Text

Susquehanna, Units 1 and 2 - Evaluation of Proposed Change Regarding Proposed License Amendment to Adopt Emergency Action Level Scheme
	ML15091A661
Person / Time
Site:	Susquehanna
Issue date:	03/19/2015
From:	Rausch T S; Susquehanna
To:	; Document Control Desk, Office of Nuclear Material Safety and Safeguards
Shared Package
ML15091A657	List: ML15091A660; ML15091A661; NEI 99-01, Susquehanna, Units 1 and 2 - Evaluation of Proposed Change Regarding Proposed License Amendment to Adopt Emergency Action Level Scheme; NEI 99-01, Susquehanna, Units 1 and 2 - Proposed License Amendment to Adopt Emergency Action Level Scheme Pursuant to NEI 99-01, Revision 6, Development of Emergency Action Levels for Non-Passive Reactors.; PLA-7285, Revision (X) to EP-RM-004, EAL Classification Bases;
References
	NEI 99-01, Rev. 6, PLA-7285
	Download: ML15091A661 (430)
	v • d • e

{{#Wiki_filter:ATTACHMENT 1 TO PLA-7285 PPL SUSQUEHANNA, LLC EVALUATION OF PROPOSED CHANGE 1. DESCRIPTION

2. BACKGROUND

3. PROPOSED CHANGE 4. TECHNICAL EVALUATION

5. REGULATORY EVALUATION 5.1 Applicable Regulatory Requirements/Criteria 5.2 Precedent 5.3 No Significant Hazards Consideration

5.4 Conclusions

6. ENVIRONMENTAL CONSIDERATIONS

7. REFERENCES Attachment 1 to PLA-7285 Page 1 of 13 PPL EVALUATION Proposed License Amendment to Adopt Emergency Action Level Scheme

1.0 DESCRIPTION

In accordance with 10 CFR 50.90, this is a request to amend Facility Operating License Nos. NPF-14 and NPF-22 for PPL Susquehanna, LLC (PPL), Susquehanna Steam Electric Station (SSES) Unit 1 and Unit 2. The proposed changes replace the approved Emergency Plan (EP) Emergency Action Level (EAL) scheme that has been based on the Nuclear Energy Institute's (NEI's) guidance established in NEI 99-01, Revision 4 (Reference 10), and the NRC approval (Reference

9) for its use at SSES, with the EAL scheme based on the guidance provided in NEI 99-01, Revision 6, (Reference 3). NEI 99-01, Revision 6 has been endorsed by the NRC (Reference 1).The proposed EAL changes were reviewed considering the requirements of 10 CFR 50.5 4 (q), paragraph (b) of 10 CFR 50.47, "Emergency plans, " 10 CFR 50 Appendix E,"Emergency Planning and Preparedness for Production and Utilization Facilities," Regulatory Issue Summary (RIS) 2003-018, "Use of NEI 99-01, Methodology for Development of Emergency Action Levels" (including supporting supplements), and RIS 2005-02, Revision 1, "Clarifying the Process for Making Emergency Plan Changes." The proposed changes to the EAL scheme contained in this submittal do not reduce the capability to meet the applicable emergency planning requirements established in 10 CFR 50.47 and 10 CFR 50, Appendix E. Adopting NEI 99-01, Revision 6 will continue to provide consistent emergency classifications to the greatest extent possible, limited only by plant-specific design or location.10 CFR 50, Appendix E, Section IV.B.2 requires prior NRC approval when a licensee is changing from one NRC-approved EAL scheme to another EAL scheme.2.0 BACKGROUND NEI 99-01, Revision 6, addresses lessons-learned since the implementation of NEI 99-01, Revision 4. The existing EAL scheme in use at SSES is based on guidance in NEI 99-01, Revision 4. In January 2003, NEI published NEI 99-01, Revision 4. In 2003, (Reference 7), the NRC endorsed NEI 99-01 Revision 4 as an acceptable method for developing standard emergency classification scheme. Subsequently, on October 27, 2003, (Reference 8), PPL submitted revised SSES EALs that were based upon the NEI 99-01 Revision 4, for NRC review and approval.

On July 21, 2004, (Reference 9), the NRC approved PPL's plan to implement the NEI 99-01, Revision 4, EAL scheme as part of Revision 45 to the SSES EP. Attachment 1 to PLA-7285 Page 2 of 13 In February 2008, NEI published NEI 99-01, Revision 5, in order to clarify the development guidance for numerous EALs, and enhance the guidance associated with the development of security-related EALs. In November 2012, NEI published NEI 99-01, Revision 6. The NRC formally endorsed the NEI 99-01, Revision 6 guidance as documented in a letter dated March 28, 2013 (Reference 1).NEI 99-01, Revision 6, represents the most recently accepted EAL methodology endorsed by the NRC. The latest revision addresses changes recommended by the NRC in a letter to NEI on October 12, 2010, along with many enhancements identified by industry during implementation of Revision 5. These enhancements include: 1. Revising EAL Basis format (Sections 5.5 through 5.11) to separate Developer Notes from Technical Basis Information.

2. Revising EAL Basis (under proposed Developer Notes) to clarify how specific instrumentation, alarms, or readings should be developed.

3. Clarifying where site-specific definitions are required (e.g., CONTAINMENT CLOSURE).4. Clarifying or proposing alternatives on the seismic and fire EALs for licensees where licensees may not have adequate instrumentation to ensure timely classification from within the Control Room.5. Revising NEI 99-01 guidance information to include a section for development of EALs applicable to new "non-passive" designs (e.g., digital instrument and controls, etc.).6. Revising front sections of the NEI 99-01 document to eliminate redundancy and inconsistency and to clearly differentiate between information that is useful for understanding how the document was put together and information that is expected to be carried over into a licensee's technical basis document.7. Conducting a review of all Unusual Events (UEs) to determine if they should be revised/eliminated or added to include a discussion of any revision proposals for the corresponding Alerts. (Note: this is primarily for events that are based upon situations where emergency response organization activation is the goal rather than a precursor to escalated EALs.)

Attachment 1 to PLA-7285 Page 3 of 13 3.0 PROPOSED CHANGES The proposed changes involve revising an EAL scheme that is currently based on NEI 99-01, Revision 4, to a scheme based on NEI 99-01, Revision 6, which has been endorsed by the NRC (Reference 1). Enhancements over earlier revision guidance (i.e., NEI 99-01, Revision 4) include: 1. Clarifying numerous EALs that have been typically misinterpreted by the industry in the development of their site-specific EAL scheme.2. Clarifying the intent of EALs that have been historically misclassified.

3. Incorporating lessons-learned from industry events (i.e., Fukushima and others)and NUREG/CR-7154, "Risk Informing Emergency Preparedness Oversight:

Evaluation of Emergency Action Levels -A Pilot Study of Peach Bottom, Surry and Sequoyah." 4. Performing a detailed review of the guidance to re-validate that the EALs are appropriate and are at the necessary emergency classification level based upon 32 years of industry and NRC experience with EAL scheme development and implementation. EAL Comparison Matrix Attachment 2 contains an EAL Comparison Matrix that has been developed to provide a tabular format of the Initiating Conditions (ICs), Mode Applicability, and EALs (Threshold Values) in NEI 99-01, Revision 6 for comparison to the proposed EALs. The matrix provides a means of assessing the proposed EAL in terms of "Differences" and"Deviations" from the NRC-endorsed guidance provided in NEI 99-01, Revision 6.The proposed EAL changes were evaluated in accordance with applicable regulatory requirements (e.g., 10 CFR 50.54(q) and Appendix E, Section IV.3.1). The evaluation assessed the conformance of the proposed EAL changes to those described in the NEI 99-01, Revision 6 guidance. The evaluation determined if the proposed EAL wording change resulted in "No Change" to the guidance, a "Difference" in the wording provided, or a "Deviation" from the NEI guidance contained in Revision 6.Any items considered to be "Differences" or "Deviations" were based on the definitions provided in RIS 2003-18, "Use of NEI 99-01, Methodology for Development of Emergency Action Levels, "and supporting supplements (References 4, 5, and 6). The RIS and supporting supplements were issued to clarify technical positions regarding the revision of EALs. Specifically, the RIS documentation provides clarification on the level of detail licensees need to provide to support the proposed changes to EALs. The RIS documents suggest that specific information be included with the EAL revision submittal to help facilitate the review process. The RIS information defines an EAL "Difference" and "Deviation" as follows: Attachment 1 to PLA-7285 Page 4 of 13 A "Difference" is an EAL change where the basis scheme guidance (e.g., NUREG, NUMARC, and NEI) differs in wording but agrees in meaning and intent, such that classification of an event would be the same, whether using the basis scheme guidance or the site-specific proposed EAL. Examples of "Differences" include the use of site-specific terminology or administrative reformatting of site-specific EALs.A "Deviation " is an EAL change where the basis scheme guidance differs in wording and is altered in meaning or intent, such that classification of the event could be different between the basis scheme guidance and the site-specific proposed EAL. Examples of "Deviations" include the use of altered mode applicability, altering key words or time limits, or changing words of physical reference (protected area, safety-related equipment, etc.).Any "Differences" identified between the NEI 99-01, Revision 6 EALs as approved by the NRC and the proposed EALs being developed by PPL in accordance with NEI 99-01, Revision 6, have been identified and are listed in the EAL Comparison Matrix (i.e., in Attachment

2) as well as the Global Differences listed below.Global Differences The following "Differences" apply throughout the set of EALs and are not specifically identified as "Differences" in the EAL Comparison Matrix: 1. The NEI phrase "Notification of Unusual Event" has been changed to "Unusual Event" or abbreviated "UE" to reduce EAL user reading burden.2. NEI 99-01 IC Example EALs are implemented in separate plant EALs to improve clarity and readability.

For example, NEI lists all IC HU3 Example EALs under one IC. The corresponding SSES EALs appear as unique EALs (e.g., HU3.1 through HU3 .4).3. Mode applicability identifiers (numbers/letter) modify the NEI mode applicability names as follows: 1 -Power Operation, 2 -Startup, 3 -Hot Shutdown, 4 -Cold Shutdown, 5 -Refueling, D -Defueled, and All. NEI 99-01 defines Defueled as follows: "Reactor Vessel contains no irradiated fuel (full core off-load during refueling or extended outage)." 4 NEI 99-01 uses the terms greater than, less than, greater than or equal to, etc. in the wording of some example EALs. For consistency and to reduce EAL-user burden, SSES has adopted use of Boolean symbols in place of the NEI 99-01 text modifiers within the EAL wording.5. "min." is the standard abbreviation for "minutes" and is used to reduce EAL user reading burden. Attachment 1 to PLA-7285 Page 5 of 13 6. IC/EAL identification: " NEI Recognition Category A "Abnormal Radiation Levels! Radiological Effluents" has been changed to Category R "Abnormal Rad Levels /Rad Effluent." The designator "R" is more intuitively associated with radiation (rad)or radiological events. NEI IC designators beginning with "A" have likewise been changed to "R."* NEI 99-01 defines the thresholds requiring emergency classification (example EALs) and assigns them to ICs which, in turn, are grouped in "Recognition Categories. "" The SSES IC/EAL scheme includes the following features: a. Division of the NEI EAL set into three groups: o EALs applicable under all plant operating modes -This group would be reviewed by the EAL-user any time emergency classification is considered. o EALs applicable only under hot operating modes -This group would only be reviewed by the EAL-user when the plant is in Hot Shutdown, 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.

b. Within each of the above three groups, assignment of EALs to categories/subcategories

-Category and subcategory titles are selected to represent conditions that are operationally significant to the EAL-user.Subcategories are used as necessary to further divide the EALs of a category into logical sets of possible emergency classification thresholds. The SSES EAL categories/subcategories and their relationship to NEI Recognition Categories are listed in Table 1 of Attachment 2.c. Unique identification of each EAL -Four characters comprise the EAL identifier as illustrated in Figure 1 of Attachment

2.

Attachment I to PLA-7285 Page 6 of 13 Figure 1 -EAL Identifier EAL Identifier XXX.X Category (R, H. E, S, F, C) -J L Sequential number within subcategory/classirlcation Emergency classification (G, S. A. U) Subcategory number (1 if no subcategory) The first character is a letter associated with the category in which the EAL is located. The second character is a letter associated with the emergency classification level (G for General Emergency, S for Site Area Emergency, A for Alert, and U for Notification of Unusual Event). The third character is a number associated with one or more subcategories within a given category.Subcategories are sequentially numbered beginning with the number "1". If a category does not have a subcategory, this character is assigned the number"1". The fourth character is a number preceded by a period for each EAL within a subcategory. EALs are sequentially numbered within the emergency classification level of a subcategory beginning with the number "1".The EAL identifier is designed to fulfill the following objectives: o Uniqueness -The EAL identifier ensures that there can be no confusion over which EAL is driving the need for emergency classification. o Speed in locating the EAL of concern -When the EALs are displayed in a matrix format, knowledge of the EAL identifier alone can lead the EAL-user to the location of the EAL within the classification matrix. The identifier conveys the category, subcategory and classification level. This assists ERO responders (who may not be in the same facility as the ED/RM) to find the EAL of concern in a timely manner without the need for a word description of the classification threshold. o Possible classification upgrade -The category/subcategory/identifier scheme helps the EAL-user find higher emergency classification EALs that may become active if plant conditions worsen.Table 2 of Attachment 2 lists the SSES ICs and EALs that correspond to the NEI ICs/Example EALs when the above EAL/IC organization and identification scheme is implemented. PPL determined that these "Differences" do not result in a reduction in effectiveness or change the intent of the new NEI 99-01, Revision 6 EALs.Any plant EAL (IC or Threshold Value) that does not meet the "intent" of the NEI 99-01, Revision 6 guidance, or may result in an event being classified differently from the guidance, would be identified as a "Deviation. " The evaluation determined that there are no "Deviations" in converting from the existing EALs based on NEI 99-01, Revision 4 as currently approved, to an EAL scheme based on the NEI 99-01, Revision 6 guidance. Attachment 1 to PLA-7285 Page 7 of 13 The Emergency Plan contains the station's EALs. The proposed EAL changes are discussed in Attachments 2, 3 and 4.4.0 TECHNICAL EVALUATION The proposed changes have been evaluated to determine whether applicable regulations and requirements continue to be met. NEI 99-01 guidance methodology includes many years of development along with use and implementation. The guidance has been subject to NRC reviews and approval. The PPL EAL scheme currently in place is based on the EAL methodology outlined in NEI 99-01, Revision 4. NEI 99-01, Revision 6 is the latest guidance endorsed by the NRC and provides guidance to nuclear power plant operators for the development of a site-specific emergency classification scheme.10 CFR 50.47(b)(4) stipulates that Emergency Plans include a standard emergency classification and action level scheme. This scheme is a fundamental component of an Emergency Plan, in that it provides the defined thresholds that will allow site personnel to rapidly implement a range of pre-planned emergency response measures. An emergency classification scheme also facilitates timely decision-making by an Offsite Response Organization (ORO) concerning the implementation of precautionary or protective actions for the public.NEI 99-01, Revision 6 contains a set of generic ICs, EALs, and fission product barrier status thresholds. It also includes supporting technical basis information, developer notes, and recommended classification instructions for users. The methodology described in this document is consistent with NRC requirements and guidance. In particular, this methodology was specifically endorsed by the NRC as documented in a March 28, 2013, letter (Reference

1) and determined to provide an acceptable approach in meeting the requirements of 10 CFR 50.47(b)(4), applicable requirements of 10 CFR 50, Appendix E, and the associated planning standard evaluation elements established in NUREG-0654/FEMA-REP-1, Revision 1, "Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants, "dated November 1980.10 CFR 50, Appendix E, Section IV.B.2 stipulates that a licensee desiring to change its entire EAL scheme shall submit an application for an amendment to its license and receive NRC approval before implementing the change. RIS 2005-02, Revision 1 (Reference

2) also indicates that a revision to an entire EAL scheme to another NRC-endorsed EAL scheme, must be submitted for prior NRC approval as specified in Section IV.B of Appendix E to 10 CFR 50.The proposed changes to the EAL scheme for adopting the NEI 99-01, Revision 6 guidance do not reduce the capability to meet the applicable emergency planning requirements established in 10 CFR 50.47 and 10 CFR 50, Appendix E.

Attachment I to PLA-7285 Page 8 of 13 The proposed changes to adopt the NEI 99-01, Revision 6, EAL scheme will continue to provide consistent emergency classifications to the greatest extent possible, limited only by plant-specific design or location. Changes to the Emergency Plan and procedures resulting from implementation of the revised EALs will be evaluated in accordance with the requirement of 10 CFR 50.54(q), subsequent to NRC approval.Accordingly, pursuant to the requirements of 10 CFR 50, Appendix E, Section IV.B.2, PPL requests NRC review and approval of the proposed changes to the EAL scheme as license amendment requests for the Facility Operating Licenses of SSES Unit I and Unit 2, in accordance with 10 CFR 50.90.

5.0 REGULATORY EVALUATION

5.1 Applicable Regulatory Reg uirements/Criteria The proposed changes have been evaluated to determine whether applicable regulations and requirements continue to be met.The regulations in 10 CFR 50.54(q) provide direction to licensees seeking to revise their Emergency Plan. The requirements related to nuclear power plant Emergency Plans are given in the standards in 10 CFR 50.47, "Emergency plans," and the requirements of Appendix E, "Emergency Planning and Preparedness for Production and Utilization Facilities, " to 10 CFR 50.Paragraph (a)(1) to 10 CFR 50.47 states that no operating license for a nuclear power reactor will be issued unless a finding is made by the NRC that there is reasonable assurance that adequate protective measures can and will be taken in the event of a radiological emergency. Section 50.47 establishes standards that onsite and offsite emergency response plans must meet for the NRC staff to make a positive finding that there is reasonable assurance that adequate protective measures can and will be taken in the event of a radiological emergency. One of these standards, 10 CFR 50.47(b)(4), stipulates that Emergency Plans include a standard emergency classification and action level scheme.Section IV.B, "Assessment Actions, "to 10 CFR 50, Appendix E, "Emergency Planning and Preparedness for Production and Utilization Facilities, " stipulates that Emergency Plans include EALs, which are to be used as criteria for determining the need for notification and participation of local and State agencies, and for determining when and what type of protective measures should be considered to protect the health and safety of individuals both onsite and offsite.EALs are to be based on plant conditions and instrumentation, as well as onsite and offsite radiological monitoring. Section IV.B of Appendix E provides that initial EALs shall be discussed and agreed on by the applicant and State and local authorities, be approved by the NRC, and reviewed annually thereafter with State and local authorities. Therefore, a revision will require NRC approval prior to Attachment 1 to PLA-7285 Page 9 of 13 implementation if it involves:

1) changing from one EAL scheme to another, such as from an EAL scheme based on NUJREG-0654/FEMA-REP-1 to one based on NUMARC/NESP-007 or NEI 99-01; 2) the licensee is proposing an alternate method for complying with the regulations; or 3) the EAL revision proposed by the licensee decreases the effectiveness of the Emergency Plan.NRC RIS 2005-02, Revision 1, "Clarifying the Process for Making Emergency Plan Changes, " (Reference

2) issued in August 2009, also discusses that a change in an EAL scheme to incorporate the improvements provided in NUMARC/NESP-007 or NEI 99-01 would not decrease the overall effectiveness of the Emergency Plan and would not expand a licensee's operating authority beyond that previously authorized by NRC. However, due to the potential safety significance of the change, the change needs prior NRC review and approval.This approval would be granted via an NRC letter and supporting Safety Evaluation Report (SER).PPL has determined that the proposed changes do not require any exemptions or relief from regulatory requirements and do not affect conformance with any General Design Criteria differently than described in the SSES Final Safety Analysis Report (FSAR).5.2 Precedent NEI 99-01, Revision 6 is the most recent NRC endorsed guidance for the EAL scheme; and as such, there has been considerable precedence of licensees submitting license amendment requests for an EAL scheme change that adopt the NEI 99-01, Revision 6 guidance.

Attachment 1 to PLA-7285 Page 10 of 13 5.3 No Significant Hazards Consideration In accordance with 10CFR 50.90, "Application for amendment of license, construction permit, or early site permit, "PPL Susquehanna, LLC (PPL) requests amendments to the Facility Operating Licenses NPF-14 and NPF-22 for the Susquehanna Steam Electric Station (SSES) Unit 1 and Unit 2. The requests support the adoption of Emergency Action level (EAL) scheme based on NEI 99-01, Revision 6, which has been endorsed by the NRC as documented in a letter dated March 28, 2013 (Reference 1).The proposed changes to the EAL scheme to adopt the guidance in NEI 99-01, Revision 6 do not reduce the capability to meet the emergency planning requirements established in 10 CFR 50.47 and 10 CFR 50, Appendix E. The proposed changes do not reduce the functionality, performance, or capability of the SSES Emergency Response Organization (ERO) to respond in mitigating the consequences of accidents. All ERO functions will continue to be performed as required.The proposed changes have been reviewed considering the applicable requirements of 10 CFR 50.47, 10 CFR 50, Appendix E, and other applicable NRC documents. PPL has evaluated the proposed changes to the Emergency Plans and determined that the changes do not involve a Significant Hazards Consideration. In support of this determination, an evaluation of each of the three (3) standards, set forth in 10 CFR 50.92, "Issuance of amendment, "is provided below.1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated? No. The proposed changes to the EAL scheme to adopt the NRC-endorsed guidance in NEI 99-01, Revision 6, "Development of Emergency Action Levels for Non-Passive Reactors, "do not reduce the capability to meet the emergency planning requirements established in 10 CFR 50.47 and 10 CFR 50, Appendix E. The proposed changes do not reduce the functionality, performance, or capability of the ERO to respond in mitigating the consequences of any design basis accident.The probability of a reactor accident requiring implementation of Emergency Plan EALs has no relevance in determining whether the proposed changes to the EALs reduce the effectiveness of the Emergency Plan. As discussed in Section I.D, "Planning Basis, "ofNUREG-0654, Revision 1, "Criteria for Attachment 1 to PLA-7285 Page 11 of 13 Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants;""... The overall objective of emergency response plans is to provide dose savings (and in some cases immediate life saving) for a spectrum of accidents that could produce offsite.doses in excess of Protective Action Guides (PAGs). No single specific accident sequence should be isolated as the one for which to plan because each accident could have different consequences, both in nature and degree. Further, the range ofpossible selection for a planning basis is very large, starting with a zero point of requiring no planning at all because significant offsite radiological accident consequences are unlikely to occur, to planning for the worst possible accident, regardless of its extremely low likelihood... Therefore, risk insights are not considered for any specific accident initiation or progression in evaluating the proposed changes.The proposed changes do not involve any physical changes to plant equipment or systems, nor do they alter the assumptions of any accident analyses. The proposed changes do not adversely affect accident initiators or precursors nor do they alter the design assumptions, conditions, and configuration or the manner in which the plants are operated and maintained. The proposed changes do not adversely affect the ability of Structures, Systems, or Components (SSCs) to perform their intended safety functions in mitigating the consequences of an initiating event within the assumed acceptance limits.Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

No. The proposed changes to the EAL scheme to adopt the NRC-endorsed guidance in NEI 99-01, Revision 6 do not involve any physical changes to plant systems or equipment. The proposed changes do not involve the addition of any new plant equipment. The proposed changes will not alter the design configuration, or method of operation of plant equipment beyond its normal functional capabilities. All ERO functions will continue to be performed as required. The proposed changes do not create any new credible failure mechanisms, malfunctions, or accident initiators. Therefore, the proposed changes do not create the possibility of a new or different kind of accident from those that have been previously evaluated. Attachment 1 to PLA-7285 Page 12 of 13 3. Does the proposed change involve a significant reduction in a margin of safety?No. The proposed changes to the EAL scheme to adopt the NRC-endorsed guidance in NEI 99-01, Revision 6 do not alter or exceed a design basis or safety limit. There is no change being made to safety analysis assumptions, safety limit, or limiting safety system settings that would adversely affect plant safety as a result of the proposed changes. There are no changes to setpoints or environmental conditions of any SSC or the manner in which any SSC is operated. Margins of safety are unaffected by the proposed changes to adopt the NEI 99-01, Revision 6 EAL scheme guidance. The applicable requirements of 10 CFR 50.47 and 10 CFR 50, Appendix E will continue to be met.Therefore, the proposed changes do not involve any reduction in a margin of safety.Based upon the above responses, PPL concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), "Issuance of Amendment, " and, accordingly, a finding of no significant hazards consideration is justified.

5.3 Conclusions

In conclusion, and based on the considerations discussed above: (1) there is reasonable assurance that the health and safety of the public will not be endangered by the proposed changes to adopt the EAL scheme established in NEI 99-01, Revision 6, as endorsed by the NRC; (2) the changes will be in compliance with the NRC's regulations; and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public, 6.0 ENVIRONMENTAL CONSIDERATION The proposed changes are applicable to emergency planning requirements involving the proposed adoption of NRC-approved EAL guidance as described in NEI 99-0 1, Revision 6 and do not reduce the capability to meet the emergency planning standards established in 10 CFR 50.47 and 10 CFR 50, Appendix E. The proposed changes do not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in the individual or cumulative occupational radiation exposure. Attachment 1 to PLA-7285 Page 13 of 13 Accordingly, the proposed changes meet the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed change.

7.0 REFERENCES

1. Letter from Marc Thaggard to Susan Perkins-Grew, NEI -U.S. Nuclear Regulatory Commission Review and Endorsement of NEI 99-01, Revision 6, November 2012, dated March 28, 2013 (ML 12346A463) and (NRC/NSIR"Regulatory and Technical Evaluation for Endorsement," 11/2012 Version, ML13008A736)

2. RIS 2005-02, Revision 1, "Clarifying the Process for Making Emergency Plan Changes, "dated August 19, 2011 3. NEI 99-01, Revision 6, "Development of Emergency Action Levels for Non-Passive Reactors, "dated November 2012 (ML12326A805)

4. RIS 2003-18, "Use of NEI 99-01, Methodology for Development of Emergency Action Levels," dated October 8, 2003.5. RIS 2003-18, Supplement 1, "Use of Nuclear Energy Institute (NEI) 99-01, Methodology for Development of Emergency Action Levels, "dated July 13, 2004 6. RIS 2003-18, Supplement 2, "Use of Nuclear Energy Institute (NEI) 99-01, Methodology for Development of Emergency Action Levels, "dated December 12, 2005 7. ML031250542, "Regulatory Analysis, Revision 4 of RG 1.101 to Accept the Guidance in NEI 99-01 as an Alternative Methodology for the Development of Emergency Action Levels, " dated May 14, 2003 [This endorses the guidance contained in NEI 99-01, Revision 4]8. PPL letter PLA-5632, "Revision to Emergency Action Levels, "dated October 27, 2003 9. NRC letter to SSES Units 1 and 2 -"Proposed Revision to Emergency Action Levels (TAC Nos. MC12 70 and MC1271), "dated July 21, 2004 (Accession No.ML042030383)

10. NEI 99-01, Revision 4, "Methodology for Development ofEmergency Action Levels, "(NUMARC/NESP-007), dated January 2003 (ML041470143)

ATTACHMENT 2 TO PLA-7285 PPL SUSQUEHANNA, LLC EAL COMPARISON MATRIX REVISION 0 Susquehanna Steam Electric Station NEI 99-01 Revision 6 EAL Comparison Matrix Revision 0 SSES EAL Comparison Matrix Table of Contents Section Paqe I 1r 1o CL i o[ I ---------.-.--------.-.-------.-.--.-.--.-.-------.-.-.-.--.-.--.--------.-.-.. ......... ....... ..- ------- -------1 Comparison Matrix Format -......................... ...--- -- -- ----- .-- -EAL Emphasis Techniques---

Global Differences---

Differences and Deviations

-....----------------------------........------- Category R -Abnormal Rad Levels / Rad Effiuent-.......- Category C -Cold Shutdown / Refueling System Malfunction-- Category D -Permanently Defueled Station Malfunction------- Category E -Independent Spent Fuel Storage Installation (ISFSI)-Category F -Fission Product Barrier Degradation Category H -Hazards and Other Conditions Affecting Plant Safety Category S -System Malfunction

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85 Table 1 -SSES EAL Categories/Subcategories-----------

Table 2 -NEI / SSES EAL Identification Cross-Reference Table 3 -Summary of Deviations--------

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-6 6 11---------------------------------------------------------------------------------- i ofi SSES EAL Comparison Matrix Introduction This document provides a line-by-line comparison of the Initiating Conditions (ICs), Mode Applicability and Emergency Action Levels (EALs) in NEI 99-01 Rev. 6 Final, Development of Emergency Action Levels for Non-Passive Reactors, ADAMS Accession Number ML1 10240324, and the Susquehanna Steam Electric Station (SSES) ICs, Mode Applicability and EALs. This document provides a means of assessing SSES differences and deviations from the NRC endorsed guidance given in NEI 99-01. Discussion of SSES EAL bases and lists of source document references are given in the EAL Technical Bases Document. It is, therefore, advisable to reference the EAL Technical Bases Document for background information while using this document. As shown in Table 3, SSES took no deviations from the generic NEI 99-01 Revision 6 guidance.Comparison Matrix Format The ICs and EALs discussed in this document are grouped according to NEI 99-01 Recognition Categories. Within each Recognition Category, the ICs and EALs are listed in tabular format according to the order in which they are given in NEI 99-01. Generally, each row of the comparison matrix provides the following information:

NEI EAL/IC identifier
NEI EAL/IC wording* SSES EAL/IC identifier
SSES EAL/IC wording* Description of any differences or deviations EAL Emphasis Techniques Due to the width of the table columns and table formatting constraints in this document, line breaks and indentation may differ slightly from the appearance of comparable wording in the source documents.

NEI 99-01 is the source document for the NEI EALs; the SSES EAL Technical Bases Document for the SSES EALs.The print and paragraph formatting conventions summarized below guide presentation of the SSES EALs in accordance with the EAL writing criteria.Space restrictions in the EAL table of this document sometimes override these criteria in cases when following the criteria would introduce undesirable complications in the EAL layout." Upper case-bold print is used for the logic terms AND, OR and EITHER." Bold font is used for certain logic terms, negative terms (not, cannot, etc.), any, all.* Upper case print is reserved for defined terms, acronyms, system abbreviations, logic terms (and, or, etc. when not used as a conjunction), annunciator window engravings." Three or more items in a list are normally introduced with "Any of the following..." or "All of the following..." Items of the list begin with bullets when a priority or sequence is not inferred.* The use of AND/OR logic within the same EAL has been avoided when possible. When such logic cannot be avoided, indentation and separation of subordinate contingent phrases is employed.Global Differences The differences listed below generally apply throughout the set of EALs and are not repeated in the Justification sections of this document. The global differences do not decrease the effectiveness of the intent of NEI 99-01.1. The NEI phrase "Notification of Unusual Event" has been changed to"Unusual Event" or abbreviated "UE" to reduce EAL-user reading burden.2. NEI 99-01 IC Example EALs are implemented in separate plant EALs to improve clarity and readability. For example, NEI lists all IC HU3 Example EALs under one IC. The corresponding SSES EALs appear as unique EALs (e.g., HU3.1 through HU3.4).3. Mode applicability identifiers (numbers/letter) modify the NEI 99-01 mode applicability names as follows: 1 -Power Operation, 2 -Startup, 3 -Hot Shutdown, 4 -Cold Shutdown, 5 -Refueling, D -Defueled, and All. NEI 99-01 defines Defueled as follows: "Reactor Vessel contains no irradiated fuel (full core off-load during refueling or extended outage)." 4. NEI 99-01 uses the terms greater than, less than, greater than or equal to, etc. in the wording of some example EALs. For consistency 1 of 110 SSES EAL Comparison Matrix and to reduce EAL-user reading burden, SSES has adopted use of Boolean symbols in place of the NEI 99-01 text modifiers within the EAL wording.5. "min." is the standard abbreviation for "minutes" and is used to reduce EAL user reading burden.6. IC/EAL identification:

NEI Recognition Category A "Abnormal Radiation Levels/Radiological Effluents" has been changed to Category R"Abnormal Rad Levels / Rad Effluent." The designator "R" is more intuitively associated with radiation (rad) or radiological events. NEI IC designators beginning with "A" have likewise been changed to "R."* NEI 99-01 defines the thresholds requiring emergency classification (example EALs) and assigns them to ICs which, in turn, are grouped in "Recognition Categories."* The SSES IC/EAL scheme includes the following features: a. Division of the NEI EAL set into three groups: o EALs applicable under all plant operating modes -This group would be reviewed by the EAL-user any time emergency classification is considered.

o EALs applicable only under hot operating modes -This group would only be reviewed by the EAL-user when the plant is in Hot Shutdown, Startup or Power Operation mode.o EALs applicable only under cold operating modes7-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.

b. Within each of the above three groups, assignment of EALs to categories/subcategories

-Category and subcategory titles are selected to represent conditions that are operationally significant to the EAL-user.Subcategories are used as necessary to further divide the EALs of a category into logical sets of possible emergency classification thresholds. The SSES EAL categories/subcategories and their relationship to NEI Recognition Categories are listed in Table 1.c. Unique identification of each EAL- Four characters comprise the EAL identifier as illustrated in Figure 1.Figure 1 -EAL Identifier EAL Identifier XXX.X Category (R, H, E. S. F. C) L Sequential numberwithin SUbcategory/ciasslflcation Emergency classification (G S.A. U I I Subcategory number (I if no subcategory) The first character is a letter associated with the category in which the EAL is located. The second character is a letter associated with the emergency classification level (G for General Emergency, S for Site Area Emergency, A for Alert, and U for Notification of Unusual Event). The third character is a number associated with one or more subcategories within a given category. Subcategories are sequentially numbered beginning with the number "1". If a category does not have a subcategory, this character is assigned the number "1". The fourth character is a number preceded by a period for each EAL within a subcategory. EALs are sequentially numbered within the emergency classification level of a subcategory beginning with the number "1".The EAL identifier is designed to fulfill the following objectives: 2 of 110 SSES EAL Comparison Matrix o Uniqueness -The EAL identifier ensures that there can be no confusion over which EAL is driving the need for emergency classification. o Speed in locating the EAL of concern -When the EALs are displayed in a matrix format, knowledge of the EAL identifier alone can lead the EAL-user to the location of the EAL within the classification matrix. The identifier conveys the category, subcategory and classification level. This assists ERO responders (who may not be in the same facility as the ED/RM) to find the EAL of concern in a timely manner without the need for a word description of the classification threshold. o Possible classification upgrade -The category/subcategory/identifier scheme helps the EAL-user find higher emergency classification EALs that may become active if plant conditions worsen.Table 2 lists the SSES ICs and EALs that correspond to the NEI ICs/Example EALs when the above EAL/IC organization and identification scheme is implemented. Differences and Deviations In accordance NRC Regulatory Issue Summary (RIS) 2003-18 "Use of Nuclear Energy Institute (NEI) 99-01, Methodology for Development of Emergency Action Levels" Supplements 1 and 2, a difference is an EAL change in which the basis scheme guidance differs in wording but agrees in meaning and intent, such that classification of an event would be the same, whether using the basis scheme guidance or the SSES EAL. A deviation is an EAL change in which the basis scheme guidance differs in wording and is altered in meaning or intent, such that classification of the event could be different between the basis scheme guidance and the SSES proposed EAL.Administrative changes that do not actually change the textual content are neither differences nor deviations. Likewise, any format change that does not alter the wording of the IC or EAL is considered neither a difference nor a deviation. The following are examples of differences:

Choosing the applicable EAL based upon plant type (i.e., BWR vs.PWR)." Using a numbering scheme other than that provided in NEI 99-01 that does not change the intent of the overall scheme.* Where the NEI 99-01 guidance specifically provides an option to not include an EAL if equipment for the EAL does not exist at SSES (e.g., automatic real-time dose assessment capability).
Pulling information from the bases section up to the actual EAL that does not change the intent of the EAL.* Choosing to state ALL Operating Modes are applicable instead of stating N/A, or listing each mode individually under the Abnormal Rad Level/Radiological Effluent and Hazard and Other Conditions Affecting Plant Safety sections.* Using synonymous wording (e.g., greater than or equal to vs. at or above, less than or equal vs. at or below, greater than or less than vs. above or below, etc.)* Adding SSES equipment/instrument identification and/or noun names to EALs." Combining like ICs that are exactly the same but have different operating modes as long as the intent of each IC is maintained and the overall progression of the EAL scheme is not affected." Any change to the IC and/or EAL, and/or basis wording, as stated in NEI 99-01, that does not alter the intent of the IC and/or EAL, i.e., the IC and/or EAL continues to: o Classify at the correct classification level.o Logically integrate with other EALs in the EAL scheme." Ensure that the resulting EAL scheme is complete (i.e., classifies all potential emergency conditions).

The following are examples of deviations: " Use of altered mode applicability.

Altering key words or time limits.* Changing words of physical reference (protected area, safety-related equipment, etc.).3 of 110 SSES EAL Comparison Matrix" Eliminating an IC. This includes the removal of an IC from the Fission Product Barrier Degradation category as this impacts the logic of Fission Product Barrier ICs." Changing a Fission Product Barrier from a Loss to a Potential Loss or vice-versa.
Not using NEI 99-01definitions as the intent is for all NEI 99-01 users to have a standard set of defined terms as defined in NEI 99-01.Differences due to plant types are permissible (BWR or PWR).Verbatim compliance to the wording in NEI 99-01 is not necessary as long as the intent of the defined word is maintained.

Use of the wording provided in NEI 99-01 is encouraged since the intent is for all users to have a standard set of defined terms as defined in NEI 99-01.* Any change to the IC and/or EAL, and/or basis wording as stated in NEI 99-01 that does alter the intent of the IC and/or EAL, i.e., the IC and/or EAL: o Does not classify at the classification level consistent with NEI 99-01.o Is not logically integrated with other EALs in the EAL scheme.o Results in an incomplete EAL scheme (i.e., does not classify all potential emergency conditions). The "Difference/Deviation Justification" columns in the remaining sections of this document identify each difference between the NEI 99-01 IC/EAL wording and the SSES IC/EAL wording. An explanation that justifies the reason for each difference is then provided. If the difference is determined to be a deviation, a statement is made to that affect and explanation is given that states why classification may be different from the NEI 99-01 IC/EAL and the reason for its acceptability. In all cases, however, the differences and deviations do not decrease the effectiveness of the intent of NEI 99-01. A summary list of SSES EAL deviations from NEI 99-01 is given in Table 3.4 of 110 SSES EAL Comparison Matrix Table 1 -SSES EAL Categories/Subcategories SSES EALs NEI Category I Subcategory Recognition Category Group: All Operating Mode: 1 -Radiological Effluent Abnormal Rad Levels/Radiological Effluent R -Abnormal Rad Levels/Rad Effluent 2 -Irradiated Fuel Event ICs/EALs 3- Area Radiation Levels 1 -Security Hazards and Other Conditions Affecting 2 -Seismic Event 2 -eisic EentPlant Safety ICs/EALs H -Hazards and Other Conditions Affecting 3 -Natural or Technological Hazard Plar Sanethr C4 -Fire Plant Safety 5 -Hazardous Gas 6 -Control Room Evacuation 7 -ED/RM Judgment E -ISFSI 1 -Confinement Boundary ISFSI ICs/EALs Group: Hot Conditions: 1 -Loss of Essential AC Power System Malfunction ICs/EALs 2 -Loss of Vital DC Power 3 -Loss of Control Room Indications S -System Malfunction 4 -RCS Activity 5 -RCS Leakage 6 -RPS Failure 7 -Loss of Communications 8 -Hazardous Event Affecting Safety Systems F -Fission Product Barrier Degradation None Fission Product Barrier ICs/EALs Group: Cold Conditions: 1 -RPV Level Cold Shutdown./ Refueling System 2 -Loss of Essential AC Power Malfunction ICs/EALs C -Cold Shutdown I Refueling System 3 -RCS Temperature Malfunction 4 -Loss of Vital DC Power 5 -Loss of Communications 6 -Hazardous Event Affecting Safety Systems 5 of 110 SSES EAL Comparison Matrix Table 2 -NEI / SSES EAL Identification Cross-Reference NEI SSES Example Category and Subcategory EAL AU1 1 R- Abnormal Rad Levels / Rad Effluent, I -Radiological Effluent RUI.1 AU1 2 R -Abnormal Rad Levels / Rad Effluent, 1 -Radiological Effluent RUI.1 AU1 3 R -Abnormal Rad Levels / Rad Effluent, 1 -Radiological Effluent RU1.2 AU2 1 R -Abnormal Rad Levels / Rad Effluent, 2 -Irradiated Fuel Event RU2.1 AA1 1 R -Abnormal Rad Levels / Rad Effluent, 1 -Radiological Effluent RAI.1 AA1 2 R -Abnormal Rad Levels / Rad Effluent, 1 -Radiological Effluent RA1.2 AA1 3 R -Abnormal Rad Levels / Rad Effluent, 1 -Radiological Effluent RA1.3 AA1 4 R -Abnormal Rad Levels / Rad Effluent, 1 -Radiological Effluent RA1.4 AA2 1 R -Abnormal Rad Levels / Rad Effluent, 2 -Irradiated Fuel Event RA2.1 AA2 2 R -Abnormal Rad Levels / Rad Effluent, 2 -Irradiated Fuel Event RA2.2 AA2 3 R- Abnormal Rad Levels / Rad Effluent, 2 -Irradiated Fuel Event RA2.3 AA3 1 R -Abnormal Rad Levels / Rad Effluent, 3 -Area Radiation Levels RA3.1 AA3 2 R -Abnormal Rad Levels / Rad Effluent, 3 -Area Radiation Levels RA3.2 AS1 1 R -Abnormal Rad Levels / Rad Effluent, 1 -Radiological Effluent RS1.1 AS1 2 R -Abnormal Rad Levels / Rad Effluent, 1 -Radiological Effluent RS1.2 AS1 3 R -Abnormal Rad Levels / Rad Effluent, 1 -Radiological Effluent RS1.3 6 of 110 SSES EAL Comparison Matrix NEI SSES Example Category and Subcategory EAL* IC EAL AS2 1 R -Abnormal Rad Levels/ Rad Effluent, 2- Irradiated Fuel Event RS2.1 AG1 1 R -Abnormal Rad Levels / Rad Effluent, 1 -Radiological Effluent RG1.1 AG1 2 R -Abnormal Rad Levels / Rad Effluent, 1 -Radiological Effluent RG1.2 AG1 3 R -Abnormal Rad Levels / Rad Effluent, 1 -Radiological Effluent RG1.3 AG2 1 R -Abnormal Rad Levels / Rad Effluent, 2 -Irradiated Fuel Event RG2.1 CUl 1 C -Cold SD/ Refueling System Malfunction, 1 -RPV Level CU1.1 CU1 2 C -Cold SD/ Refueling System Malfunction, 1 -RPV Level CU1.2 CU2 1 C -Cold SD/ Refueling System Malfunction, 2 -Loss of Essential AC Power CU2.1 CU3 1 C -Cold SD/ Refueling System Malfunction, 3 -RCS Temperature CU4.1 CU3 2 C -Cold SD/ Refueling System Malfunction, 3 -RCS Temperature CU4.2 CU4 1 C -Cold SD/ Refueling System Malfunction, 4 -Loss of Essential DC Power CU3.1 CU5 1, 2, 3 C -Cold SD/ Refueling System Malfunction, 5 -Loss of Communications CU5.1 CAI 1 C -Cold SD/ Refueling System. Malfunction, I -RPV Level CA1.1 CAl 2 C -Cold SD/ Refueling System Malfunction, 1 -RPV Level CA1.2 CA2 1 C -Cold SD/ Refueling System Malfunction, 1 -Loss of Essential AC Power CA2.1 CA3 1, 2 C -Cold SD/ Refueling System Malfunction, 3 -RCS Temperature CA3.1 CA6 1 C -Cold SD/ Refueling System Malfunction, 6 -Hazardous Event Affecting Safety Systems CA6.1 CS1 1 C -Cold SD/ Refueling System Malfunction, 1 -RPV Level CS1.1 7 of 110 SSES EAL Comparison Matrix NEI SSES IC Example ExEAL Category and Subcategory EAL CS1 2 C -Cold SD/ Refueling System Malfunction, 1 -RPV Level CS1.2 CS1 3 C -Cold SD/ Refueling System Malfunction, 1 -RPV Level CS1.3 CG1 1 C -Cold SD/ Refueling System Malfunction, 1 -RPV Level CG1.1 CG1 2 C -Cold SD/ Refueling System Malfunction, 1 -RPV Level CG1.2 E-HU1 1 E -ISFSI EU1.1 FAI 1 F -Fission Product Barrier Degradation FA1.1 FS1 1 F -Fission Product Barrier Degradation FS1.1 FG1 1 F -Fission Product Barrier Degradation FG1.1 HUL1 1, 2, 3 H -Hazards and Other Conditions Affecting Plant Safety, I -Security HU1.1 HU2 1 H -Hazards and Other Conditions Affecting Plant Safety, 2 -Seismic Event HU2.1 HU3 1 H -Hazards and Other Conditions Affecting Plant Safety, 3 -Natural or Technological Hazard HU3.1 HU3 2 H -Hazards and Other Conditions Affecting Plant Safety, 3 -Natural or Technological Hazard HU3.2 HU3 3 H -Hazards and Other Conditions Affecting Plant Safety, 3 -Natural or Technological Hazard HU3.3 HU3 4 H -Hazards and Other Conditions Affecting Plant Safety, 3 -Natural or Technological Hazard HU3.4 HU3 5 N/A N/A HU4 1 H -Hazards and Other Conditions Affecting Plant Safety, 4 -Fire HU4.1 HU4 2 H -Hazards and Other Conditions Affecting Plant Safety, 4 -Fire HU4.2 HU4 3 H -Hazards and Other Conditions Affecting Plant Safety, 4 -Fire HU4.3 8 of 110 SSES EAL Comparison Matrix NEI SSES Example Category and Subcategory EAL HU4 4 H -Hazards and Other Conditions Affecting Plant Safety, 4 -Fire HU4.4 HU7 1 H -Hazards and Other Conditions Affecting Plant Safety, 7 -ED/RM Judgment HU7.1 HAl 1, 2 H -Hazards and Other Conditions Affecting Plant Safety, 1 -Security HAl.1 HA5 1 H -Hazards and Other Conditions Affecting Plant Safety, 5 -Hazardous Gas HA5.1 HA6 1 H -Hazards and Other Conditions Affecting Plant Safety, 6 -Control Room Evacuation HA6.1 HA7 1 H -Hazards and Other Conditions Affecting Plant Safety, 7 -ED/RM Judgment HA7.1 HS1 1 H -Hazards and Other Conditions Affecting Plant Safety, 1 -Security HS1.1 HS6 1 H -Hazards and Other Conditions Affecting Plant Safety, 6 -Control Room Evacuation HS6.1 HS7 1 H -Hazards and Other Conditions Affecting Plant Safety, 7 -ED/RM Judgment HS7.1 HGI 1 H -Hazards and Other Conditions Affecting Plant Safety, 1 -Security HG1.1 HG7 1 H -Hazards and Other Conditions Affecting Plant Safety, 7 -ED/RM Judgment HG7.1 SUl 1 S -System Malfunction, 1 -Loss of Essential AC Power SU1.1 SU2 I S -System Malfunction, 3 -Loss of Control Room Indications SU3.1 SU3 1 S -System Malfunction, 4 -RCS Activity SU4.1 SU3 2 S -System Malfunction, 4 -RCS Activity SU4.2 SU4 1, 2, 3 S -System Malfunction, 5 -RCS Leakage SU5.1 SU5 1 S -System Malfunction, 6 -RPS Failure SU6.1 SU5 2 S -System Malfunction, 6 -RPS Failure SU6.2 9 of 110 SSES EAL Comparison Matrix NEI SSES Example Category and Subcategory EAL SU6 1, 2, 3 S -System Malfunction, 7 -Loss of Communications SU7.1 SU7 1, 2 N/A (PWR only) N/A SA1 1 S -System Malfunction, 1 -Loss of Essential AC Power SAI.1 SA2 1 S -System Malfunction, 3 -Loss of Control Room Indications SA3.1 SA5 1 S -System Malfunction, 6 -RPS Failure SA6.1 SA9 1 S -System Malfunction, 8 -Hazardous Event Affecting Safety Systems SA8.1 SS1 I S -System Malfunction, 1 -Loss of Essential AC Power SSI.1 SS5 1 S -System Malfunction, 6 -RPS Failure SS6.1 SS8 1 S -System Malfunction, 2 -Loss of Essential DC Power SS2.1 SG1 1 S -System Malfunction, 1 -Loss of Essential AC Power SG1.1 SG8 2 S -System Malfunction, 1 -Loss of Essential AC Power SG1.2 10 of 110 SSES EAL Comparison Matrix Table 3 -Summary of Deviations 11 of 110 SSES EAL Comparison Matrix Category R Abnormal Rad Levels I Radiological Effluent 12 of 110 SSES EAL Comparison Matrix NEI IC# NEI IC Wording and Mode SSES SSES IC Wording and Mode Difference/Deviation Justification Applicability IC#(s) Applicability AU1 Release of gaseous or liquid RU1 Release of gaseous or liquid The SSES TRM is the site-specific effluent release controlling radioactivity greater than 2 times radioactivity greater than 2 times the document.the (site-specific effluent release TRM limits for 60 minutes or longer.controlling document) limits for MODE: All 60 minutes or longer.MODE: All NEI Ex. NEI Example EAL Wording SSES SSES EAL Wording Difference/Deviation Justification EAL # EAL #Reading on ANY effluent Gaseous or liquid effluent > Table R-1 Example EALs #1 and #2 have been combined into a single radiation monitor greater than 2 column "UE" for > 60 min. EAL to simplify presentation. times the (site-specific effluent (Notes 1, 2, 3) The NEI phrase "...effluent radiation monitor greater than 2 release controlling document) times the (site-specific effluent release controlling limits for 60 minutes or longer: document)" and "effluent radiation monitor greater than 2 (site-specific monitor list and times the alarm setpoint established by a current radioactivity threshold values corresponding discharge permit " have been replaced with "... Gaseous or to 2 times the controlling RU1.1 liquid effluent > Table R-1 column "UE" document limits) UE thresholds for all SSES continuously monitored gaseous 2 Reading on ANY effluent and liquid release pathways are listed in Table R-1 to radiation monitor greater than 2 consolidate the information in a single location and, thereby, times the alarm setpoint simplify identification of the thresholds by the EAL user. The established by a current values shown in Table R-1 column "UE", consistent with the radioactivity discharge permit for NEI bases, represent two times the TRM release limits for 60 minutes or longer, both liquid and gaseous release.13 of 110 SSES EAL Comparison Matrix NEI Ex. NEI Example EAL Wording EAL # SSES EAL Wording Difference/Deviation Justification 3 Sample analysis for a gaseous or RU1.2 Sample analysis for a gaseous or The SSES TRM is the site-specific effluent release liquid release indicates a liquid release indicates a concentration controlling document.concentration or release rate or release rate > 2 x TRM limits for >greater than 2 times the (site- 60 min.specific effluent release (Notes 1, 2)controlling document) limits for 60 minutes or longer.Notes The Emergency Director N/A Note 1: The ED/RM should declare The classification timeliness note has been standardized should declare the Unusual the event promptly upon across the SSES EAL scheme by referencing the "time limit" Event promptly upon determining that time limit specified within the EAL wording.determining that 60 minutes has been exceeded, or will has been exceeded, or will likely be exceeded.likely be exceeded. Note 2: If an ongoing release is* If an ongoing release is detected and the release The classification timeliness note has been standardized detected and the release start time is unknown, across the SSES EAL scheme by referencing the "time limit" start time is unknown, assume that the release specified within the EAL wording.assume that the release duration has exceeded the duration has exceeded 60 specified time limit.minutes. Note 3: If the effluent flow past an* If the effluent flow past an effluent monitor is known to None effluent monitor is known to have stopped, indicating that have stopped due to actions the release path is isolated, to isolate the release path, the effluent monitor reading then the effluent monitor is no longer valid for reading is no longer valid for classification purposes.classification purposes.14 of 110, SSES EAL Comparison Matrix Table R-1 Effluent Monitor Classification Thresholds (Note 4)Release Point Monitor GE SAE Alert UEPlant Vent 01.9E+09 1.9E+08 1.9E+07 4.OE+06 0 Plan Vent0C630 (noble gas) 00677 pCi/min pCi/min pCi/min pCi/min (site total) (site total) (site total) (site total)LRW RR-06433 ............- 2 x hi alarm". 1(2) RHRSW A/B RR-D12-1R606 ............- 2 x hi alarm , 1(2) SW/SDHR RR-D12-1 R604 ............- 2 x hi alarm 15 of 110 SSES EAL Comparison Matrix NEI IC# NEI IC Wording and Mode SSES SSES IC Wording and Mode Difference/Deviation Justification Applicability IC#(s) Applicability AU2 UNPLANNED loss of water level RU2 Unplanned loss of water level above None above irradiated fuel. irradiated fuel MODE: All MODE: All NEI Ex. SSES EAL # NEI Example EAL Wording EAL # SSES EAL Wording Difference/Deviation Justification

a. UNLANND waer lvel U2.1 UNPLANNED water level drop in the Si a. UNPLANNED water level RU2.1 REFUELING PATHWAY as indicated Site-specific level indications and area radiation monitors are drop in the REFUELING by any of the following on EITHER listed in bullet format for clarification.

PATHWAY as indicated by unit: Added "... on EITHER unit..." to clarify application to a multi-ANY of the following:

Fuel Pool Water Low Level unit site.(site-specific level alarm indications).

0 Skimmer Surge Tank Low Level AND alarm b. UNPLANNED rise in area 0 Visual observation of a water radiation levels as indicated level drop below a fuel pool by ANY of the following skimmer surge tank inlet radiation monitors. a Observation of water draining (site-specific list of area down the outside wall of primary radiation monitors) containment AND UNPLANNED rise in area radiation levels as indicated by any of the following radiation monitors:* Channel 14 Spent Fuel Pool Area Criticality Monitor" Channel 15 Refueling Floor Area* Channel 42 Refueling Floor Area 16 of 110 SSES EAL Comparison Matrix SSES NEI IC# NEI IC Wording IC#(s) SSES IC Wording Difference/Deviation Justification AA1 Release of gaseous or liquid RA1 Release of gaseous or liquid SSES utilizes child thyroid for dose assessments and radioactivity resulting in offsite radioactivity resulting in offsite dose protective action recommendations. dose greater than 10 mrem TEDE greater than 10 mrem TEDE or 50 or 50 mrem thyroid CDE. mrem child thyroid CDE MODE: All MODE: All NEI Ex. SSES EA E NEI Example EAL Wording SL SSES EAL Wording Difference/Deviation Justification EAL # EAL #1 Reading on ANY of the following RAI.1 Gaseous effluent > Table R-1 column The SSES radiation monitors that detect gaseous radioactivity radiation monitors greater than "Alert" for > 15 min. (Notes 1, 2, 3, 4) effluent release to the environment are listed in Table R-1.the reading shown for 15 minutes UE, Alert, SAE and GE thresholds for all SSES continuously or longer: monitored gaseous release pathways are listed in Table R-1 (site-specific monitor list and to consolidate the information in a single location and, threshold values) thereby, simplify identification of the thresholds by the EAL-user.2 Dose assessment using actual RA1.2 Dose assessment using actual The EMERGENCY PLAN BOUNDARY is the site-specific meteorology indicates doses meteorology indicates doses > 10 receptor point.greater than 10 mrem TEDE or mrem TEDE or 50 mrem child thyroid 50 mrem thyroid CDE at or CDE at or beyond the EMERGENCY beyond (site-specific dose PLAN BOUNDARY (Notes 3, 4)receptor point).3 Analysis of a liquid effluent RA1.3 Analysis of a liquid effluent sample The EMERGENCY PLAN BOUNDARY is the site-specific sample indicates a concentration indicates a concentration or release receptor point.or release rate that would result rate that would result in doses > 10 in doses greater than 10 mrem mrem TEDE or 50 mrem child thyroid The alert classification for liquid releases must be declared TEDE or 50 mrem thyroid CDE at CDE at or beyond the EMERGENCY based on RA1.3 by sample analysis per the TRM or or beyond (site-specific dose PLAN BOUNDARY for 60 min. of applicable off normal procedure. receptor point) for one hour of exposure (Notes 1, 2)exposure.17 of 110 SSES EAL Comparison Matrix 4 Field survey results indicate EITHER of the following at or beyond (site-specific dose receptor point): " Closed window dose rates greater than 10 mR/hr expected to continue for 60 minutes or longer." Analyses of field survey samples indicate thyroid CDE greater than 50 mrem for one hour of inhalation. RA1.4 Field survey results indicate EITHER of the following at or beyond the EMERGENCY PLAN BOUNDARY: " Closed window dose rates > 10 mR/hr expected to continue for> 60 min." Analyses of field survey samples indicate child thyroid CDE > 50 mrem for 60 min. of inhalation.(Notes 1, 2)The EMERGENCY PLAN BOUNDARY is the site-specific receptor point.4 1- I.Notes" The Emergency Director should declare the Alert promptly upon determining that the applicable time has been exceeded, or will likely be exceeded." If an ongoing release. is detected and the release start time is unknown, assume that the release duration has exceeded 15 minutes." 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." The pre-calculated effluent monitor values presented in EAL #1 should be used for emergency classification assessments until the results N/A Note 1: The ED/RM 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 RAi.1, RSI.1 and RGI.1 should be used for emergency classification assessments until the The classification timeliness note has been standardized across the SSES EAL scheme by referencing the "time limit" specified within the EAL wording.The classification timeliness note has been standardized across the SSES EAL scheme by referencing the "time limit" specified within the EAL wording.None Incorporated site-specific EAL numbers associated with generic EAL#1.I 18 of 110 SSES EAL Comparison Matrix from a dose assessment results from a dose using actual meteorology are assessment using actual available. meteorology are available. 19 of 110 SSES EAL Comparison Matrix SSES NEI IC# NEI IC Wording IC#(s) SSES IC Wording Difference/Deviation Justification AA2 Significant lowering of water RA2 Significant lowering of water level None level above, or damage to, above, or damage to, irradiated fuel irradiated fuel. MODE: All MODE: All NEI Ex. NEI Example EAL Wording SSES SSES EAL Wording Difference/Deviation Justification EAL # EAL #1 Uncovery of irradiated fuel in the RA2.1 Uncovery of irradiated fuel in the None REFUELING PATHWAY. REFUELING PATHWAY .2 Damage to irradiated fuel RA2.2 Damage to irradiated fuel resulting in a Deleted the NEI phrase "from the fuel" because it is redundant resulting in a release of release of radioactivity to the preceding phrase "irradiated fuel." radioactivity from the fuel as AND Site-specific list of radiation monitors are listed in bullet format indicated by ANY of the following for clarification. Listed in bullet format for clarification. radiation monitors: Any of the following radiation monitor indications: (site-specific listing of radiation monitors, and the associated ° Refuel Floor High Exhaust (>readings, setpoints and/or 18 mR/hr)alarms) a Refuel Floor Wall Exhaust (>21 mR/hr)a Channel 14 Spent Fuel Pool Area Criticality Monitor (> 100 mR/hr)* Channel 15 Refueling Floor Area (> 80 mR/hr)* Channel 42 Refueling Floor Area (> 80 mR/hr)0 Channel 47 (Ul) / 44 (U2)Spent Fuel Pool Area Criticality Monitor (> 100 20 of 110 SSES EAL Comparison Matrix mR/hr)* Channel 49 Refueling Floor High Range Monitor (on scale)3 Lowering of spent fuel pool level RA2.3 Lowering of spent fuel pool level to Post-Fukushima order EA-12-051 required the installation of to (site-specific Level 2 value). < 10 ft. above the top of the spent fuel reliable SFP level indication capable of identifying normal level[See Developer Notes] racks (Level 1), SFP level 10 ft. above the top of the fuel racks (Level 2 or el. 804'-4") and SFP level at the top of the fuel racks (Level 3 or el.794'-1 0" which for SSES is 6 inches above the top of the fuel racks). Each spent fuel pool is equipped with primary and backup guided wave radar probes to measure pool level. The range is continuous from the high pool level elevation (el. 817'-7") to the top of the spent fuel racks (0 inches or el. 794'-4")21 of 110 SSES EAL Comparison Matrix SSES NEI IC# NEI IC Wording IC#(s) SSES IC Wording Difference/Deviation Justification AA3 Radiation levels that impede RA3 Radiation levels that IMPEDE access RA3.2 mode applicability is Mode 3 only because the Mode access to equipment necessary to equipment necessary for normal dependent safe operation and shutdown areas are only for normal plant operations, plant operations, cooldown or applicable in Mode 3 per Attachment 3 of the bases cooldown or shutdown shutdown. document.MODE: All MODE: All (except RA3.2 Mode 3 only)NEI Ex. NEI Example EAL Wording SSES SSES EAL Wording Difference/Deviation Justification EAL # EAL #1 Dose rate greater than 15 mRlhr RA3 Dose rates > 15 mR/hr in any of the The Radwaste Control Room is another site-specific area in ANY of the following areas: following areas: Room e Main Control RoomreurncotnoscupcyaSES" Control Room

Radwaste Control Room Central Alarm Station (CAS) and Secondary Alarm Station" Central Alarm Station
Both the Central Alarm Station (SAS) are both included in this EAL because either security (CAS) and Secondary Alarm station can effectively permit access to areas required to* (other site-specific Station (SAS) assure safe plant operations.

areas/rooms) 2 An UNPLANNED event results RA3.2 An UNPLANNED event results in The site-specific list of plant rooms or areas with entry-related in radiation levels that prohibit or radiation levels that prohibit or mode applicability are listed in Table R-2 for clarification. impede access to any of the IMPEDE access to any Table R-2 following plant rooms or areas: rooms or areas (Note 5)(site-specific list of plant rooms or areas with entry-related mode applicability identified) Note If the equipment in the listed N/A Note 5 If the equipment in the listed None room or area was already area was already inoperable inoperable or out-of-service or out-of-service before the before the event occurred, then event occurred, then no no emergency classification is emergency classification is warranted. warranted. 22 of 110 SSES EAL Comparison Matrix Table R-2 Safe Operation & Shutdown Areas Elevation Unit I Area(s) ** Unit 2 Area(s) ** Mode(s)670' RB 27 RB 32 3,4,5 683' RB 27, 28, 29 RB 32, 33, 34 3,4,5 703' RB 28, 29 RB 33, 34 3,4,5 719' RB 25, 29 RB 30, 34 3,4,5 749' RB 25, 29 RB 32, 33 3,4,5** See Chart I for location of plant areas Chart 1- Plant Area Key Plan m LOW LEVEL iRADWASTE ACID AND CHLORINE BUILDING WATER TREATMENT 53 52 51 50 PUMPHOUSE UNIT #2 TURBINE BLDG.El UNIT#1 TURBINE BLDG.E SPRAY POND VALVE VAULT ESSW PUMPHOUSE RADWASTE 38 37 40 39 42 41 TURB.16 15 14 1 13 4 13 2 1 E CK 20 1 19 18 I 17 8 17 6 5 L. I I I 1 1. ~-I -I -24 23 22 1 21 12 1 11 10 9 (36 44 35 43 COND & REF STORAGE DIESEL GENERATOR , -.j-RIVER INTAKE 57 STRUCTURE REACTOR REACTOR 1 E DIESEL GENERATOR 23 of 110 SSES EAL Comparison Matrix SSES NEI IC Wording IC#(s) SSES IC Wording Difference/Deviation Justification Release of gaseous radioactivity RS1 Release of gaseous radioactivity SSES utilizes child thyroid for dose assessments and resulting in offsite dose greater resulting in offsite dose greater than protective action recommendations. than 100 mrem TEDE or 500 100 mrem TEDE or 500 mrem child mrem thyroid CDE thyroid CDE MODE: All MODE: All NEI Ex. NEI Example EAL Wording EA SSES EAL Wording Difference/Deviation Justification EAL # EAL #1 Reading on ANY of the following RSI.1 Gaseous effluent > Table R-1 column The SSES radiation monitors that detect radioactivity effluent radiation monitors greater than "SAE" for > 15 min. release to the environment are listed in Table R-1. UE, Alert, the reading shown for 15 (Notes 1, 2, 3, 4) SAE and GE thresholds for all SSES continuously monitored minutes or longer: gaseous release pathways are listed in Table R-1 to (site-specific monitor list and consolidate the information in a single location and, thereby, threshold values) simplify identification of the thresholds by the EAL-user.2 Dose assessment using actual RS1.2 Dose assessment using actual The EMERGENCY PLAN BOUNDARY is the site-specific meteorology indicates doses meteorology indicates doses > 100 receptor point.greater than 100 mrem TEDE or mrem TEDE or 500 mrem child thyroid 500 mrem thyroid CDE at or CDE at or beyond the EMERGENCY beyond (site-specific dose PLAN BOUNDARY (Notes 3, 4)receptor point)3 Field survey results indicate RS1.3 Field survey results indicate EITHER The EMERGENCY PLAN BOUNDARY is the site-specific EITHER of the following at or of the following at or beyond the receptor point.beyond (site-specific dose EMERGENCY PLAN BOUNDARY: receptor point): .Closed window dose rates >* Closed window dose rates 100 mR/hr expected to continue greater than 100 mR/hr for > 60 min.expected to continue for 60 minutes or longer.

Analyses of field survey" Analyses of field survey samples indicate child thyroid samples indicate thyroid CDE > 500 mrem for 60 min. of CDE greater than 500 inhalation.

24 of 110 SSES EAL Comparison Matrix mrem for one hour of inhalation.(Notes 1, 2)Notes I *The Emergency Director should declare the Site Area Emergency promptly upon determining that the applicable time has been exceeded, or will likely be exceeded." If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 15 minutes." 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." The pre-calculated effluent monitor values presented in EAL #1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available. Note 1: The ED/RM 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 RAI.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. The classification timeliness note has been standardized across the SSES EAL scheme by referencing the "time limit" specified within the EAL wording.The classification timeliness note has been standardized across the SSES EAL scheme by referencing the "time limit" specified within the EAL wording.None Incorporated site-specific EAL numbers associated with generic EAL#1.25 of 110 SSES EAL Comparison Matrix~SSES NEI IC# NEI IC Wording IC#(S) SSES IC Wording Difference/Deviation Justification AS2 Spent fuel pool level at (site- RS2 Spent fuel pool level at the top of the Top of the fuel racks is the site specific Level 3.specific Level 3 description) fuel racks MODE: All NEI Ex.SSES NEI Ex. NEI Example EAL Wording EA SSES EAL Wording Difference/Deviation Justification EAL # EAL #1 Lowering of spent fuel pool level RS2.1 Lowering of spent fuel pool level to Post-Fukushima order EA-12-051 required the installation of to (site-specific Level 3 value) < 0.5 ft. above the top of the spent reliable SFP level indication capable of identifying normal level fuel racks (Level 1), SFP level 10 ft. above the top of the fuel racks (Level 2 or el. 804'-4") and SFP level at the top of the fuel racks (Level 3 or el.794'-1 0" which for SSES is 6 inches above the top of the fuel racks). Each spent fuel pool is equipped with primary and backup guided wave radar probes to measure pool level. The range is continuous from the high pool level elevation (el. 817'-7") to the top of the spent fuel racks (0 inches or el. 794'-4")26 of 110 SSES EAL Comparison Matrix SSES NEI IC# NEI IC Wording IC#(s) SSES IC Wording Difference/Deviation Justification AG1 Release of gaseous radioactivity RG1 Release of gaseous radioactivity SSES utilizes child thyroid for dose assessments and protective resulting in offsite dose greater resulting in offsite dose greater action recommendations. than 1,000 mrem TEDE or than 1,000 mrem TEDE or 5,000 5,000 mrem thyroid CDE. mrem child thyroid CDE MODE: All MODE: All NEI Ex. NEI Example EAL Wording SSES SSES EAL Wording Difference/Deviation Justification EAL # EAL #1 Reading on ANY of the following RG1.1 Gaseous effluent > Table R-1 The SSES radiation monitors that detect radioactivity effluent radiation monitors greater than column "GE" for > 15 min. release to the environment are listed in Table R-1. UE, Alert, SAE the reading shown for 15 (Notes 1, 2, 3, 4) and GE thresholds for all SSES continuously monitored gaseous minutes or longer: release pathways are listed in Table R-1 to consolidate the (site-specific monitor list and information in a single location and, thereby, simplify identification of threshold values) the thresholds by the EAL-user.2 Dose assessment using actual RG1.2 Dose assessment using actual The EMERGENCY PLAN BOUNDARY is the site-specific receptor meteorology indicates doses meteorology indicates doses > point.greater than 1,000 mrem TEDE 1000 mrem TEDE or or 5,000 mrem thyroid CDE at 5000 mrem child thyroid CDE at or or beyond (site-specific dose beyond the EMERGENCY PLAN receptor point). BOUNDARY (Notes 3, 4)3 Field survey results indicate RG1.3 Field survey results indicate The EMERGENCY PLAN BOUNDARY is the site-specific receptor EITHER of the following at or EITHER of the following at or point.beyond (site-specific dose beyond the EMERGENCY PLAN receptor point): BOUNDARY: " Closed window dose rates

Closed window dose rates >greater than 1,000 mR/hr 1000 mR/hr expected to expected to continue for 60 continue for > 60 min.minutes or longer." Analyses of field survey
Analyses of field survey samples indicate thyroid CDE samples indicate child thyroid greater than 5,000 mrem for CDE > 5000 mrem for 60 min.27 of 110 SSES EAL Comparison Matrix one hour of inhalation, of inhalation.

______ ____________________ ~(Notes 1, 2) _____________________ Notes" The Emergency Director should declare the Site Area Emergency promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.* If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 15 minutes." 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." The pre-calculated effluent monitor values presented in EAL #1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available. Note 1: The ED/RM 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 5 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. The classification timeliness note has been standardized across the SSES EAL scheme by referencing the "time limit" specified within the EAL wording.The classification timeliness note has been standardized across the SSES EAL scheme by referencing the "time limit" specified within the EAL wording.None Incorporated site-specific EAL numbers associated with generic EAL#1.28 of 110 SSES EAL Comparison Matrix SSES NEI IC# NEI IC Wording IC#(s) SSES IC Wording Difference/Deviation Justification AG2 Spent fuel pool level cannot be RG2 Spent fuel pool level cannot be Top of the fuel racks is the site specific Level 3.restored to at least (site-specific restored to at least the top of the fuel Level 3 description) for 60 racks for 60 minutes or longer minutes or longer MODE: All MODE: All NEI Ex. SSES EAL # NEI Example EAL Wording EAL # SSES EAL Wording Difference/Deviation Justification Spent fuel pool level cannot be RG2.1 Spent fuel pool level CANNOT BE Post-Fukushima order EA-12-051 required the installation of restored to at least (site-specific RESTORED to at least 0.5 ft. above reliable SFP level indication capable of identifying normal level Level 3 value) for 60 minutes or the top of the spent fuel racks for (Level 1), SFP level 10 ft. above the top of the fuel racks longer > 60 min. (Note 1) (Level 2 or el. 804'-4") and SFP level at the top of the fuel racks (Level 3 or el.794'-10" which for SSES is 6 inches above the top of the fuel racks). Each spent fuel pool is equipped with primary and backup guided wave radar probes to measure pool level. The range is continuous from the high pool level elevation (el. 817'-7") to the top of the spent fuel racks (0 inches or el. 794'-4")Note The Emergency Director should Note 1: The ED/RM should declare The classification timeliness note has been standardized declare the General Emergency the event promptly upon across the SSES EAL scheme by referencing the "time limit" promptly upon determining that determining that time limit specified within the EAL wording.60 minutes has been exceeded, has been exceeded, or will or will likely be exceeded. likely be exceeded.29 of 110 SSES EAL Comparison Matrix Category C Cold Shutdown / Refueling System Malfunction 30 of 110 SSES EAL Comparison Matrix SSES NEI IC# NEI IC Wording IC#(s) SSES IC Wording Difference/Deviation Justification Cui UNPLANNED loss of (reactor CUi UNPLANNED loss of RPV None vessel/RCS [PWR] or RPV inventory for 15 minutes or[BWR]) inventory for 15 minutes longer or longer. MODE: 4 -Cold Shutdown, 5 -MODE: Cold Shutdown, Refueling Refueling NEI Ex. SSES EAL # NEI Example EAL Wording EAL SSES EAL Wording Difference/Deviation Justification 1 UNPLANNED loss of reactor CUI.1 UNPLANNEDloss of reactor None coolant results in (reactor coolant results in RPV level vessel/RCS [PWR] or RPV less than a required lower limit[BWR]) level less than a for > 15 min. (Note 1)required lower limit for 15 minutes or longer.2 a. (Reactor vessel/RCS [PWR] CU1.2 RPV level cannot be monitored Site-specific applicable sumps and tanks are listed in Table C-1 to or RPV [BWR]) level cannot AND improve the readability of the EAL.be monitored. The phrase "due to a loss of RPV inventory" has been added to the AND UNPLANNED increase in any SSES EAL for clarification. This wording implements the intent of the Table C-1 sump or tank levels NEI EAL basis which states" "Sump and/or tank level changes must b. UNPLANNED increase in due to a loss of RPV inventory be evaluated against other potential sources of water flow to ensure (site-specific sump and/or they are indicative of leakage from the RCS." tank) levels. Although "Visual Observation" in Table C-1 is neither a sump nor tank, it is included in order to implement the intent of the NEI basis which states: "... operators may determine that an inventory loss is occurring by observing changes..." 31 of 110 SSES EAL Comparison Matrix Note The Emergency Director should declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.N/A Note 1: The ED/RM should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.The classification timeliness note has been standardized across the SSES EAL scheme by referencing the "time limit" specified within the EAL wording.Table C-1 Sumps & Tanks* Drywell equipment drain tank* Drywell sumps* Reactor Building sump* LRW collection tanks* Main condenser hotwell* Suppression pool* Visual observation 32 of 110 SSES EAL Comparison Matrix SSES NEI IC# NEI IC Wording IC#(s) SSES IC Wording Difference/Deviation Justification CU2 Loss of all but one AC power CU2 Loss of all but one AC power "essential" is the site-specific designation for the SSES emergency source to emergency buses for source to essential buses for 15 AC buses.15 minutes or longer. minutes or longer.MODE: Cold Shutdown, MODE: 4 -Cold Shutdown, 5 -Refueling, Defueled Refueling, D -Defueled NEI Ex. SSES EA # NEI Example EAL Wording EA SSES EAL Wording Difference/Deviation Justification EAL # EAL #a. AC power capability to (site- CU2.1 AC power capability to all 4.16 ESS buses are the site-specific emergency buses.specific emergency buses) is kV ESS buses on EITHER unit Added "...on EITHER unit..." to clarify application to a multi-unit site.reduced to a single power reduced to a single power source for 15 minutes or source for > 15 min. (Note 1)longer. AND AND Any additional single power b. Any additional single power source failure will result in loss of source failure will result in all AC power to SAFETY(loss of all AC power to SYSTEMS SAFETY SYSTEMS.Note The Emergency Director should N/A Note 1: The ED/RM should The classification timeliness note has been standardized across the declare the Unusual Event declare the event SSES EAL scheme by referencing the "time limit" specified within promptly upon determining that promptly upon the EAL wording.15 minutes has been exceeded, determining that time or will likely be exceeded. limit has been exceeded, or will likely be exceeded.33 of 110 SSES EAL Comparison Matrix N/A N/A N/A Note 10: Credit may be taken for Added site specific note 10 to clarify that if a CTG is not already one of the four CTGs as aligned at the time of the AC power loss credit cannot be taken for an onsite AC power that CTG source because the time required to align a CTG to an source only if the CTG essential bus exceeds 15 min.is already aligned and capable of powering an essential bus within 15 mi.34 of 110 SSES EAL Comparison Matrix SSES NEI IC# NEI IC Wording IC#(s) SSES IC Wording Difference/Deviation Justification CU3 UNPLANNED increase in RCS CU4 UNPLANNED increase in RCS Reordered subcategory for loss of DC power to follow loss of AC temperature temperature power.MODE: Cold Shutdown, MODE: 4 -Cold Shutdown, 5 -Refueling Refueling NEI Ex. SSES EI E NEI Example EAL Wording ES SSES EAL Wording Difference/Deviation Justification EAL i# EAL #1 UNPLANNED increase in RCS CU4.1 UNPLANNED increase in RCS 200'F is the site-specific Tech. Spec. cold shutdown temperature temperature to greater than (site- temperature to > 200°F due to limit.specific Technical Specification loss of decay heat removal Added the phrase "due to loss of decay heat removal capability" to cold shutdown temperature limit) capability clarify that the increase in temperature is related to such capability as specified in the generic bases.2 Loss of ALL RCS temperature CU4.2 Loss of all RCS temperature and None and (reactor vessel/RCS [PWR] RPV level indication for > 15 min.or RPV [BWR]) level indication (Note 1)for 15 minutes or longer.Note The Emergency Director should N/A Note 1: The ED/RM should The classification timeliness note has been standardized across the declare the Unusual Event declare the event SSES EAL scheme by referencing the "time limit" specified within promptly upon determining that promptly upon the EAL wording.15 minutes has been exceeded, determining that time or will likely be exceeded limit has been exceeded, or will likely be exceeded.35 of 110 SSES EAL Comparison Matrix SSES NEI IC# NEI IC Wording IC#(s) SSES IC Wording Difference/Deviation Justification CU4 Loss of Vital DC power for 15 CU3 Loss of vital DC power for 15 Reordered subcategory for loss of DC power to follow loss of AC minutes or longer. minutes or longer, power.MODE: Cold Shutdown, MODE: 5 -Cold Shutdown, 6 -Refueling Refueling EAL Ex NEI Example EAL Wording SSES SSES EAL Wording Difference/Deviation Justification NEI Ex EAL # SSEL odn 1 Indicated voltage is less than CU3.1 < 105 VDC bus voltage < 105 VDC is the site-specific minimum Vital DC bus design (site-specific bus voltage value) indications on Technical voltage.on required Vital DC buses for Specification required 125 VDC Clarified that the plant design of indicated voltage for the vital 125 15 minutes or longer, buses on the affected unit for > VDC main distribution buses is local only. Local voltage indication 15 min. (Note 1) is available for each bus based on dispatching a field operator in accordance with Control Room alarm response procedure AR-1(2)06-001 (A12, B12,C12,D12). Field observation of indicated voltage constitutes the point in time when availability of indications to plant operators that an emergency action level has been, or may be, exceeded.Note The Emergency Director should N/A Note 1:The ED/RM should The classification timeliness note has been standardized across declare the Unusual Event declare the event promptly upon the SSES EAL scheme by referencing the "time limit" specified promptly upon determining that determining that time limit has within the EAL wording.15 minutes has been exceeded, been exceeded, or will likely be or will likely be exceeded. exceeded.36 of 110 SSES EAL Comparison Matrix NEI IC# NEI IC Wording CU5 Loss of all onsite or offsite communications capabilities. MODE: Cold Shutdown, Refueling, Defueled SSES IC Wording Difference/Deviation Justification Loss of all onsite or offsite None communications capabilities. MODE: 4- Cold Shutdown, 5 -Refueling, D -Defueled NEI Ex. NEI Example EAL Wording EA SSES EAL Wording Difference/Deviation Justification EAL # EAL #1 Loss of ALL of the following CU5.1 Loss of all Table C-4 onsite Example EALs #1, 2 and 3 have been combined into a single EAL onsite communication methods: communication methods for simplification of presentation.(site specific list of OR Table C-4 provides a site-specific list of onsite, ORO and NRC communications methods) Loss of all Table C-40RO communications methods.2 Loss of ALL of the following ORO communication methods communications methods: OR (site specific list of Loss of all Table C-4 NRC communications methods) communication methods 3 Loss of ALL of the following NRC communications methods: (site specific list of communications methods)37 of 110 SSES EAL Comparison Matrix Table C-4 Communication Methods System Onsite ORO NRC UHF Radio X Plant PA System X Dedicated Conference Lines X Commercial Telephone Systems X X X Cellular Telephone X X FTS-2001 (ENS) X X 38 of 110 SSES EAL Comparison Matrix SSES NEI IC# NEI IC Wording IC#(s) SSES IC Wording Difference/Deviation Justification CA1 Loss of (reactor vessel/RCS CA1 Loss of RPV inventory None[PWR] or RPV [BWR]) inventory MODE: 4 -Cold Shutdown, 5 -MODE: Cold Shutdown, Refueling Refueling NEI Ex. NEI Example EAL Wording SSES SSES EAL Wording Difference/Deviation Justification EAL # EAL #Loss of (reactor vessel/RCS CA1.1 Loss of RPV inventory as -38 in is the site-specific level corresponding to the Level 2 trip[PWR] or RPV [BWR]) inventory indicated by RPV level < -38 in. setpoint.as indicated by level less than (Level 2)(site-specific level).2 a. (Reactor vessel/RCS [PWR] CA1.2 RPV level cannot be monitored Site-specific applicable sumps and tanks are listed in Table C-1 to or RPV [BWR]) level cannot for > 15 min. (Note 1) improve the readability of the EAL.be monitored for 15 minutes AND Although "Visual Observation" in Table C-1 is neither a sump nor orlonger UPtank, it is included in order to implement the intent of the NEI basis AND UNPLANNED increase in any which states: "...operators may determine that an inventory loss is Table C-i sump or tank levels b. UNPLANNED increase in due to a loss of RPV inventory occurring by.observing changes..." (site-specific sump and/or tank) levels due to a loss of (reactor vessel/RCS [PWR]or RPV [BWR]) inventory. Note The Emergency Director should N/A Note 1:The ED/RM should The classification timeliness note has been standardized across the declare the Alert promptly upon declare the event promptly upon SSES EAL scheme by referencing the "time limit" specified within the determining that 15 minutes has determining that time limit has EAL wording.been exceeded, or will likely be been exceeded, or will likely be exceeded exceeded.39 of 110 SSES EAL Comparison Matrix SSES NEI IC# NEI IC Wording IC#(s) SSES IC Wording Difference/Deviation Justification CA2 Loss of all offsite and all onsite CA2 Loss of all offsite and all onsite "essential" is the site-specific designation for the SSES emergency AC power to emergency buses AC power to essential buses for AC buses.for 15 minutes or longer 15 minutes or longer.MODE: Cold Shutdown, MODE: Cold Shutdown, Refueling, Defueled Refueling, Defueled NEI Ex. NEI Example EAL Wording SSES SSES EAL Wording Difference/Deviation Justification EAL # EAL #1 Loss of ALL offsite and ALL CA2.1 Loss of all offsite and all onsite ESS buses are the site-specific emergency buses.onsite AC Power to (site- AC power capability to all 4.16 Added the words "on EITHER unit" to clarify intent for a multi-unit specific emergency buses) for kV ESS buses on EITHER unit site.15 minutes or longer. for > 15 min. (Note 1)Note The Emergency Director should N/A Note 1: The ED/RM should The classification timeliness note has been standardized across the declare the Unusual Event declare the event SSES EAL scheme by referencing the "time limit" specified within promptly upon determining that promptly upon the EAL wording.15 minutes has been exceeded, determining that time or will likely be exceeded. limit has been exceeded, or will likely be exceeded.40 of 110 SSES EAL Comparison Matrix SSES NEI IC# NEI IC Wording IC#(s) SSES IC Wording Difference/Deviation Justification CA3 Inability to maintain the plant in CA4 Inability to maintain the plant in Reordered subcategory for loss of DC power to follow loss of AC cold shutdown. cold shutdown. power.MODE: Cold Shutdown, MODE: 4 -Cold Shutdown, 5 -Refueling Refueling NEI Ex. SSES EAL # NEI Example EAL Wording EAL # SSES EAL Wording Difference/Deviation Justification 1 UNPLANNED increase in RCS CA4.1 UNPLANNED increase in RCS Example EALs #1 and #2 have been combined in one SSES EAL for temperature to greater than temperature to > 200°F for > simplification.(site-specific Technical Table C-3 duration 200'F is the site-specific Tech. Spec. cold shutdown temperature Specification cold shutdown (Note 1) limit.temperature limit) for greater than the duration specified in the OR Table C-3 is the site-specific implementation of the generic RCS following table. UNPLANNED RPV pressure Heat-up Duration Threshold table.increase > 10 psig due to loss 10 psig is the site-specific pressure increase readable by Control 2 UNPLANNED RCS pressure of decay heat removal Room indications. increase greater than (site- capability specific pressure reading). (This Added "due to loss of decay heat removal capability" consistent with EAL does not apply during the generic bases wording.water-solid plant conditions. [PWR])Note The Emergency Director should N/A Note 1: The ED/RM should The classification timeliness note has been standardized across the declare the Unusual Event declare the event SSES EAL scheme by referencing the "time limit" specified within the promptly upon determining that promptly upon EAL wording.15 minutes has been exceeded, determining that time or will likely be exceeded. limit has been exceeded, or will likely be exceeded.41 of 110 SSES EAL Comparison Matrix Table: RCS Heat-up Duration Thresholds RCS Status Containment Closure Status Heat-up Duration Intact (but not at reduced Not applicable 60 minutes*inventory [PWR])Not intact (or at reduced Established 20 minutes*inventory [PWR]) Not Established 0 minutes* If an RCS heat removal system is in operation within this time frame and RCS temperature is being reduced, the EAL is not applicable. Table C-3 RCS Heat-up Duration Thresholds RCS Status Containment Closure Status Heat-up Duration INTACT N/A 60 min.*established 20 min.*Not INTACT not established 0 min.* If an RCS heat removal system is in operation within this time frame and RCS temperature is being reduced, the EAL is not applicable. 42 of 110 SSES EAL Comparison Matrix SSES NEI IC# NEI IC Wording IC#(s) SSES IC Wording Difference/Deviation Justification CA6 Hazardous event affecting a CA6 Hazardous event affecting a None SAFETY SYSTEM needed for SAFETY SYSTEM needed for the current operating mode. the current operating mode.MODE: Cold Shutdown, MODE: 4 -Cold Shutdown, 5 -Refueling Refueling NEI Ex. SSES EAL # NEI Example EAL Wording EAL # SSES EAL Wording Difference/Deviation Justification

a. The occurrence of ANY of the CA6.1 The occurrence of any Table C- The hazardous events have been listed in Table C-5 to improve the following hazardous events: 5 hazardous event readability of the SSES EAL." Seismic event AND (earthquake)

EITHER: " Internal or external flooding event Event damage has caused* High winds or tornado indications of degraded strike performance in at least one" FIRE train of a SAFETY SYSTEM* EXPLOSION needed for the current" (site-specific hazards) operating mode" Other events with similar OR hazard characteristics as determined by the Shift The event has caused Manager VISIBLE DAMAGE to a AND SAFETY SYSTEM b. EITHER of the following: component or structure 1. Event damage has caused needed for the current indications of degraded operating mode performance in at least one train of a SAFETY SYSTEM needed for the current operating mode.OR 43 of 110 SSES EAL Comparison Matrix 2. The event has caused VISIBLE DAMAGE to a SAFETY SYSTEM component or structure needed for the current_________ operating mode._________________________________________________ Table C-5 Hazardous Events" Seismic event (earthquake)

Internal or external FLOODING event" High winds or tornado strike" FIRE* EXPLOSION" Other events with similar hazard characteristics as determined by the Shift Manager 44 of 110 SSES EAL Comparison Matrix SSES NEI IC# NEI IC Wording IC#(s) SSES IC Wording Difference/Deviation Justification CS1 Loss of (reactor vessel/RCS CS1 Loss of RPV inventory affecting None[PWR] or RPV [BWR]) inventory core decay heat removal affecting core decay heat capability removal capability.

MODE: 4 -Cold Shutdown, 5 -MODE: Cold Shutdown, Refueling Refueling NEI Ex. SSES EAL E NEI Example EAL Wording EAL # SSES EAL Wording Difference/Deviation Justification 1 a. CONTAINMENT CLOSURE CS1.1 CONTAINMENT CLOSURE not -129 in. is the RPV water level that corresponds to the Level 1 trip not established, established setpoint.AND AND b. (Reactor vessel/RCS [PWR] RPV level < -129 in. (Level 1)or RPV [BWR]) level less than (site-specific level).2 a. CONTAINMENT CLOSURE CS1.2 CONTAINMENT CLOSURE -161 in. is the RPV water level that corresponds to the top of active established, established fuel.AND AND b. (Reactor vessel/RCS [PWR] RPV level < -161 in. (TAF)or RPV [BWR]) level less than (site-specific level).3 a. (Reactor vessel/RCS [PWR] CS1.3 RPV level cannot be monitored Site-specific applicable sumps and tanks are listed in Table C-1 to or RPV [BWR]) level cannot for > 30 min. (Note 1) improve readability of the EAL.be monitored for 30 minutes AND Although "Visual Observation" in Table C-1 is neither a sump nor or longer. Utank, it is included in order to implement the intent of the NEI basis AND Table C-1 sump or tank level which states: "... operators may determine that an inventory loss is b. Core uncovery is indicated by due to a loss of RPV inventory occurring by observing changes..." ANY of the following: of sufficient magnitude to The phrase "due to a loss of RPV inventory" has been added to the 45 of 110 SSES EAL Comparison Matrix* (Site-specific radiation monitor) reading greater than (site-specific value)" Erratic source range monitor indication [PWR]" UNPLANNED increase in (site-specific sump and/or tank) levels of sufficient magnitude to indicate core uncovery* (Other site-specific indications) indicate core uncovery SSES EAL for clarification. This wording implements the intent of the NEI EAL basis which states "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." SSES does not have any installed or temporary area radiation monitors located in line of sight to the reactor cavity that can be used as an alternative indication of irradiate fuel uncovery in the RPV. No alternative site-specific level indications of core uncover exist at SSES.Because the SSES SRMs are located near the core mid-plane they are not a viable indicator of core uncover.Note The Emergency Director should N/A Note 1: The ED/RM should The classification timeliness note has been standardized across the declare the Site Area declare the event SSES EAL scheme by referencing the "time limit" specified within the Emergency promptly upon promptly upon EAL wording.determining that 30 minutes has determining that time been exceeded, or will likely be limit has been exceeded exceeded, or will likely be exceeded.46 of 110 SSES EAL Comparison Matrix SSES NEI IC# NEI IC Wording IC#(s) SSES IC Wording Difference/Deviation Justification CG1 Loss of (reactor vessel/RCS CG1 Loss of RPV inventory affecting None[PWR] or RPV [BWR]) inventory fuel clad integrity with affecting fuel clad integrity with Containment challenged containment challenged MODE: 4 -Cold Shutdown, 5 -MODE: Cold Shutdown, Refueling Refueling NEI Ex. NEI Example EAL Wording EAL#S SSES EAL Wording Difference/Deviation Justification

a. (Reactor vessel/RCS

[PWR] or CG1.1 RPV level < -161 in. (TAF) for > -161 in. is the RPV water level corresponding to the top of active RPV [BWR]) level less than 30 min. fuel.(site-specific level) for 30 minutes or longer. AND The Containment Challenge Table is Table C-2.AND Any Containment Challenge 6% hydrogen concentration in the presence of oxygen is the indication, Table C-2 minimum necessary to support a hydrogen deflagration.

b. ANY indication from the The Max Safe Operating Radiation Levels are the highest value of Containment Challenge Table these parameters at which neither: (1) equipment necessary for (see below), the safe shutdown of the plant will fail, nor (2) personnel access necessary for the safe shutdown of the plant will be precluded.

These are the site-specific secondary containment radiation monitor readings and are listed in EOP-1 04.The Max Safe Operating Radiation Levels are restricted to only those that can be read within the Control Room to support prompt classification. 2 a. (Reactor vessel/RCS [PWR] or CG1.2 RPV level cannot be monitored Site-specific applicable sumps and tanks are listed in Table C-1 to RPV [BWR]) level cannot be for > 30 min. (Note 1) improve the readability of the EAL.monitored for 30 minutes or AND The phrase "due to a loss of RPV inventory" has been added to the AND UNPLANNED increase in any SSES EAL for clarification. This wording implements the intent of Table C-1 sump or tank level due the NEI EAL basis which states" "Sump and/or tank level changes b. Core uncovery is indicated by to a loss of RPV inventory of must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the RCS." 47 of 110 SSES EAL Comparison Matrix ANY of the following: 0 (Site-specific radiation monitor) reading greater than (site-specific value)* Erratic source range monitor indication [PWR]* UNPLANNED increase in (site-specific sump and/or tank) levels of sufficient magnitude to indicate core uncovery* (Other site-specific indications) AND c. ANY indication from the Containment Challenge Table (see below).sufficient magnitude to indicate core uncovery AND Any Containment Challenge indication, Table C-2 Although "Visual Observation" in Table C-1 is neither a sump nor tank, it is included in order to implement the intent of the NEI basis which states: "...operators may determine that an inventory loss is occurring by observing changes..." The Containment Challenge Table is Table C-2.SSES does not have any installed or temporary area radiation monitors located in line of sight to the reactor cavity that can be used as an alternative indication of irradiate fuel uncovery in the RPV. No alternative site-specific level indications of core uncover exist at SSES.Because the SSES SRMs are located near the core mid-plane they are not a viable indicator of core uncover.4 4 4 4 Note The Emergency Director should declare the General Emergency promptly upon determining that 30 minutes has been exceeded, or will likely be exceeded.N/A N/A Note 1: The ED/RM 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.The classification timeliness note has been standardized across the SSES EAL scheme by referencing the "time limit" specified within the EAL wording.Note 6 implements the asterisked note associated with the generic Containment Challenge table.48 of 110 SSES EAL Comparison Matrix Containment Challenge Table N CONTAINMENT CLOSURE not established*

(Explosive mixture) exists inside containment
UNPLANNED increase in containment pressure* Secondary containment radiation monitor reading above (site-specific value) [BWR]* If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, then declaration of a General Emergency is not required.Table C-2 Containment Challenge Indications
CONTAINMENT CLOSURE not established (Note 6)* PC hydrogen concentration

> 6%* UNPLANNED rise in PC pressure* Exceeding one or more Secondary Containment Control Max Safe Radiation Levels (EO-000-1 04 Table 9) that can be read in the Control Room (Table C-6)Reference to EO-000-104 Table and Table C-6 have been added for clarification. This Containment Challenge Indication threshold is limited to those Max Safe Reactor Building Radiation Limits that can be remotely determined from within the control room to support prompt classification. Reference PPL letters to NRC: SUSQUEHANNA STEAM ELECTRIC STATION REPLY TO A NOTICE OF VIOLATION PLA-7212 dated 08/29/2014 SUSQUEHANNA STEAM ELECTRIC STATION NOTIFICATION OF COMMITMENT CHANGE PLA-7264 dated 12/05/2014 49 of 110 SSES EAL Comparison Matrix Table C-6 Max Safe Reactor Building Radiation Limits Max Safe Rad RB Area ARM Channel Limit Elevation (ft) ARM Number Description (RiHR)818 49 Refuel Floor Area 10 749 52 RWCU Recirc PP Access 10 54 Fuel Pool PP Area 719 50 CRD North 10 51 CRD South 670 53 Remote Shutdown Room 10 645 48 HPCI PP & Turbine Room 10 57 RCIC PP & Turbine Room 55 RHRA C PP Room 56 RHR B D PP Room 50 of 110 SSES EAL Comparison Matrix Category D Permanently Defueled Station Malfunction 51 of 110 SSES EAL Comparison Matrix SSES NEI IC# NEI IC Wording IC#(s) SSES IC Wording Difference/Deviation Justification PD-AU1 Recognition Category D N/A N/A NEI Recognition Category PD ICs and EALs are applicable only to PD-AU2 Permanently Defueled Station permanently defueled stations. SSES is not a defueled station.PD-SUI PD-HU1 PD-HU2 PD-HU3 PD-AA1 PD-AA2 PD-HA1 PD-HA3 52 of 110 SSES EAL Comparison Matrix Category E Independent Spent Fuel Storage Installation (ISFSI)53 of 110 SSES EAL Comparison Matrix SSES NEI IC# NEI IC Wording IC#(s) SSES IC Wording Difference/Deviation Justification E-HU1 Damage to a loaded cask EU1 Damage to a loaded cask None CONFINEMENT BOUNDARY CONFINEMENT BOUNDARY MODE: All MODE: All NEI Ex. SSES EAL E NEI Example EAL Wording EAL # SSES EAL Wording Difference/Deviation Justification 1 Damage to a loaded cask EUI.1 Damage to a loaded canister The values shown represent 2 times the limits specified in the ISFSI CONFINEMENT BOUNDARY as confinement boundary as Certificate of Compliance Technical Specification 1.2.7, HSM Dose indicated by an on-contact indicated by a radiation reading Rates.radiation reading greater than (2 on a loaded spent fuel cask > Removed phrase" which is also used in Recognition Category A IC times the site-specific cask any of the following: AUI" from bases since was confusing to end users.specific technical specification allowable radiation level) on the 800 mrem/hr at 3 ft from surface of the spent fuel cask. the HSM surface* 200 mrem/hr on contact on the outside of the HSM door centerline

40 mrem/hr on contact on the end shield wall exterior 54 of 110 SSES EAL Comparison Matrix Category F Fission Product Barrier Degradation 55 of 110 SSES EAL Comparison Matrix SSES NEI IC# NEI IC Wording IC#(s) SSES IC Wording Difference/Deviation Justification FA1 Any Loss or any Potential Loss of FA1 Any loss or any potential loss of Deleted "Hot Standby" because BWRs do not have this operating either the Fuel Clad or RCS EITHER Fuel Clad or RCS mode.barrier. MODE: 1 -Power Operation, 2 -MODE: Power Operation, Hot Startup, 3 -Hot Shutdown Standby, Startup, Hot Shutdown NEI Ex. NEI Example EAL Wording SSES SSES EAL Wording Difference/Deviation Justification EAL # EAL #1 Any Loss or any Potential Loss of FAI.1 Any loss or any potential loss of Table F-1 provides the fission product barrier loss and potential loss either the Fuel Clad or RCS EITHER Fuel Clad or RCS thresholds.

barrier. barrier (Table F-i) Table F-2 provides a human factors enhancement mechanism to track the threshold conditions that define the Loss and Potential Loss of the three fission product barriers (Fuel Clad, Reactor Coolant System, and Containment) to assist with quickly determining which initiating condition for EALs FG1.1, FS1.1, or FA1.1 Js met.56 of 110 SSES EAL Comparison Matrix SSES NEI IC# NEI IC Wording IC#(s) SSES IC Wording Difference/Deviation Justification FS1 Loss or Potential Loss of any two FS1 Loss or potential loss of any two Deleted "Hot Standby" because BWRs do not have this operating barriers barriers mode.MODE: Power Operation, Hot MODE: 1 -Power Operation, 2 -Standby, Startup, Hot Shutdown Startup, 3 -Hot Shutdown NEI Ex. EIE IELWrin SSESI NEAL Ex NEI Example EAL Wording SEA SSES EAL Wording Difference/Deviation Justification EAL # ameg EAL _ __#i ___ 1 _____1 Loss or Potential Loss of any two barriers FS1.1 Loss or potential loss of any two barriers (Table F-1)Table F-1 provides the fission product barrier loss and potential loss thresholds. Table F-2 provides a human factors enhancement mechanism to track the threshold conditions that define the Loss and Potential Loss of the three fission product barriers (Fuel Clad, Reactor Coolant System, and Containment) to assist with quickly determining which initiating condition for EALs FG1.1, FS1.1, or FAI.1 is met.57 of 110 SSES EAL Comparison Matrix SSES NEI IC# NEI IC Wording IC#(s) SSES IC Wording Difference/Deviation Justification FG1 Loss of any two barriers and FG1 Loss of any two barriers and loss Deleted "Hot Standby" because BWRs do not have this operating Loss or Potential Loss of third or potential loss of the third mode.barrier barrier MODE: Power Operation, Hot MODE: 1 -Power Operation, 2 -Standby, Startup, Hot Shutdown Startup, 3 -Hot Shutdown NEI Ex. NEI Example EAL Wording SSES SSES EAL Wording Difference/Deviation Justification EAL # EAL # SSEL odn 1 Loss of any two barriers and FG1.1 Loss of any two barriers Table F-1 provides the fission product barrier loss and potential loss Loss or Potential Loss of third AND thresholds. barrier Table F-2 provides a human factors enhancement mechanism to Loss or potential loss of the third track the threshold conditions that define the Loss and Potential Loss barrier (Table F-i) of the three fission product barriers (Fuel Clad, Reactor Coolant System, and Containment) to assist with quickly determining which initiating condition for EALs FGI.1, FSI.1, or FAI.1 is met.58 of 110 SSES EAL Comparison Matrix BWR Fuel Clad Fission Product Barrier Degradation Thresholds NEI SSES NEI NEI Threshold Wording FPB SSES FPB Wording Difference/Deviation Justification FPB# FPB #(s)FC Loss RCS Activity FC Loss Primary coolant activity > 300 300 pCi/gm DE-1 31 is the site-specific indication for this reactor 1 A. (Site-specific indications that D2 pCi/gm 1-131 Dose Equivalent coolant activity.reactor coolant activity is greater In the NEI 99-01 Bases, revised the lower value from 2% to 1% in than 300 pCi/gm dose the sentence "Reactor coolant activity above this level is greater equivalent 1-131). than that expected for iodine spikes and corresponds to an approximate range of 1% to 5% fuel clad damage" to avoid confusion when interpreting the bases. 1% fuel clad damage has been calculated to correspond to 300 pCi/gm dose equivalent 1-131 at SSES.FC Loss RPV Water Level FC Loss SAGs entered Revised to read "SAGs entered." Requirements for Primary 2 A. Primary containment flooding Al Containment Flooding correspond to entry into the Severe Accident Guidelines (SAGs) and are established in EOP RPV Control and required. EOP RPV Flooding. Per the developers guide "The phrase, "Primary containment flooding required," should be modified to agree with the site-specific EOP phrase indicating exit from all EOPs and entry to the SAGs (e.g., drywell flooding required, etc.)." FC Loss Not Applicable N/A N/A N/A 3 Not Applicable 59 of 110 SSES EAL Comparison Matrix NEI SSES FPB NEI Threshold Wording FPB SSES FPB Wording Difference/Deviation Justification FPB# FPB #(s)FC Loss Primary Containment FC Loss CHRRM radiation > 3.OE+3 R/hr For a fuel failure event equivalent to approximately 1% of cladding Radiation D1 failure and an instantaneous and complete release of reactor A. Primary containment coolant to the primary containment, the response of the CHRRM radiation monitor reading monitors in the drywell will be approximately 3,450 R/hr immediately greater than (site-specific after shutdown (rounded to 3,000 R/hr, which approximates the value). dose rate 10 minutes after shutdown). This assumes that the release has occurred soon after reactor shutdown, and that the fuel cladding failures produce a coolant source term of 300 pCi/gm of I-131 dose equivalent just prior to the release into primary containment. In the NEI 99-01 Bases, revised the lower value from 2% to 1% in the sentence "Reactor coolant activity above this level is greater than that expected for iodine spikes and corresponds to an approximate range of 1% to 5% fuel clad damage" to avoid confusion when interpreting the bases. 1% fuel clad damage has been calculated to correspond to 300 pCi/gm dose equivalent 1-131 at SSES.Last sentence of NEI 99-01 Bases was changed to state -"There is no Fuel Clad Barrier Potential Loss threshold associated with Primary Containment Radiation." to eliminate confusion associated with PC -Potential Loss threshold for Primary Containment Radiation. FC Loss Other Indications N/A N/A No other site-specific Fuel Clad Loss indication has been identified 5 for SSES.A. (site-specific as applicable) FC Loss Emergency Director FC Loss Any condition in the opinion of None 6 Judgment F1 the Emergency Director/Recovery Manager that A. ANY condition in the opinion indicates loss of the fuel clad of the Emergency Director that indicates Loss of the Fuel Clad barrier Barrier.60 of 110 SSES EAL Comparison Matrix NEI SSES FPB NEI Threshold Wording FPB SSES FPB Wording Difference/Deviation Justification FPB# FPB #(s)FC RCS Activity N/A N/A N/A P-Loss Not Applicable FC RPV Water Level FC RPV level CANNOT BE -161 in. is the site-specific RPV water level corresponding to the top P-Loss A. RPV water level cannot be P-Loss RESTORED AND MAINTAINED of active fuel.> -161 in. or CANNOT be 2 restored and maintained above Al determined (site-specific RPV water level corresponding to the top of active fuel) or cannot be determined. FC Not Applicable N/A N/A N/A P-Loss Not Applicable 3 FC Primary Containment N/A N/A N/A P-Loss Radiation 4 Not Applicable FC Other Indications N/A N/A No other site-specific Fuel Clad Potential Loss indication has been P-Loss A. (site-specific as applicable) identified for SSES.5 FC Emergency Director FC Any condition in the opinion of None P-Loss Judgment P-Loss the Emergency A. Any condition in the opinion F1 Director/Recovery Manager that 6 f A.e Any rnditn Dinector the opindicates potential loss of the fuel of the Emergency Director that clad barrier indicates Potential Loss of the Fuel Clad Barrier.61 of 110 SSES EAL Comparison Matrix BWR RCS Fission Product Barrier Degradation Thresholds NEI SSES FPB NEI NEI IC Wording s) SSES FPB Wording Difference/Deviation Justification FPB# #(s)RCS Primary Containment RCS Loss Primary Containment 1.72 psig is the site-specific primary containment pressure Loss Pressure C1 pressure > 1.72 psig due to corresponding to the drywell high pressure scram and isolation 1 A. Primary containment pressure RCS leakage setpoint.greater than (site-specific value)due to RCS leakage.RCS RPV Water Level RCS Loss RPV level CANNOT BE -161 in. is the site-specific RPV water level corresponding to the Loss A.RPV water level cannot be Al RESTORED AND top of active fuel.MAINTAINED > -161 in. or 2 restored and maintained above CANNOT b deterined (site-specific RPV water level corresponding to the top of active fuel) or cannot be determined. RCS RCS Leak Rate RCS Loss UNISOLABLE break in any of Main Steam Line, HPCI Steam Line, RCIC Steam Line, RWCU, Loss A.UNISOLABLE break in ANY of B1 the following: and Feedwater are the site-specific systems with potential for high 3 the following: (site-specific

Main Steam Line energy line breaks.systems with potential for high-
HPCI Steam Line energy line breaks)OR
RCIC Steam Line B.Emergency RPV .RWCU Depressurization.
Feedwater RCS Loss Emergency RPV Added "... is required" to be consistent with EOP terminology.

B2 Depressurization is required 62 of 110 SSES EAL Comparison Matrix NEI NEI IC Wording SSES FPB SSES FPB Wording Difference/Deviation Justification FPB# #(s)RCS Primary Containment RCS Loss CHRRM radiation > 7.OE+0 Indication of a RCS leak into the drywell is added to qualify the Loss Radiation D1 R/hr with indication of a RCS radiation monitor indication to avoid declaring the loss of the RCS 4 A.Primary containment radiation leak inside the drywell barrier for situations where the radiation increase is not due to a monitor reading greater than primary system leak. For situations that involve failure of the Fuel (site-specific value). Clad Barrier alone, containment Radiation levels would increase to greater than 30 R/hr potentially giving a false indication of a loss of the RCS barrier. Therefore the EAL contains a qualifier to preclude over classification of the event if only the fuel clad barrier has failed. Indication of a leak should be determined by observing other containment indications such as sump level, drywell pressure and ambient temperature. CHRRM readings of approximately 3 R/hr indicate an instantaneous release of reactor coolant at normal operating concentrations of 1-131 to the drywell atmosphere. Adding this value to the normal CHRRM background readings of 3-4 R/hr (100% power normal operation) provides the value of 7 R/hr.Last sentence of NEI 99-01 Bases was changed to state -"There is no RCS Barrier Potential Loss threshold associated with Primary Containment Radiation." to eliminate confusion associated with PC -Potential Loss threshold for Primary Containment Radiation. RCS Other Indications N/A N/A No other site-specific RCS Loss indication has been identified for Loss A. (site-specific as applicable) SSES.5 RCS Emergency Director Judgment RCS Loss Any condition in the opinion None Loss A. ANY condition in the opinion F1 of the Emergency Director/Recovery Manager of the Emergency Director that that indicates loss of the RCS indicates Loss of the RCSbarrier Barrier.63 of 110 SSES EAL Comparison Matrix NEI SSES FPB FPB NEI IC Wording s) SSES FPB Wording Difference/Deviation Justification FPB# #(s)RCS Primary Containment N/A N/A N/A P-Loss 1 Pressure Not Applicable RCS RPV Water Level N/A N/A N/A P-Loss 2 Not Applicable RCS RCS Leak Rate RCS UNISOLABLE primary system Reference to EO-000-104 Tables 8 and 9 and Table F-3 and F-4 P-Loss 3 A.UNISOLABLE primary system P-Loss leakage that results in have been added for clarification. leakage that results in exceeding B1 exceeding EITHER of the This RCS Potential Loss threshold is limited to those Max Normal EITHER of the following: following: Reactor Building Radiation and Temperature Limits that can be 1 .Max Normal Operating a One or more Max Normal remotely determined from within the control room to support Temperature Reactor Building Radiation prompt classification. Limits Reference PPL letters to NRC: OR (EO-000-104 Table 9) that 2.Max Normal Operating Area can be read in the control SUSQUEHANNA STEAM ELECTRIC STATION REPLY TO A Radiation Level. room (Table F-3) NOTICE OF VIOLATION PLA-7212 dated 08/29/2014 OR SUSQUEHANNA STEAM ELECTRIC STATION NOTIFICATION SOne or more Max Normal OF COMMITMENT CHANGE PLA-7264 dated 12/05/2014 Reactor Building Area Temperature Limits (EO-000-1 04 Table 8) that can be read in the control room (Table F-4)RCS Primary Containment N/A N/A N/A P-Loss 4 Radiation Not Applicable RCS Other Indications N/A N/A No other site-specific RCS Potential Loss indication has been identified for SSES.P-Loss 5 A. (site-specific as applicable) 64 of 110 SSES EAL Comparison Matrix NEI SSES FPB NEB NEI IC Wording s) SSES FPB Wording Difference/Deviation Justification FPB# #(s)RCS Emergency Director Judgment RCS Any condition in the opinion of None P-Loss 6 A. ANY condition in the opinion P-Loss the Emergency Director/Recovery Manager that of the Emergency Director that Fl indicates potential loss of the indicates Potential Loss of the RCS barrier RCS Barrier.BWR Containment Fission Product Barrier Degradation Thresholds NEI SSES NEI IC Wording FPB SSES FPB Wording Difference/Deviation Justification FPB# FPB #(s)PC Loss Primary Containment Conditions PC Loss UNPLANNED rapid drop in Primary NEI PC Loss 1 has been split into two thresholds. 1 A. UNPLANNED rapid drop in primary C1 Containment pressure following Primary containment pressure following primary Containment pressure rise containment pressure rise OR PC Loss Primary Containment pressure NEI PC Loss 1 has been split into two thresholds. B. Primary containment pressure C2 response not consistent with LOCA response not consistent with LOCA conditions conditions. PC Loss RPV Water Level N/A N/A N/A 2 Not Applicable PC Loss Primary Containment Isolation PC Loss UNISOLABLE direct downstream NEI PC Loss 3 has been split into two thresholds. 3 Failure El pathway to the environment exists after A. UNISOLABLE direct downstream Primary Containment isolation signal pathway to the environment exists after 65 of 110 SSES EAL Comparison Matrix NEI SSES FPB NEI IC Wording FPB SSES FPB Wording Difference/Deviation Justification FPB# FPB #(s)primary containment isolation signal PC Loss Intentional Primary Containment venting NEI PC Loss 3 has been split into two thresholds. OR E2 per EP-DS-004 RPV and PC Venting EP-DS-004 RPV and PC Venting provides the B. Intentional primary containment guidance for PC venting.venting per EOPs OR C. UNISOLABLE primary system PC Loss UNISOLABLE primary system leakage Reference to EO-000-104 Tables 8 and 9 and Table leakage that results in exceeding B1 that results in exceeding EITHER of the F-5 and F-6 have been added for clarification. EITHER of the following: following: This RCS Potential Loss threshold is limited to those 1. Max Safe Operating

One or more Max Safe Reactor Max Safe Reactor Building Radiation and Temperature.

Building Radiation Limits Temperature Limits that can be remotely determined (EO-000-1 04 Table 9) that can be from within the control room to support prompt OR read in the control room (Table F-5) classification.

2. Max Safe Operating Area OR Reference PPL letters to NRC: Radiation Level.R One or more Max Safe Reactor SUSQUEHANNA STEAM ELECTRIC STATION Building area temperature Limits REPLY TO A NOTICE OF VIOLATION PLA-7212 (EO-000-104 Table 8) that can be dated 08/29/2014 read in the control room (Table F-6) SUSQUEHANNA STEAM ELECTRIC STATION NOTIFICATION OF COMMITMENT CHANGE PLA-7264 dated 12/05/2014 PC Loss Primary Containment Radiation N/A N/A N/A 4 Not Applicable PC Loss Other Indications N/A N/A No other site-specific Containment Loss indication has 5 A. (site-specific as applicable) been identified for SSES.PC Emergency Director Judgment PC Any condition in the opinion of the None Loss ANY condition in the opinion of the Loss Emergency Director/Recovery Manager 6 Emergency Director that indicates Loss F1 that indicates loss of the Primary of the Containment Barrier. Containment barrier 66 of 110 SSES EAL Comparison Matrix NEI SSES FPB NEI IC Wording FPB SSES FPB Wording Difference/Deviation Justification FPB# FPB #(s)PC Primary Containment Conditions PC Primary Containment pressure > 53 NEI PC Pot. Loss 1 has been split into three P-Loss A. Primary containment pressure P-Loss psig thresholds.

1 greater than (site-specific value) C1 53 psig is the maximum SSES containment pressure OR allowed by design.B. (site-specific explosive mixture) PC Deflagration concentrations exist inside NEI PC Pot. Loss 1 has been split into three exists inside primary containment P-Loss PC (H 2 > 6% AND 02 -> 5%) thresholds. OR C2 The minimum global deflagration hydrogen/oxygen C. HCTL exceeded. concentrations are 6% and 5%, respectively PC Heat Capacity Temperature Limit NEI PC Pot. Loss 1 has been split into three P-Loss (Figure -HCTL) exceeded thresholds. C3 PC RPV Water Level PC SAGs entered Revised to read "SAGs entered." Requirements for P-Loss A. Primary containment flooding P-Loss Primary Containment Flooding correspond to entry.Preuimary. c n into the Severe Accident Guidelines (SAGs) and are 2 required. Al established in EOP RPV Control and EOP RPV Flooding. Per the developers guide "The phrase,"Primary containment flooding required," should be modified to agree with the site-specific EOP phrase indicating exit from all EOPs and entry to the SAGs (e.g., drywell flooding required, etc.)." PC Primary Containment Isolation N/A N/A N/A P-Loss Failure 3 Not Applicable PC Primary Containment Radiation PC CHRRM radiation > 4.OE+4 R/hr A reading of 40,000 R/hr indicates the release of P-Loss A. Primary containment radiation P-Loss reactor coolant into the drywell with elevated activity 4 monitor reading greater than (site- D1 indicative of 20% fuel clad damage.specific value).67 of 110 SSES EAL Comparison Matrix NEINE CWodn SSES NEI IC Wording SSES FPB Wording Difference/Deviation Justification FPB# FPB #(s)PC Other Indications N/A N/A No other site-specific Containment Potential Loss P-Loss A. (site-specific as applicable) indication has been identified for SSES.5 PC Emergency Director Judgment PC Any condition in the opinion of the None P-Loss A. ANY condition in the opinion of the P-Loss Emergency Director/Recovery 6 Emergency Director that indicates 6A Manager that indicates potential loss Potential Loss of the Containment of the Primary Containment barrier Barrier.68 of 110 SSES EAL Comparison Matrix Category H Hazards and Other Conditions Affecting Plant Safety 69 of 110 SSES EAL Comparison Matrix SSES NEI IC# NEI IC Wording IC#(s) SSES IC Wording Difference/Deviation Justification HU1 Confirmed SECURITY HU1 Confirmed SECURITY None CONDITION or threat CONDITION or threat.MODE: All MODE: All SSES NEI Example EAL Wording EAL # SSES EAL Wording Difference/Deviation Justification A SECURITY CONDITION that HU1.1 A SECURITY CONDITION that Example EALs #1, 2 and 3 have been combined into a single EAL does not involve a HOSTILE does not involve a HOSTILE for ease of presentation and use.ACTION as reported by the (site- ACTION as reported by Security specific security shift supervision). Shift Supervision Notification of a credible security OR threat directed at the site. Notification of a credible security threat directed at the site A validated notification from the OR NRC providing information of an aircraft threat. A validated notification from the NRC providing information of an aircraft threat 70 of 110 SSES EAL Comparison Matrix SSES NEI IC# NEI IC Wording IC#(s) SSES IC Wording Difference/Deviation Justification HU2 Seismic event greater than OBE HU2 Seismic event greater than OBE None levels levels MODE: All MODE: All NEI Ex. NIEapeELWrig SSES ENEI Example EAL Wording E SSES EAL Wording Difference/Deviation Justification EAL # EAL #1 Seismic event greater than HU2.1 Seismic event greater than Ground motion acceleration of 0.05g is the Operating Basis Operating Basis Earthquake OPERATING BASIS Earthquake for Susquehanna. Earthquake Monitoring Panel 0C696 (OBE) as indicated by: EARTHQUAKE (OBE) as provides indication of strong motion, OBE, or SSE events for the indicated by seismic Unit 1 Containment Foundation, Unit 2 Containment Foundation, (seite-specifc ivendti tt ar einstrumentation in the Control and the ESW Pumphouse (as well as tape printout). Input from all OBE limits) Room recording level greater six channels are recorded when a trigger initiates the system. A than an OBE seismic event generally starts with an indication in the Control Room, Annunciator SEISMIC MON SYSTEM TRIGGERED (AR-016-G06) on 0C653. The OBE is signaled by an LED illuminated green on the upper panel adjacent to the label, OPERATING BASIS EARTHQUAKE. 71 of 110 SSES EAL Comparison Matrix SSES NEI IC# NEI IC Wording IC#(s) SSES IC Wording Difference/Deviation Justification HU3 Hazardous event. HU3 Hazardous event None MODE: All MODE: All NEI Ex. SSES EA E NEI Example EAL Wording EA SSES EAL Wording Difference/Deviation Justification EAL # EAL #1 A tornado strike within the HU3.1 A tornado strike within the None PROTECTED AREA. PROTECTED AREA 2 Internal room or area flooding of a HU3.2 Internal room or area FLOODING None magnitude sufficient to require of a magnitude sufficient to manual or automatic electrical require manual or automatic isolation of a SAFETY SYSTEM electrical isolation of a SAFETY component needed for the current SYSTEM component needed for operating mode. the current operating mode 3 Movement of personnel within the HU3.3 Movement of personnel within the None PROTECTED AREA is impeded PROTECTED AREA is due to an offsite event involving IMPEDED due to an offsite event hazardous materials (e.g., an involving hazardous materials offsite chemical spill or toxic gas (e.g., an offsite chemical spill or release). toxic gas release)4 A hazardous event that results in HU3.4 A hazardous event that results in Added reference to Note 7.on-site conditions sufficient to on-site conditions sufficient to prohibit the plant staff from prohibit the plant staff from accessing the site via personal accessing the site via personal vehicles. vehicles (Note 7)5 (Site-specific list of natural or N/A N/A No other site-specific hazard has been identified for SSES.technological hazard events)Note EAL #3 does not apply to routine N/A Note 7: This EAL does not The SSES note is intended to apply to generic example EAL #4, not 72 of 110 SSES EAL Comparison Matrix traffic impediments such as fog, apply to routine traffic #3 as specified in the generic guidance.snow, ice, or vehicle breakdowns impediments such as or accidents. fog, snow, ice, or vehicle breakdowns or accidents. 73 of 110 SSES EAL Comparison Matrix SSES NEI IC# NEI IC Wording IC#(s) SSES IC Wording Difference/Deviation Justification HU4 FIRE potentially degrading the HU4 FIRE potentially degrading the None level of safety of the plant. level of safety of the plant MODE: All MODE: All NEI Ex. SSES EA Ex NEI Example EAL Wording SL SSES EAL Wording Difference/Deviation Justification EAL # EAL #a.A FIRE is NOT extinguished HU4.1 A FIRE is not extinguished within Site-specific plant rooms and areas are listed in Table H-1 to within 15-minutes of ANY of the 15 min. of any of the following improve the readability of the EAL.following FIRE detection FIRE detection indications (Note The NEI 99-01 bases was clarified to read in part: indications: 1): " Report from the field (i.e.,

Report from the field (i.e., Upon receipt, operators will take prompt actions to confirm the visual observation) visual observation) validity of an initial fire alarm, indication, or report. For EAL assessment purposes, the knowledge that a fires exists starts at the" Receipt of multiple (more 9 Receipt of multiple (more time of a report from the field, .receipt of multiple fire detection alarms than 1) fire alarms or than 1) fire alarms or or indications, or field confirmation of a single alarm or indication.

indications indications The fire duration clock starts at the time of knowledge that a fire" Field verification of a single

Field verification of a single exists.fire alarm fire alarm The original language AND AND "Upon receipt, operators will take prompt actions to confirm the validity of an initial fire alarm, indication, or report. For EAL b.The FIRE is located within ANY The FIRE is located within any assessment purposes, the emergency declaration clock starts at the of the following plant rooms or Table H-i area time that the initial alarm, indication, or report was received, and not the time that a subsequent verification action was performed.(site-specific list of plant rooms or Similarly, the fire duration clock also starts at the time of receipt of areas) the initial alarm, indication or report" could be interpreted to preclude the need to have the distinction of"receipt of multiple (more than 1) fire alarms or indications" and "field verification of a single fire alarm" in the EAL. It would also preclude the need for HU4.2 in its entirety.74 of 110 SSES EAL Comparison Matrix 2 a.Receipt of a single fire alarm (i.e., no other indications of a FIRE).AND b.The FIRE is located within ANY of the following plant rooms or areas: (site-specific list of plant rooms or areas)AND c.The existence of a FIRE is not verified within 30-minutes of alarm receipt.HU4.2 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 (i.e., proved or disproved) within 30 min. of alarm receipt (Note 1)Site-specific plant rooms and areas are listed in Table H-1 to improve the readability of the EAL.Added the phrase (i.e., proved or disproved) to eliminate the possibility of misinterpretation.

Phrase is directly from the NEI 99-01 Bases for this EAL.3 A FIRE within the plant or ISFSI HU4.3 A FIRE within the plant SSES does not have an ISFSI located outside the plant Protected[for plants with an ISFSI outside PROTECTED AREA not Area.the plant Protected Area] extinguished within 60 min. of the PROTECTED AREA not initial report, alarm or indication extinguished within 60-minutes of (Note 1)the initial report, alarm or indication. 4 A FIRE within the plant or ISFSI HU4.4 A FIRE within the plant SSES does not have an ISFSI located outside the plant Protected[for plants with an ISFSI outside PROTECTED AREA that Area.the plant Protected Area] requires firefighting support by PROTECTED AREA that requires an offsite fire response agency to firefighting support by an offsite extinguish fire response agency to extinguish. Note Note:The Emergency Director N/A Note 1:The ED/RM should The classification timeliness note has been standardized across the should declare the Unusual Event declare the event promptly upon SSES EAL scheme by referencing the "time limit" specified within promptly upon determining that determining that time limit has the EAL wording.the applicable time has been been exceeded, or will likely be exceeded, or will likely be exceeded._exceeded. I I 75 of 110 SSES EAL Comparison Matrix Table H-I Fire Areas* Control Structure" Diesel Generator Buildings* ESSW Pump House* Reactor Buildings* Turbine Buildings* ISFSI 76 of 110 SSES EAL Comparison Matrix SSES NEI IC# NEI IC Wording IC#(s) SSES IC Wording Difference/Deviation Justification HU7 Other conditions exist which in the HU7 Other conditions existing that in None judgment of the Emergency the judgment of the ED/RM Director warrant declaration of a warrant declaration of a UE (NO)UE MODE: All MODE: All NEI Ex. SSES EA E NEI Example EAL Wording EL SSES EAL Wording Difference/Deviation Justification EAL # EAL #Other conditions exist which in the HU7.1 Other conditions exist which in None judgment of the Emergency the judgment of the ED/RM Director indicate that events are in indicate that events are in progress or have occurred which progress or have occurred which indicate a potential degradation of indicate a potential degradation the level of safety of the plant or of the level of safety of the plant indicate a security threat to facility or indicate a security threat to protection has been initiated. No facility protection has been releases of radioactive material initiated. No releases of requiring offsite response or radioactive material requiring monitoring are expected unless offsite response or monitoring further degradation of safety are expected unless further systems occurs. degradation of SAFETY SYSTEMS occurs.77 of 110 SSES EAL Comparison Matrix SSES NEI IC# NEI IC Wording IC#(s) SSES IC Wording Difference/Deviation Justification HAI HOSTILE ACTION within the HAl HOSTILE ACTION within the None OWNER CONTROLLED AREA or OWNER CONTROLLED AREA airborne attack threat within 30 or airborne attack threat within 30 minutes. minutes MODE: All MODE: All NEI Ex. SSES EL E NEI Example EAL Wording EA SSES EAL Wording Difference/Deviation Justification EAL # EAL #1 A HOSTILE ACTION is occurring or HAl.1 A HOSTILE ACTION is Example EALs #1 and #2 have been combined into a single EAL has occurred within the OWNER occurring or has occurred within for ease of use.CONTROLLED AREA as reported the OWNER CONTROLLED by the (site-specific security shift AREA as reported by the supervision). Security Shift Supervision 2 A validated notification from NRC of OR an aircraft attack threat within 30 A validated notification from minutes of the site. NRC of an aircraft attack threat within 30 min. of the site 78 of 110 SSES EAL Comparison Matrix SSES NEI IC# NEI IC Wording IC#(s) SSES IC Wording Difference/Deviation Justification HA5 Gaseous release impeding HA5 Gaseous release IMPEDING HA5.1 mode applicability is Mode All for the control rooms and access to equipment necessary access to equipment necessary Mode 3 only for the remaining areas because those Mode for normal plant operations, for normal plant operations, dependent safe operation and shutdown areas are only applicable in cooldown or shutdown. cooldown or shutdown Mode 3 per Attachment 3 of the bases document.MODE: All MODE: All NEI Ex. SSES EAL E NEI Example EAL Wording EAL # SSES EAL Wording Difference/Deviation Justification a.Release of a toxic, corrosive, HA5.1 Release of a toxic, corrosive, Plant rooms or areas with entry-related mode applicability are listed asphyxiant or flammable gas asphyxiant or flammable gas into in Table H-2 to improve the readability of the EAL.into any of the following plant any Table H-2 area rooms or areas: AND (site-specific list of plant rooms Entry into the area is prohibited or or areas with entry-related mode Entr iNote a o applicability identified) IMPEDED (Note 5)AND b. Entry into the room or area is prohibited or impeded.Note Note:lf the equipment in the N/A Note 5:If the equipment in the None listed room or area was already listed area was already inoperable or out-of-service inoperable or out-of-before the event occurred, then service before the event no emergency classification is occurred, then no warranted. emergency classification is warranted. 79 of 110 SSES EAL Comparison Matrix Table H-2 Safe Operation & Shutdown Areas Elevation Unit I Area(s) ** Unit 2 Area(s) ** Mode(s)670' RB 27 RB 32 3,4,5 683' RB 27, 28, 29 RB 32, 33, 34 3,4,5 703' RB 28, 29 RB 33, 34 3,4,5 719' RB 25, 29 RB 30, 34 3,4,5 749' RB 25, 29 RB 32, 33 3,4,5 729' CS 12, 21 CS 12, 21 1, 2, 3, 4, 5, D** See Chart I for location of plant areas Chart 1- Plant Area Key Plan 70 LOW LEVEL RADWASTE_ ACID AND CHLORINE WATER TREATMENT 53 52 51 50 H SPRAY POND VALVE VAULT 47 BUlL.DING PUMPHOUSE ESSW y 332 ($~ ~ cmcPUMPHOUSE -I UNIT#2 ? UNIT91 ?TURBINE BLDG. TURBINE BLDG RADWASTE TUR. 1 16 15 14 1 13 4 13 2 1 E , 20 19 18 1 17 8 17 6 5 38 1137 040+39 42 4 19 24 23 Q-/ 45 COND. STORAGE 22 21 12 11 10 9 1=1

I = -32 30 27 25 36 44 35 43COND & REF STORAGE-DIESEL GENERATOR E #2#1_ RIVER INTAKE STRUCTURE REACTOR REACTOR+81 E DIESEL GENERATOR 80 of 110 SSES EAL Comparison Matrix SSES NEI IC# NEI IC Wording IC#(s) SSES IC Wording Difference/Deviation Justification HA6 Control Room evacuation HA6 Control Room evacuation None resulting in transfer of plant resulting in transfer of plant control to alternate locations, control to alternate locations MODE: All MODE: All NEI Ex. SSES EAL # NEI Example EAL Wording EAL # SSES EAL Wording Difference/Deviation Justification 1 An event has resulted in plant HA6.1 An event has resulted in plant Remote Shutdown Panels are the SSES site-specific remote control being transferred from the control being transferred from the shutdown panels and local control stations.Control Room to (site-specific Control Room to the Remote remote shutdown panels and Shutdown Panels local control stations).

81 of 110 SSES EAL Comparison Matrix NEI C# NI I WoringSSES NEI IC# NEI IC Wording IC#(s) SSES IC Wording Difference/Deviation Justification HA7 Other conditions exist which in the HA7 Other conditions exist that in the None judgment of the Emergency Director judgment of the ED/RM warrant warrant declaration of an Alert. declaration of an Alert MODE: All MODE: All NEI Ex. SSES EA Ex NEI Example EAL Wording EA SSES EAL Wording Difference/Deviation Justification EAL # EAL #1 Other conditions exist which, in the HA7.1 Other conditions exist which, in the None judgment of the Emergency Director, judgment of the ED/RM, indicate that indicate that events are in progress or events are in progress or have occurred have occurred which involve an actual or which involve an actual or potential potential substantial degradation of the substantial degradation of the level of level of safety of the plant or a security safety of the plant or a security event that event that involves probable life involves probable life threatening risk to threatening risk to site personnel or site personnel or damage to site damage to site equipment because of equipment because of HOSTILE HOSTILE ACTION. Any releases are ACTION. Any releases are expected to expected to be limited to small fractions be limited to small fractions of the EPA of the EPA Protective Action Guideline Protective Action Guideline exposure exposure levels, levels.82 of 110 SSES EAL Comparison Matrix SSES NEI IC# NEI IC Wording IC#(s) SSES IC Wording Difference/Deviation Justification HS1 HOSTILE ACTION within the HS1 HOSTILE ACTION within the None PROTECTED AREA PROTECTED AREA MODE: All MODE: All NEI Ex. NEI Example EAL Wording SSES SSES EAL Wording Difference/Deviation Justification EAL # EAL #1 A HOSTILE ACTION is occurring HS1.1 A HOSTILE ACTION is occurring or has None or has occurred within the occurred within the PROTECTED AREA PROTECTED AREA as reported as reported by the Security Shift by the (site-specific security shift Supervision supervision). 83 of 110 SSES EAL Comparison Matrix SSES NEI IC# NEI IC Wording IC#(s) SSES IC Wording Difference/Deviation Justification HS6 Inability to control a key safety HS6 Inability to control a key safety function None function from outside the Control from outside the Control Room Room. MODE: All MODE: All NEI Ex. NEI Example EAL Wording SSES SSES EAL Wording Difference/Deviation Justification EAL# EAL #a.An event has resulted in plant HS6.1 An event has resulted in plant control Remote Shutdown Panels are the SSES site-specific remote control being transferred from the being transferred from the Control Room shutdown panels and local control stations.Control Room to (site-specific to the Remote Shutdown Panels Deleted the word "control" after "reactivity" as it is redundant. remote shutdown panels and local AND control stations). AND Control of any of the following key safety functions is not reestablished within 15 b.Control of ANY of the following min. (Note 1): key safety functions is not reestablished within (site-specific number of minutes). e RPV water level" Reactivity control

RCS heat removal" Core cooling [PWR] I RPV water level [BWR]" RCS heat removal 84 of 110 SSES EAL Comparison Matrix SSES NEI IC# NEI IC Wording IC#(s) SSES IC Wording Difference/Deviation Justification HS7 Other conditions exist which in HS7 Other conditions existing that in the None the judgment of the Emergency judgment of the ED/RM warrant Director warrant declaration of a declaration of a Site Area Emergency Site Area Emergency.

MODE: All MODE: All NEI Ex. SSES EA NEI Example EAL Wording SL SSES EAL Wording Difference/Deviation Justification EAL # EAL #Other conditions exist which in HS7.1 Other conditions exist which in the The EMERGENCY PLAN BOUNDARY is the site-specific the judgment of the Emergency judgment of the ED/RM indicate that receptor point.Director indicate that events are events are in progress or have occurred in progress or have occurred which involve actual or likely major which involve actual or likely failures of plant functions needed for major failures of plant functions protection of the public or HOSTILE needed for protection of the ACTION that results in intentional damage public or HOSTILE ACTION that or malicious acts, (1) toward site results in intentional damage or personnel or equipment that could lead to malicious acts, (1) toward site the likely failure of or, (2) that prevent personnel or equipment that effective access to equipment needed for could lead to the likely failure of the protection of the public. Any releases or, (2) that prevent effective are not expected to result in exposure access to equipment needed for levels which exceed EPA Protective the protection of the public. Any Action Guideline exposure levels beyond releases are not expected to the EMERGENCY PLAN BOUNDARY.result in exposure levels which exceed EPA Protective Action Guideline exposure levels beyond the site boundary.85 of 110 SSES EAL Comparison Matrix SSES NEI IC Wording IC#(s) SSES IC Wording Difference/Deviation Justification HOSTILE ACTION resulting in HG1 HOSTILE ACTION resulting in loss of None loss of physical control of the physical control of the facility facility. MODE: All MODE: All NEI Ex. SSES EAL # NEI Example EAL Wording EAL # SSES EAL Wording Difference/Deviation Justification 1 a.A HOSTILE ACTION is HG1.1 A HOSTILE ACTION is occurring or has Deleted the word "control" after "reactivity" as it is redundant. occurring or has occurred within occurred within the PROTECTED AREA the PROTECTED AREA as as reported by the Security Shift reported by the (site-specific Supervision security shift supervision). AND AND EITHER of the following has occurred: b.EITHER of the following has Any of the following safety functions occurred: cannot be controlled or maintained

1. ANY of the following safety

Reactivity functions cannot be controlled or maintained.

9 RPV water level" Reactivity control

RCS heat removal" Core cooling OR[PWR]IRPV water Damage to spent fuel has occurred level [BWR] or is IMMINENT" RCS heat removal OR 2. Damage to spent fuel has occurred or is IMMINENT.86 of 110 SSES EAL Comparison Matrix SSES NEI IC# NEI IC Wording IC#(s) SSES IC Wording Difference/Deviation Justification HG7 Other conditions exist which in HG7 Other conditions exist which in the None the judgment of the Emergency judgment of the ED warrant declaration Director warrant declaration of a of a General Emergency General Emergency MODE: All MODE: All NE! Ex. SSES EA E NEI Example EAL Wording EA SSES EAL Wording Difference/Deviation Justification EAL # EAL #1 Other conditions exist which in HG7.1 Other conditions exist which in the None the judgment of the Emergency judgment of the ED/RM indicate that Director indicate that events are events are in progress or have occurred in progress or have occurred which involve actual or IMMINENT which involve actual or substantial core degradation or melting IMMINENT substantial core with potential for loss of containment degradation or melting with integrity or HOSTILE ACTION that potential for loss of containment results in an actual loss of physical integrity or HOSTILE ACTION control of the facility.

Releases can be that results in an actual loss of reasonably expected to exceed EPA physical control of the facility. Protective Action Guideline exposure Releases can be reasonably levels offsite for more than the expected to exceed EPA immediate site area.Protective Action Guideline exposure levels offsite for more than the immediate site area.87 of 110 SSES EAL Comparison Matrix Category S System Malfunction 88 of 110 SSES EAL Comparison Matrix SSES NEI IC# NEI IC Wording IC#(s) SSES IC Wording Difference/Deviation Justification SU1 Loss of all offsite AC power SUl Loss of all offsite AC power "essential" is the site-specific designation for the SSES emergency capability to emergency buses for capability to essential buses for AC buses.15 minutes or longer. 15 minutes or longer MODE: Power Operation, Startup, MODE: 1 -Power Operation, 2 -Hot Standby, Hot Shutdown Startup, 3 -Hot Shutdown NEI Ex.SSES NEI Ex. NEI Example EAL Wording EA SSES EAL Wording Difference/Deviation Justification EAL # EAL #1 Loss of ALL offsite AC power SUI.1 Loss of all offsite AC power The ESS buses are the site-specific emergency buses.capability to (site-specific capability to all 4.16 kV ESS Added "on EITHER unit" to clarify applicability to either unit.emergency buses) for 15 minutes buses on EITHER unit for or longer. > 15 min. (Note 1)Note The Emergency Director should N/A Note 1: The ED/RM should The classification timeliness note has been standardized across the declare the Unusual Event declare the event SSES EAL scheme by referencing the "time limit" specified within promptly upon determining that promptly upon the EAL wording.15 minutes has been exceeded, determining that time or will likely be exceeded. limit has been exceeded, or will likely be exceeded.89 of 110 SSES EAL Comparison Matrix SSES NEI IC# NEI IC Wording IC#(s) SSES IC Wording Difference/Deviation Justification SU2 UNPLANNED loss of Control SU3 UNPLANNED loss of Control None Room indications for 15 minutes Room indications for 15 minutes or longer, or longer.MODE: Power Operation, MODE: 1 -Power Operation, 2 -Startup, Hot Standby, Hot Startup, 3 -Hot Shutdown Shutdown NEI Ex. SSES EAL #E NEI Example EAL Wording EAL # SSES EAL Wording Difference/Deviation Justification 1 An UNPLANNED event results in SU3.1 An UNPLANNED event results in The site-specific Safety System Parameter list is tabulated in Table the inability to monitor one or the inability to monitor one or S-1.more of the following parameters more Table S-1 parameters from from within the Control Room for within the Control Room for > 15 15 minutes or longer. min. (Note 1)Note The Emergency Director should N/A Note 1: The ED/RM should The classification timeliness note has been standardized across the declare the Unusual Event declare the event SSES EAL scheme by referencing the "time limit" specified within the promptly upon determining that promptly upon EAL wording.15 minutes has been exceeded, determining that time or will likely be exceeded. limit has been exceeded, or will likely be exceeded.90 of 110 SSES EAL Comparison Matrix[BWR parameter list] [PWR parameter list]Reactor Power Reactor Power RPV Water Level RCS Level RPV Pressure RCS Pressure Primary Containment Pressure In-Core/Core Exit Temperature Suppression Pool Level Levels in at least (site-specific number)steam generators Suppression Pool Temperature Steam Generator Auxiliary or Emergency Feed Water Flow Table S-I Safety System Parameters

Reactor power* RPV water level* RPV pressure* Primary Containment pressure* Suppression Pool water level* Suppression Pool temperature 91 of 110 SSES EAL Comparison Matrix NEI C# NI I WoringSSES NEI IC# NEI IC Wording IC#(s) SSES IC Wording Difference/Deviation Justification SU3 Reactor coolant activity greater SU4 Reactor coolant activity greater None than Technical Specification than Technical Specification allowable limits, allowable limits MODE: Power Operation, Startup, MODE: 1 -Power Operation, 2 -Hot Standby, Hot Shutdown Startup, 3 -Hot Shutdown NEI Ex. NEI Example EAL Wording SSES SSES EAL Wording Difference/Deviation Justification EAL # EAL #1 (Site-specific radiation monitor) SU4.1 Offgas pretreatment monitor high- The Offgas Pretreatment RMS monitors radioactivity in the Offgas reading greater than (site-specific high radiation alarm system downstream of the Motive Steam Jet Condenser.

The value), monitor detects the radiation level that is attributable to the fission gases produced in the reactor and transported with steam through the turbine to the condenser. 2 Sample analysis indicates that a SU4.2 Coolant activity > 0.2 pCi/gm The specific iodine activity is limited to < 0.2 pCi/gm Dose Equivalent reactor coolant activity value is dose equivalent 1-131 for > 48 1-131. This limit ensures the source term assumed in the safety greater than an allowable limit hours analysis for the Main Steam Line Break (MSLB) outside containment specified in Technical is not exceeded, so any release of radioactivity to the environment Specifications. OR during an MSLB is less than a small fraction of the regulatory limits.Coolant activity > 4.0 pCi/gm The upper limit of 4.0 pCi/gm Dose Equivalent 1-131ensures that the dose equivalent 1-131 at any time thyroid dose from an MSLB will not exceed the dose guidelines of 10 CFR 50.67 or Control Room operator dose limits specified in GDC 19 of 10 CFR 50, Appendix A.92 of 110 SSES EAL Comparison Matrix SSES NEI IC# NEI IC Wording IC#(s) SSES IC Wording Difference/Deviation Justification SU4 RCS leakage for 15 minutes or SU5 RCS leakage for 15 minutes or None longer, longer MODE: Power Operation, MODE: 1 -Power Operation, 2 -Startup, Hot Standby, Hot Startup, 3 -Hot Shutdown Shutdown NEI Ex. NEI Example EAL Wording SSES SSES EAL Wording Difference/Deviation Justification EAL # EAL #1 RCS unidentified or pressure SU5.1 RCS unidentified or pressure Example EALs #1, 2 and 3 have been combined into a single EAL boundary leakage greater than boundary leakage> 10 gpm for > for usability.(site-specific value) for 15 15 min. Statements in the SSES bases were added for clarification: minutes or longer. OR Drywell leakage calculations in SO-1 00(200)-006 take a finite 2 RCS identified leakage greater RCS identified leakage > 25 gpm period of time to complete. Leakage rates cannot be determined than (site-specific value) for 15 for > 15 min. quickly by merely observing an indicator. For this reason, the 15 minutes or longer. Ominutes clock starts after it is determined that leakage rates exceed gOR the entry value. Upon determination that leakage has increased 3 Leakage from the RCS to a Leakage from the RCS to a substantially, effort should be made to quantify this leakage in a location outside containment location outside Primary timely manner.greater than 25 gpm for 15 Containment > 25 gpm for > 15 ON-1(2)00-005, "Excessive Drywell Leakage Identification", minutes or longer. mi. contains methods of quickly estimating drywell leakage. These (Note 1) methods can be used in lieu of completing the calculations contained in SO-1(2)00-006. Means to directly quantify RCS leakage outside containment may not be available. For this reason, judgment must be used for assessment of the 25 gpm leak rate criterion. For example, a short steam plume that does not appreciably change room temperature or room radiation levels can be judged to be less than 25 gpm. A leak that causes room temperature to rise rapidly above maximum safe temperatures could be judged to be greater than 25 gpm in the absence of measurable leak rates, and thus judgment is an acceptable method to evaluate this criterion. 93 of 110 SSES EAL Comparison Matrix Note The Emergency Director should N/A Note 1: The ED/RM should The classification timeliness note has been standardized across the declare the Unusual Event declare the event SSES EAL scheme by referencing the "time limit" specified within promptly upon determining that promptly upon the EAL wording.15 minutes has been exceeded, determining that time or will likely be exceeded. limit has been exceeded, or will likely be exceeded.SSES NEI IC# NEI IC Wording IC#(s) SSES IC Wording Difference/Deviation Justification SU5 Automatic or manual (trip SU6 Automatic or manual scram fails Included Mode 2 Startup consistent with developer note. Reactor[PWR] / scram [BWR]) fails to to shut down the reactor power can be above the APRM downscale shutdown threshold of 5%shutdown the reactor. MODE: 1 -Power Operation, 2 -while still in Mode 2.MODE: Power Operation Startup NEI Ex. SSES EAL E NEI Example EAL Wording EAL # SSES EAL Wording Difference/Deviation Justification 1 a.An automatic (trip [PWR] / SU6.1 An automatic scram did not shut Added the phrase "... after any RPS setpoint is exceeded "to clarify scram [BWR]) did not shutdown down the reactor after any RPS that it is a failure of the automatic scram when a valid scram signal the reactor. setpoint is exceeded has been exceeded.AND AND Added "subsequent automatic scram or..." to take credit for automatic initiation of ARI subsequent to an RPS failure to complete b. A subsequent manual action A subsequent automatic scram an automatic scram. Automatic initiation of ARI to initiate an taken at the reactor control or manual scram action taken at automatic scram is a design feature of ARI.consoles is successful in the reactor control console shutting down the reactor. (Manual PBs, Mode Switch, ARI) Manual PBs, Mode Switch, and initiation of ARI are the manual is successful in shutting down the actions taken to shut down the reactor.reactor as indicated by reactor Reactor power below 5% (APRM downscale) is the site-specific power < 5% (APRM downscale) indication of a successful reactor scram.(Note 8)2 a.A manual trip ([PWR] / scram SU6.2 A manual scram did not shut Added the phrase "... after any manual scram action was initiated" to I [BWR]) did not shutdown the down the reactor after any clarify that it is a failure of any manual scram when an actual manual 94 of 110 SSES EAL Comparison Matrix manual scram action was reactor.AND b.EITHER of the following:

1. A subsequent manual action taken at the reactor control consoles is successful in shutting down the reactor.OR 2 A subsequent automatic (trip [PWR] / scram[BWR]) is successful in shutting down the reactor.manual scram action was initiated AND A subsequent automatic scram or manual scram action taken at the reactor control console (Manual PBs, Mode Switch, ARI)is successful in shutting down the reactor as indicated by reactor power < 5% (APRM downscale)(Note 8)scram signal has been inserted.Combined conditions b.1 and b.2 into a single statement to simplify the. presentation.

Manual PBs, Mode Switch, and initiation of ARI are the manual actions taken to shut down the reactor.Reactor power below 5% (APRM downscale) is the site-specific indication of a successful reactor scram.Notes Note:A manual action is any N/A Note 8: A manual scram action Added the scram to actions to be consistent with EAL wording.operator action, or set of actions, is any operator action, or which causes the control rods to set of actions, which be rapidly inserted into the core, causes the control rods and does not include manually to be rapidly inserted driving in control rods or into the core, and does implementation of boron not include manually injection strategies. driving in control rods or implementation of boron injection strategies. 95 of 110 SSES EAL Comparison Matrix SSES NEI IC# NEI IC Wording IC#(s) SSES IC Wording Difference/Deviation Justification SU6 Loss of all onsite or offsite SU7 Loss of all onsite or offsite None communications capabilities, communications capabilities. MODE: Power Operation, MODE: 1 -Power Operation, 2 -Startup, Hot Standby, Hot Startup, 3 -Hot Shutdown Shutdown NEI Ex. SSES EAL # NEI Example EAL Wording EAL # SSES EAL Wording Difference/Deviation Justification 1 Loss of ALL of the following SU7.1 Loss of all Table S-3 onsite Example EALs #1, 2 and 3 have been combined into a single EAL onsite communication methods: communication methods for simplification of presentation.(site-specific list of OR Table S-3 provides a site-specific list of onsite, ORO and NRC communications methods) Loss of all Table S-30RO communications methods.2 Loss of ALL of the following communication methods ORO communications methods: OR (site-specific list of Loss of all Table S-3 NRC communications methods) communication methods 3 Loss of ALL of the following NRC communications methods: (site-specific list of communications methods)96 of 110 SSES EAL Comparison Matrix Table S-3 Communication Methods System Onsite ORO NRC UHF Radio X Plant PA System X Dedicated Conference Lines X Commercial Telephone Systems X X X Cellular Telephone X X FTS-2001 (ENS) X X 97 of 110 SSES EAL Comparison Matrix SSES NEI IC# NEI IC Wording IC#(s) SSES IC Wording Difference/Deviation Justification SU7 Failure to isolate containment or N/A N/A This IC and its associated example EALs are applicable to PWRs loss of containment pressure only and therefore not included.control. [PWR]MODE: Hot Standby, Hot Shutdown NEI Ex. SSES EA E NEI Example EAL Wording SL SSES EAL Wording Difference/Deviation Justification EAL # EAL #1 a.Failure of containment to N/A N/A This IC and its associated example EALs are applicable to PWRs isolate when required by an only and therefore not included.actuation signal.AND b.ALL required penetrations are not closed within 15 minutes of the actuation signal.2 a.Containment pressure greater N/A N/A This IC and its associated example EALs are applicable to PWRs than (site-specific pressure). only and therefore not included.AND b.Less than one full train of (site-specific system or equipment) is operating per design for 15 minutes or longer.98 of 110 SSES EAL Comparison Matrix SSES NEI IC# NEI IC Wording IC#(s) SSES IC Wording Difference/Deviation Justification SA1 Loss of all but one AC power SA1 Loss of all but one AC power 'essential" is the site-specific designation for the SSES emergency source to emergency buses for source to essential buses for 15 AC buses.15 minutes or longer. minutes or longer.MODE: Power Operation, MODE: 1 -Power Operation, 2 -Startup, Hot Standby, Hot Startup, 3 -Hot Shutdown Shutdown NEI Ex. NEI Example EAL Wording SSES SSES EAL Wording Difference/Deviation Justification EAL # EAL #1 a.AC power capability to (site- SAI.1 AC power capability to all 4.16 ESS buses are the site-specific emergency buses.specific emergency buses) is kV ESS buses on EITHER unit Added "on EITHER unit" to clarify applicability to either unit.reduced to a single power reduced to a single power source for 15 minutes or longer, source for> 15 min. (Note 1)AND AND b. Any additional single power Any additional single power source failure will result in a loss source failure will result in loss of of all AC power to SAFETY all AC power to SAFETY SYSTEMS. SYSTEMS Note The Emergency Director should N/A Note 1: The ED/RM should The classification timeliness note has been standardized across declare the Alert promptly upon declare the event the SSES EAL scheme by referencing the "time limit" specified determining that 15 minutes has promptly upon within the EAL wording.been exceeded, or will likely be determining that time exceeded. limit has been exceeded, or will likely be exceeded.99 of 110 SSES EAL Comparison Matrix SSES NEI IC# NEI IC Wording IC#(s) SSES IC Wording Difference/Deviation Justification SA2 UNPLANNED loss of Control SA3 UNPLANNED loss of Control None Room indications for 15 minutes Room indications for 15 minutes or longer with a significant or longer with a significant transient in progress. transient in progress.MODE: Power Operation, MODE: 1 -Power Operation, 2 -Startup, Hot Standby, Hot Startup, 3 -Hot Shutdown Shutdown NEI Ex. SSES EA E NEI Example EAL Wording SL SSES EAL Wording Difference/Deviation Justification EAL # EAL #1 An UNPLANNED event results in SA3.1 An UNPLANNED event results in The site-specific Safety System Parameters are listed in Table S-1.the inability to monitor one or the inability to monitor one or more of the following parameters more Table S-1 parameters from The significant transient list has been tabularized in Table S-2 for from within the Control Room for within the Control Room for > 15 15 minutes or longer. min. (Note 1) Revised transient runback to read "Runback > 25% reactor power".AND AND Full electrical load can change based on ambient conditions, which affect condenser backpressure and turbine efficiency, and is also ANY of the following transient Any significant transient is in affected by the generator capability curve.events in progress. progress, Table S-2 Deleted "Electrical load rejection > 25% electrical load" since any" Automatic or manual load rejection that occurs when >25% power results in a reactor runback greater than 25% scram.thermal reactor power Added RRC pump trip while > 25% reactor power as an additional" Electrical load rejection BWR transient that is significant. greater than 25% full electrical load* Reactor scram [BWR] I trip[PWR]" ECCS (SI) actuation" Thermal power oscillations 100 of 110 SSES EAL Comparison Matrix greater than 10% [BWR]F -4 F Note The Emergency Director should declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.N/A Note 1: The ED should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.The classification timeliness note has been standardized across the SSES EAL scheme by referencing the "time limit" specified within the EAL wording.[BWR parameter list] [PWR parameter list]Reactor Power Reactor Power RPV Water Level RCS Level RPV Pressure RCS Pressure Primary Containment Pressure In-Core/Core Exit Temperature Suppression Pool Level Levels in at least (site-specific number)steam generators Suppression Pool Temperature Steam Generator Auxiliary or Emergency Feed Water Flow Table S-1 Safety System Parameters

Reactor power* RPV water level* RPV pressure* Primary Containment pressure* Suppression Pool water level* Suppression Pool temperature 101 of 110 SSES EAL Comparison Matrix Table S-2 Significant Transients
Reactor scram* Runback > 25% reactor power* RRC pump trip while > 25% reactor power* ECCS injection* Thermal power oscillations

> 10%102 of 110 SSES EAL Comparison Matrix SSES NEI IC# NEI IC Wording IC#(s) SSES IC Wording Difference/Deviation Justification SA5 Automatic or manual (trip [PWR] SA6 Automatic or manual scram fails Included Mode 2 Startup consistent with developer note. Reactor/ scram [BWR]) fails to shutdown to shut down the reactor and power can be above the APRM downscale shutdown threshold of the reactor, and subsequent subsequent manual actions 5% while still in Mode 2.manual actions taken at the taken at the reactor control reactor control consoles are not consoles are not successful in successful in shutting down the shutting down the reactor reactor. MODE: 1 -Power Operation, 2 -MODE: Power Operation Startup NEI Ex. NEI Example EAL Wording SSES SSES EAL Wording Difference/Deviation Justification EAL # EAL #1 a.An automatic or manual (trip SA6.1 An automatic or manual scram Manual PBs, Mode Switch, and initiation of ARI are the manual[PWR] / scram [BWR]) did not fails to shut down the reactor actions taken to shut down the reactor.shutdown the reactor. AND Reactor power below 5% is the site-specific indication of a AND Manual scram actions taken at successful reactor scram.b.Manual actions taken at the the reactor control console reactor control consoles are not (Manual PBs, Mode Switch, ARI)successful in shutting down the are not successful in shutting reactor. down the reactor as indicated by reactor power > 5% (Note 8)103 of 110 SSES EAL Comparison Matrix SSES NEI IC# NEI IC Wording IC#(s) SSES IC Wording Difference/Deviation Justification SA9 Hazardous event affecting a SA8 Hazardous event affecting a None SAFETY SYSTEM needed for SAFETY SYSTEM required for the current operating mode. the current operating mode.MODE: Power Operation, MODE: 1 -Power Operation, 2 -Startup, Hot Standby, Hot Startup, 3 -Hot Shutdown Shutdown NEI Ex. NEI Example EAL Wording SSES SSES EAL Wording Difference/Deviation Justification

EAL # EAL #1 a.The occurrence of ANY of the SA8.1 The occurrence of any Table S- The hazardous events have been listed in Table S-4 to improve the following hazardous events: 4 hazardous event readability of the SSES EAL." Seismic event (earthquake)

AND" Internal or external flooding EITHER: event Event damage has caused" High winds or tornado strike indications of degraded" FIRE performance in at least one train of a SAFETY SYSTEM" EXPLOSION required for the current" (site-specific hazards) operating mode OR" Other events with similar hazard characteristics as The event has caused determined by the Shift VISIBLE DAMAGE to a Manager SAFETY SYSTEM AND component or structure required for the current b.EITHER of the following: operating mode 1. Event damage has caused indications of degraded performance in at least one train of a SAFETY 104 of 110 SSES EAL Comparison Matrix SYSTEM needed for the current operating mode.OR 2. The event has caused VISIBLE DAMAGE to a SAFETY SYSTEM component or structure needed for the current operating mode.Table S-4 Hazardous Events" Seismic event (earthquake)

Internal or external FLOODING event" High winds or tornado strike* FIRE* EXPLOSION* Other events with similar hazard characteristics as determined by the Shift Manager 105 of 110 SSES EAL Comparison Matrix 1SSES NEI IC# NEI IC Wording IC#(s) SSES IC Wording Difference/Deviation Justification SS1 Loss of all offsite and all onsite SS1 Loss of all offsite and all onsite "essential" is the site-specific designation for the SSES emergency AC power to emergency buses AC power to essential buses for AC buses.for 15 minutes or longer. 15 minutes or longer MODE: Power Operation, MODE: 1 -Power Operation, 2 -Startup, Hot Standby, Hot Startup, 3 -Hot Shutdown Shutdown NEI Ex. NEI Example EAL Wording SSES SSES EAL Wording Difference/Deviation Justification EAL # EAL #1 Loss of ALL offsite and ALL SS1.1 Loss of all offsite and all onsite ESS buses are the site-specific emergency buses.onsite AC power to (site-specific AC power capability to all 4.16 Added "on EITHER unit" to clarify applicability to either unit.emergency buses) for 15 kV ESS buses on EITHER unit minutes or longer, for > 15 min. (Note 1)Note The Emergency Director should declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.N/A Note 1: The ED/RM should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.The classification timeliness note has been standardized across the SSES EAL scheme by referencing the "time limit" specified within the EAL wording.106 of 110 SSES EAL Comparison Matrix SSES NEI IC# NEI IC Wording IC#(s) SSES IC Wording Difference/Deviation Justification SS5 Inability to shutdown the reactor SS6 Inability to shut down the Included Mode 2 Startup consistent with developer note. Reactor causing a challenge to (core reactor causing a challenge to power can be above the APRM downscale shutdown threshold of cooling [PWR] I RPV water level RPV water level or RCS heat 4% while still in Mode 2.[BWR]) or RCS heat removal, removal MODE: Power Operation MODE: 1 -Power Operation, 2 -Startup NEI Ex. NEI Example EAL Wording SSES SSES EAL Wording Difference/Deviation Justification EAL # EAL #1 a.An automatic or manual (trip SS6.1 An automatic or manual scram Reactor power < 5% is the site-specific indication of a successful

[PWR] / scram [BWR]) did not fails to shut down the reactor reactor scram.shutdown the reactor. AND Deleted the term "manual actions" from the second condition. For AND All actions to shut down the generic IC SS5, all actions to shut down the reactor can be credited, b.All manual actions to shutdown reactor are not successful as including emergency boration which is not considered a 'manual" the reactor have been indicated by reactor power > 5% scram action.unsuccessful. AND Indication of an inability to adequately remove heat from the core AND occurs when RPV water level cannot be restored and maintained EITHER: above -179 in., which is the EOP RPV water level indicative of a c.EITHER of the following RPV level CANNOT BE loss of adequate core cooling.conditions exist: RESTORED AND Indication of an inability to adequately remove heat from the RCS (Site-specific indication of MAINTAINED occurs when parameters are in the unsafe region of the HCTL an inability to adequately > -179 in. or CANNOT be curve.remove heat from the determined core) OR (Site-specific indication of Suppression pool water an inability to adequately temperature AND RPV remove heat from the pessure AND BE RCS) pressure CANNOT BE RCS) MAINTAINED below the Heat Capacity Temperature Limit (Figure -HCTL)107 of 110 SSES EAL Comparison Matrix SSES NEI IC Wording IC#(s) SSES IC Wording Difference/Deviation Justification Loss of all Vital DC power for 15 SS2 Loss of all vital DC power for 15 None minutes or longer. minutes or longer.MODE: Power Operation, MODE: I -Power Operation, 2 -Startup, Hot Standby, Hot Startup, 3 -Hot Shutdown Shutdown NEI Ex. NEI Example EAL Wording EA SSES EAL Wording Difference/Deviation Justification EAL # EAL #<105VDC105 VDC on 1 D612 (2D612), 1 D622 (2D622), 1 D632 (2D632) and 1 Indicated voltage is less than SS2.1 Indicated voltage is < 105 VDC 10 DonI62(D1),162(D2),163(D3)ad Indiate votag isles thn SS.1 ndiate votag is1D642 (2D642) is the site-specific minimum vital DC bus design (site-specific bus voltage value) on all of the following vital 125 voltage.on ALL (site-specific Vital DC VDC main distribution buses on voltage.busses) for 15 minutes or longer, the affected unit for > 15 min. Clarified that the plant design of indicated voltage for the vital 125 (Note 1): VDC main distribution buses is local only. Local voltage indication is available for each bus based on dispatching a field operator in S1 D61 2 (2D61 2) accordance with Control Room alarm response procedure AR-S1 D622 (2D622) 1(2)06-001 (A12,B12,C12,D12). Field observation of indicated voltage constitutes the point in time when availability of indications to S1 D632 (2D632) plant operators that an emergency action level has been, or may be, S1 D642 (2D642) exceeded.Note The Emergency Director should N/A Note 1:The ED/RM should The classification timeliness note has been standardized across the declare the Unusual Event declare the event promptly upon SSES EAL scheme by referencing the "time limit" specified within the promptly upon determining that determining that time limit has EAL wording.15 minutes has been exceeded, been exceeded, or will likely be or will likely be exceeded. exceeded.108 of 110 SSES EAL Comparison Matrix SSES NEI IC# NEI IC Wording IC#(s) SSES IC Wording Difference/Deviation Justification SG1 Prolonged loss of all offsite and SGla Prolonged loss of all offsite and "essential" is the site-specific designation for the SSES emergency all onsite AC power to all onsite AC power to essential AC buses.emergency buses. buses Combined NEI ICs SG1 and SG8 under the loss of power category MODE: Power Operation, MODE: 1 -Power Operation, 2 -for usability. Startup, Hot Standby, Hot Startup, 3 -Hot Shutdown Shutdown NEI Ex. SSES EA E NEI Example EAL Wording SL SSES EAL Wording Difference/Deviation Justification EAL # EAL #1 a.Loss of ALL offsite and ALL SG1.1 Loss of all offsite and all onsite ESS buses are the site-specific emergency buses.onsite AC power to (site-specific AC power capability to all 4.16 Added "on EITHER unit" to clarify applicability to either unit.emergency buses). kV ESS buses on EITHER unit 4 hours is the site-specific SBO coping analysis time.AND AND Indication of an inability to adequately remove heat from the core b.EITHER of the following: EITHER: occurs when RPV water level cannot be restored and maintained SRestoration of at least Restoration of at least one above -179 in., which is the EOP RPV water level indicative of a one AC emergency bus in 4.16 kV ESS bus in < 4 hours loss of adequate core cooling.less than (site-specific is not likely (Note 1)hours) is not likely. OR* (Site-specific indication of RPV water level CANNOT an inability to adequately BE RESTORED AND remove heat from the MAINTAINED core) > -179 in.Note The Emergency Director should N/A Note 1: The ED/RM should The classification timeliness note has been standardized across the declare the General Emergency declare the event SSES EAL scheme by referencing the "time limit" specified within promptly upon determining that promptly .upon the EAL wording.(site-specific hours) has been determining that time exceeded, or will likely be limit has been exceeded. exceeded, or will likely be exceeded.109 of 110 SSES EAL Comparison Matrix SSES NEI IC# NEI IC Wording IC#(s) SSES IC Wording Difference/Deviation Justification SG8 Loss of all AC and Vital DC SG1b Loss of all essential AC and "essential" is the site-specific designation for the SSES emergency power sources for 15 minutes or vital DC power sources for 15 AC buses.longer. minutes or longer. Combined NEI ICs SG1 and SG8 under the loss of power category MODE: Power Operation, MODE: 1 -Power Operation, 2 for usability. Startup, Hot Standby, Hot -Startup, 3 -Hot Shutdown Shutdown NEI Ex. SSES EAL # NEI Example EAL Wording EAL# SSES EAL.Wording Difference/Deviation Justification 1 a.Loss of ALL offsite and ALL SG1.2 Loss of all offsite and all onsite ESS buses are the site-specific emergency buses.onsite AC power to (site-specific AC power capability to all 4.16 Added "on EITHER unit" to clarify applicability to either unit.emergency buses) for 15 minutes kV ESS buses on EITHER unit or longer. for > 15 min. 105 VDC on 1D612 (2D612), 1D622 (2D622), 1 D632 (2D632) and 1 D642 (2D642) is the site-specific minimum vital DC bus design AND AND voltage.b.lndicated voltage is less than Indicated voltage is < 105 VDC Clarified that the plant design of indicated voltage for the vital 125 (site-specific bus voltage value) on all of the following vital 125 VDC main distribution buses is local only. Local voltage indication on ALL (site-specific Vital DC VDC main distribution buses is available for each bus based on dispatching a field operator in busses) for 15 minutes or longer, on the affected unit for > 15 accordance with Control Room alarm response procedure AR-min. (Note 1): 1(2)06-001 (A12,B12,C12,D12). Field observation of indicated* 1D612 (2D612) voltage constitutes the point in time when availability of indications to plant operators that an emergency action level has been, or may S1 D622 (2D622) be, exceeded.S1 D632 (2D632)1D642 (2D642)Note The Emergency Director should N/A Note 1: The ED/RM should The classification timeliness note has been standardized across the declare the Unusual Event declare the event SSES EAL scheme by referencing the "time limit" specified within promptly upon determining that promptly upon the EAL wording.15 minutes has been exceeded, determining that time or will likely be exceeded. limit has been exceeded, or will likely be exceeded.110 of 110 ATTACHMENT 3 TO PLA-7285 PPL SUSQUEHANNA, LLC EAL WALLCHARTS

ATTACHMENT 4 TO PLA-7285 PPL SUSQUEHANNA, LLC EAL CLASSIFICATION BASES EP-RM-004 (Mark-ups and a Clean Copy) PROCEDURE COVER SHEET SUSQUEHANNA, LLC PROCEDURE 02/13/2015 EAL CLASSIFICATION BASES (MARK UP OF NEI 99-01 REV. 6) EP-RM-004 Revision [X]Page 1 of 295 ADHERENCE LEVEL: INFORMATION USE QUALITY CLASSIFICATION: APPROVAL CLASSIFICATION: (X) QAProgram ( ) Non-QA Program (X) Plant ( ) Non-Plant ( ) Instruction EFFECTIVE DATE: PERIODIC REVIEW FREQUENCY: 2 Year PERIODIC REVIEW DUE DATE: REWL X0200 RECOMMENDED REVIEWS: All Procedure Owner: Emergency Planning Responsible Supervisor: Manager-EP Responsible FUM: Manager-EP Responsible Approver: Manager-NRA EP-RM-004 Revision [X]Page 2 of 295 TABLE OF CONTENTS SECTION PAGE 1.0 PURPOSE 3 2.0 DISCUSSION 3 2.1 Background 3 2.2 Fission Product Barriers 4 2.3 Fission Product Barrier Classification Criteria 4 2.4 EAL Organization 5 2.5 Technical Bases Information 7 2.6 Operating Mode Applicability 8 3.0 GUIDANCE ON MAKING EMERGENCY CLASSIFICATIONS 9 3.1 General Considerations 9 3.2 Classification Methodology 10

4.0 REFERENCES

13 4.1 Developmental 13 4.2 Implementing 13 5.0 DEINITIONS, ACRONYMS & ABBREVIATIONS 14 6.0 Susquehanna TO NEI 99-01 Rev. 6 EAL CROSS-REFERENCE 22 7.0 ATTACHMENTS 26 1 EAL Bases 27 Cateqory R Abnormal Rad Release / Rad Effluent 28 Categqory C Cold Shutdown / Refueling System Malfunction 75 Category H Hazards 123 Category S System Malfunction 168 Category E ISFSI 214 Category F Fission Product Barrier Degradation 217 2 Fission Product Barrier Loss / Potential Loss Matrix and Bases 222 3 Safe Shutdown Room/Areas Tables R-2 & H-2 Bases 287 4 Table R -Abnormal Rad Levels / Rad Effluents (Form EP-RM-004-R) 290 5 Table C -Cold Shutdown/Refueling System Malfunctions (Form EP-RM-004-C) 291 6 Table H -Hazards (Form EP-RM-004-H) 292 7 Table S -System Malfunctions (Form EP-RM-004-S) 293 8 Table E -ISFSI (Form EP-RM-004-E) 294 9 Table F -Fission Product Barrier Degradation (Form EP-RM-004-F) 295 EP-RM-004 Revision [X]Page 3 of 295 CAUTION This Reference Manual shall follow the process described in NDAP-QA-0004, Procedure Change Process, for subsequent revisions AND NOT the change process described under EP-112, Emergency Plan Reference Manual Program. It must be maintained in accordance with 10 CFR50.54(q). 1.0 PURPOSE This document provides an explanation and rationale for each Emergency Action Level (EAL)included in the EAL Upgrade Project for Susquehanna, LLC. It should be used to facilitate review of the Susquehanna EALs and provide historical documentation for future reference. Decision-makers responsible for implementation of EP-PS-1 00, Emergency Director Control Room (ref. 4.2.1), EP-PS-101, TSC Emergency Director (ref. 4.2.2) and EP-PS-200, Recovery Manager (ref. 4.2.3), may use this document as a technical reference in support of EAL interpretation. This information may assist the Emergency Director/Recovery Manager in making classifications, particularly those involving judgment or multiple events. The bases information may also be useful in training and for explaining event classifications to off-site officials. The expectation is that emergency classifications are to be made as soon as conditions are present and recognizable for the classification, but within 15 minutes or less in all cases of conditions present. Use of this document for assistance is not intended to delay the emergency classification.

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 Susquehanna Emergency Plan.In 1992, the NRC endorsed NUMARC/NESP-007 "Methodology for Development of Emergency Action Levels" as an alternative to NUREG-0654 EAL guidance.NEI 99-01 (NUMARC/NESP-007) Revisions 4 and 5 were subsequently issued for industry implementation. Enhancements over earlier revisions included:* Consolidating the system malfunction initiating conditions and example emergency action levels which address conditions that may be postulated to occur during plant shutdown conditions.

Initiating conditions and example emergency action levels that fully address conditions that may be postulated to occur at permanently Defueled Stations and Independent Spent Fuel Storage Installations (ISFSls).* 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 ML1 10240324) (ref. EP-RM-004 Revision [X]Page 4 of 295 4.1.1), Susquehanna conducted an EAL implementation upgrade project that produced the EALs discussed herein.2.2 Fission Product Barriers Fission product barrier thresholds represent threats to the defense in depth design concept that precludes the release of radioactive fission products to the environment. This concept relies on multiple physical barriers, any one of which, if maintained intact, precludes the release of significant amounts of radioactive fission products to the environment. Many of the EALs derived from the NEI methodology are fission product barrier threshold based. That is, the conditions that define the EALs are based upon thresholds that represent the loss or potential loss of one or more of the three fission product barriers. "Loss" and "Potential Loss" signify the relative damage and threat of damage to the barrier. A "Loss" threshold means the barrier no longer assures containment of radioactive materials. A "Potential Loss" threshold implies an increased probability of barrier loss and decreased certainty of maintaining the barrier.The primary fission product barriers are: A. Fuel Clad (FC): The Fuel Clad Barrier consists of the cladding material that contains the fuel pellets.B. Reactor Coolant System (RCS): The RCS Barrier is the reactor coolant system pressure boundary and includes the Reactor Pressure Vessel (RPV) and all reactor coolant system piping up to and including the isolation valves.C. Primary Containment (PC): The Primary Containment Barrier includes the drywell, the wetwell, their respective interconnecting paths, and other connections up to and including the outermost containment isolation valves. Primary Containment Barrier thresholds are used as criteria for escalation of the Emergency Classification Level (ECL) from Alert to a Site Area Emergency or a General Emergency. 2.3 Fission Product Barrier Classification Criteria The following criteria are the bases for event classification related to fission product barrier loss or potential loss: Alert: Any loss or any potential loss of either Fuel Clad or RCS barrier Site Area Emergency: Loss or potential loss of any two barriers General Emergency: Loss of any two barriers and loss or potential loss of the third barrier EP-RM-004 Revision [X]Page 5 of 295 24 EAL Organization The Susquehanna EAL scheme includes the following features:* Division of the EAL set into three broad groups: o EALs applicable under all plant operating modes -This group would be reviewed by the EAL-user any time emergency classification is considered. o EALs applicable only under hot operating modes -This group would only be reviewed by the EAL-user when the plant is in Hot Shutdown, 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 Susquehanna EAL categories are aligned to and represent the NEI 99-01"Recognition Categories." Subcategories are used in the Susquehanna scheme as necessary to further divide the EALs of a category into logical sets of possible emergency classification thresholds. The Susquehanna EAL categories and subcategories are listed in Table 2.4-1.The primary tool for determining the emergency classification level is the EAL Classification Matrix. The bases provide the EAL user with the background and justification behind the EAL threshold values identified using the guidance set forth in NEI 99-01 Revision 6. If there is any doubt with regard to the applicability of any EAL, the technical basis should be reviewed. The user should consult Section 3.0 and Attachments 1 (EAL technical bases) & 2 (fission product barrier technical bases) of this document for such information. EP-RM-004 Revision [X]Page 6 of 295 Table 2.4-1 EAL Groups, Categories and Subcategories l Group/Category EAL Subcategory All Operating Mode: R -Abnormal Rad Levels / Rad Effluent H -Hazards and Other Conditions Affecting Plant Safety E -Independent Spent Fuel Storage Installation (ISFSI)1 -Radiological Effluent 2 -Irradiated Fuel Event 3 -Area Radiation Levels 1 -Security 2 -Seismic Event 3 -Natural or Technological Hazard 4 -Fire 5 -Hazardous Gas 6 -Control Room Evacuation 7- ED/RM Judgment 1 -Confinement Boundary Hot Conditions: S -System Malfunction F -Fission Product Barrier Degradation 1 -Loss of Essential AC Power 2 -Loss of Vital DC Power 3 -Loss of Control Room Indications 4 -RCS Activity 5 -RCS Leakage 6 -RPS Failure 7 -Loss of Communications 8 -Hazardous Event Affecting Safety Systems None Cold Conditions: C -Cold Shutdown / Refueling System 1 -RPV Level Malfunction 2 -Loss of Essential AC Power 3 -Loss of Vital DC Power 4 -RCS Temperature 5 -Loss of Communications 6 -Hazardous Event Affecting Safety Systems EP-RM-004 Revision [X]Page 7 of 295 2.5 Technical Bases Information EAL technical bases are provided in Attachment 1 for each EAL according to EAL group (All, Hot, Cold), EAL category (R, C, H, S, F and E) and EAL subcategory. A summary explanation of each category and subcategory is given at the beginning of the technical bases discussions of the EALs included in the category. For each EAL, the following information is provided: Category Letter & Title Subcategory Number & Title Initiating Condition (IC)Site-specific description of the generic IC given in NEI 99-01 Rev. 6.EAL Identifier (enclosed in rectangle) Each EAL is assigned a unique identifier to support accurate communication of the emergency classification to onsite and offsite personnel. Four characters define each EAL identifier:

1. First character (letter):

Corresponds to the EAL category as described above (R, C, H, S, F or E)2. Second character (letter): The emergency classification (G, S, A or U)G = General Emergency S = Site Area Emergency A = Alert U Unusual Event 3. Third character (number): Subcategory number within the given category.Subcategories are sequentially numbered beginning with the number one (1). If a category does not have a subcategory, this character is assigned the number one (1).4. Fourth character (number): The numerical sequence of the EAL within the EAL subcategory. If the subcategory has only one EAL, it is given the number one (1).Classification (enclosed in rectangle): Unusual Event (U), Alert (A), Site Area Emergency (S) or General Emergency (G)EAL (enclosed in rectangle) Exact wording of the EAL as it appears in the EAL Classification Matrix Mode Applicability One or more of the following plant operating conditions comprise the mode to which each EAL is applicable: 1 -Power Operations, 2 -Startup, 3 -Hot Shutdown, 4 -Cold Shutdown, 5 -Refueling, D -Defueled, or All. (See Section 2.6 for operating mode definitions) Definitions: If the EAL wording contains a defined term, the definition of the term is included in this section. These definitions can also be found in Section 5.1. Defined terms used in Initiating Condition and Emergency Action Level statements are set in all capital letters (e.g., ALL CAPS). EP-RM-004 Revision [X]Page 8 of 295 Basis: A Plant-Specific basis section that provides Susquehanna-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.Susquehanna Basis Reference(s): Site-specific source documentation from which the EAL is derived 2.6 Operating Mode Applicability (ref. 4.1.2 except Defueled)1 Power Operation Reactor is critical and the mode switch is in RUN 2 Startup The mode switch is in REFUEL (with all reactor vessel head closure bolts fully tensioned) or STARTUP/HOT STANDBY 3 Hot Shutdown The mode switch is in SHUTDOWN (with one or more reactor vessel head closure bolts less than fully tensioned) and average reactor coolant temperature is >200°F 4 Cold Shutdown The mode switch is in SHUTDOWN (with one or more reactor vessel head closure bolts less than fully tensioned) and average reactor coolant temperature is < 200OF 5 Refueling The mode switch is in REFUEL or SHUTDOWN with one or more reactor vessel head closure bolts are less than fully tensioned D Defueled All fuel removed from the reactor vessel (i.e., full core offload during refueling or extended outage) (ref. 4.1.1)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. EP-RM-004 Revision [X]Page 9 of 295 3.0 GUIDANCE ON MAKING EMERGENCY CLASSIFICATIONS 3.1 General Considerations When making an emergency classification, the Emergency Director/Recovery Manager 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. 4.1.3).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. 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 indicator operability, the condition existence, or the report accuracy is removed. Implicit in this definition is the need for timely assessment. 3.1.3 Imminent Conditions For ICs and EALs that have a stipulated time duration (e.g., 15 minutes, 30 minutes, etc.), the Emergency Director/Recovery Manager 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 systemor component. In these cases, the controls associated with the planning, preparation and execution of the work will ensure that compliance is maintained with all aspects of the operating license provided that the activity proceeds and concludes as expected. Events or conditions of this type may be subject to the reporting requirements of 10 § CFR 50.72 (ref. 4.1.4). EP-RM-004 Revision [X]Page 10 of 295 3.1.5 Classification Based on Analysis The assessment of some EALs is based on the results of analyses that are necessary to ascertain whether a specific EAL threshold has been exceeded (e.g., dose assessments, chemistry sampling, RCS leak rate calculation, etc.). For these EALs, the EAL wording or the associated basis discussion will identify the necessary analysis. In these cases, the 15-minute declaration period starts with the availability of the analysis results that show the threshold to be exceeded (i.e., this is the time that the EAL information is first available). The NRC expects licensees to establish the capability to initiate and complete EAL-related analyses within a reasonable period of time (e.g., maintain the necessary expertise on-shift). 3.1.6 Emergency Director/Recovery Manager Judgment While the EALs have been developed to address a full spectrum of possible events and conditions which may warrant emergency classification, a provision for classification based on operator/management experience and judgment is still necessary. The NEI 99-01 EAL scheme provides the Emergency Director/Recovery Manager with the ability to classify events and conditions based upon judgment using EALs that are consistent with the Emergency Classification Level (ECL) definitions (refer to Category H). The Emergency Director/Recovery Manager 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. 4.1.3).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 both 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 both units, an Alert should be declared.Related guidance concerning classification of rapidly escalating events or conditions is provided in Regulatory Issue Summary (RIS) 2007-02, Clarification of NRC Guidance for Emergency Notifications During Quickly Changing Events (ref. 4.1.5). EP-RM-004 Revision [X]Page 11 of 295 3.2.2 Consideration of Mode Changes During Classification The mode in effect at the time that an event or condition occurred, and prior to any plant or operator response, is the mode that determines whether or not an IC is applicable. If an event or condition occurs, and results in a mode change before the emergency is declared, the emergency classification level is still based on the mode that existed at the time that the event or condition was initiated (and not when it was declared). Once a different mode is reached, any new event or condition, not related to the original event or condition, requiring emergency classification should be evaluated against the ICs and EALs applicable to the operating mode at the time of the new event or condition. For events that occur in Cold Shutdown or Refueling, escalation is via EALs that are applicable in the Cold Shutdown or Refueling modes, even if Hot Shutdown (or a higher mode) is entered during the subsequent plant response. In particular, the fission product barrier EALs are applicable only to events that initiate in the Hot Shutdown mode or higher.3.2.3 Classification of Imminent Conditions Although EALs provide specific thresholds, the Emergency Director/Recovery Manager must remain alert to events or conditions that could lead to meeting or exceeding an EAL within a relatively short period of time (i.e., a change in the ECL is IMMINENT). If, in the judgment of the Emergency Director/Recovery Manager, 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. 4.1.5).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 scram the reactor followed by a successful manual scram.3.2.6 Classification of Transient Conditions Many of the ICs and/or EALs employ time-based criteria. These criteria will require that the IC/EAL conditions be present for a defined period of time before an emergency declaration is warranted. In cases where no time-based criterion is specified, it is recognized that some transient conditions may cause an EAL to be met for a brief period of time (e.g., a few seconds to a few minutes). The following guidance should be applied to the classification of these conditions. EP-RM-004 Revision [X]Page 12 of 295 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 Director/Recovery Manager 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. 4.1.6) is applicable. Specifically, the event should be reported to the NRC in accordance with 10 CFR § 50.72 (ref. 4.1.4) within one hour of the discovery of the undeclared event or condition. The licensee should also notify appropriate State and local agencies in accordance with the agreed upon arrangements. 3.2.8 Retraction of an Emergency Declaration Guidance on the retraction of an emergency declaration reported to the NRC is discussed in NUREG-1022 (ref. 4.1.6). EP-RM-004 Revision [X]Page 13 of 295

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 ML1 10240324 4.1.2 Technical Specifications Table 1.1-1 Modes 4.1.3 NSIR/DPR-ISG-01 Interim Staff Guidance, Emergency Planning for Nuclear Power Plants 4.1.4 10 § CFR 50.72 Immediate Notification Requirements for Operating Nuclear Power Reactors 4.1.5 RIS 2007-02 Clarification of NRC Guidance for Emergency Notifications During Quickly Changing Events, February 2, 2007 4.1.6 NUREG-1022 Event Reporting Guidelines: 10CFR50.72 and 50.73 4.1.7 NUH-003 Updated Final Safety Analysis Report for the Standardized NUHOMS Horizontal Modular Storage System for Irradiated Nuclear Fuel, Section 1.3.1 4.1.8 NDAP-QA-0309 Primary Containment Access and Control 4.1.9 NDAP-QA-0321 Secondary Containment Integrity Control 4.1.10 Susquehanna LLC, Susquehanna Steam Electric Station Emergency Plan, Section 1.0, Definitions 4.1.11 10 § CFR 50.73 License Event Report System 4.1.12 ON-FPC-101(201) Loss of Fuel Pool Cooling 4.1.13 EO-000-104 Secondary Containment Control 4.1.14 FSAR Section 3.7a Seismic Design 4.2 Implementing 4.2.1 EP-PS-100 Emergency Director Control Room 4.2.2 EP-PS-101 TSC Emergency Director 4.2.3 EP-PS-200 Recovery Manager 4.2.4 NEI 99-01 Rev. 6 to Susquehanna EAL Comparison Matrix EP-RM-004 Revision [X]Page 14 of 295 5.0 DEFINITIONS, ACRONYMS & ABBREVIATIONS 5.1 Definitions (ref. 4.1.1 except as noted)Selected defined 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.CAN/CANNOT BE MAINTAINED ABOVE/BELOW The value of an identified parameter is/is not able to be held within the specified limit. The determination requires an evaluation of system performance and availability in relation to parameter values and trends. An instruction prescribing action when a parameter cannot be maintained above or below a specified limit neither requires nor prohibits anticipatory action-depending upon plant conditions, the action may be taken as soon as it is determined that the limit will ultimately be exceeded, or delayed until the limit is actually reached. Once the parameter does exceed the limit, however, the action must be performed; it may not be delayed while attempts are made to restore the parameter to within the desired control band.CAN/CANNOT BE RESTORED ABOVE/BELOW The value of an identified parameter is/is not able to be brought within the specified limit. The determination requires an evaluation of system performance and availability in relation to parameter values and trends. An instruction prescribing action when a value CANNOT BE RESTORED AND MAINTAINED above or below a specified limit does not require immediate action simply because the current values is outside the ranqe, but does not permit extended operation beyond the limit: the action must be taken as soon as it is apparent that the specified range cannot be attained.CONFINEMENT BOUNDARY The barrier(s) between spent fuel and the environment once the spent fuel is processed for dry storage. As related to the Susquehanna ISFSI, Confinement Boundary is defined as the Dry Shielded Canister (DSC) (Ref. 4.1.7).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 Susquehanna, CONTAINMENT CLOSURE is established per NDAP-QA-0309 (ref 4.1.8) for Primary Containment OR is established per NDAP-QA-0321 (ref 4.1.9) for Secondary Containment. EMERGENCY PLAN BOUNDARY (EPB) (ref. 4.1.10)Same as the Exclusion Area Boundary, i.e., that area around SSES within a radius of 1800 feet determined in accordance with 10 CFR 100.11. EP-RM-004 Revision [X]Page 15 of 295 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 require Susquehanna to recommend protective actions for the -general public to offsite planning agencies.The dose program complies with the "Manual of Protective Action Guides and Protective Actions for Nuclear Incidents," (EPA-400), adopting the dose calculation methodoloqy I ICRP#26/30. The accident dose assessments are based on the adult physiology per EPA 400, except for one case -that is, child thyroid dose conversion factors are used in calculating thyroid CDE. However adult physiology is used in calculating thyroid CDE for purposes of evaluating the need for a sheltering only PAR and evaluating controlled venting of containment. Calculations of TEDE are made using the (adult) dose factors provided in EPA-400.EXPLOSION A rapid, violent and catastrophic failure of a piece of equipment due to combustion, chemical reaction or overpressurization. A release of steam (from high energy lines or components) or an electrical component failure (caused by short circuits, grounding, arcing, etc.) should not automatically be considered an explosion. Such events require a post-event inspection to determine if the attributes of an explosion are present.Faulted The te~rm applied to a steam generator that has a steamn lcak On the seGondar~y side Of sufficient size to Gaus6e an uncontrolled drop in steam generator pressure or the 6team 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.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.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 Susquehanna 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 Susquehanna. 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). EP-RM-004 Revision [X]Page 16 of 295 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.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.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 measures such as temporary shielding, SCBAs or dose extensions beyond Emergency Plan RWP that are not routinely employed to access the room/area). INTRUSION The act of entering without authorization. Discovery of a bomb in a specified area is indication of intrusion into that area by a hostile force.MAINTAIN Take appropriate action to hold the value of an identified parameter within specified limits.NORMAL LEVELS As applied to radiological IC/EALs, the highest reading in the past twenty-four hours excluding the current peak value.OPERATING BASIS EARTHQUAKE (OBE)An earthquake which, considering the regional and local geology, and seismology and specific characteristics of local subsurface material, could reasonably be expected to affect the plant site during the operating life of the plant. It is that earthquake which produces the vibratory ground motion for which these features of the nuclear power plant necessary for continued operation without undue risk to the health and safety of the public are designed to remain functional (ref.4.1.14).OWNER CONTROLLED AREA (ref. 4.1.10)Includes the area within the expanded security perimeter, i.e., the areas. that are bordered by the Vehicle Barriers System. The OWNER CONTROLLED AREA also includes the Monitored OWNER CONTROLLED AREA (MOCA) as defined in Security Procedures. PROJECTILE An object directed toward a Nuclear Power Plant that could cause concern for its continued operability, reliability, or personnel safety.PROTECTED AREA (ref. 4.1.10)Area within the station inner security fence (PROTECTED AREA Barrier) designated to implement the requirements of 10 CFR 73. EP-RM-004 Revision [X]Page 17 of 295 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).REFUELING PATHWAY The reactor refueling cavity, spent fuel pool and fuel transfer canal comprise the refueling pathway (ref. 4.1.12).Ruptured The condition of a steam generator in which primar,' to seconRdar; leakage is Of sufficient magnitude to require a safety ijcin 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. SAFE SHUTDOWN EARTHQUAKE (SSE)An earthquake which is based upon an evaluation of the maximum earthquake potential considering the regional and local geology, and seismology and specific characteristics of local subsurface material. It is that earthquake which produces the maximum vibratory ground motion for which Seismic Category I systems and components are designed to remain functional (ref 4.1.14).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 BOUNDARY The line beyond which the land is not owned, leased or otherwise controlled by the licensee (Susquehanna drawing C243786, Sh 1, "Site Facilities and Boundary Map.") (ref. 4.1.10). EP-RM-004 Revision [XI Page 18 of 295 UNISOLABLE An open or breached system line that cannot be isolated, remotely or locally.* The term UNISOLABLE also includes any decision by plant staff or procedure direction to not isolate a primary system.* Normal leakage past a closed isolation valve is not considered UNISOLABLE leakage.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.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. EP-RM-004 Revision [X]Page 19 of 295 5.2 Abbreviations/Acronyms OF .................................................................................................................... Degrees Fahrenheit 0 ......................................................................................................................................... D e g re e s AC ..................................................................................................................... Alternating Current AOP .............................................................................................. Abnormal Operating Procedure APRM ................................................................................................ Average Power Range Meter ATW S .................................................................................... Anticipated Transient W ithout Scram BW R ............................................................................................................. Boiling W ater Reactor BW ROG ............................................................................... Boiling W ater Reactor Owners Group CDE ................................................................................................... Com mitted Dose Equivalent CFR .................................................................................................. Code of Federal Regulations CS ................................................................................................................................. Core Spray DBA ....................................................................................... ..... ... Design Basis Accident DC ............................................................................................................................. Direct Current EAL .......................................................................................................... Emergency Action Level ECCS ......................................................................................... Em ergency Core Cooling System ECL ............................................................................................... Em ergency Classification Level ED ................................................................................................................... Em ergency Director EDST ........................................................................................ Evaporator Distillate Sam ple Tank EOF ................................................................................................ Em ergency Operations Facility EOP ........................................................................................... Emergency Operating Procedure EPA ............................................................................................ Environmental Protection Agency EPB ...................................................................................................... Emergency Plan Boundary EPG .....................................................

......................................

Emergency Procedure Guideline EPIP ............................................................................. Emergency Plan Im plem enting Procedure ESF ...................................................................................................... Engineered Safety Feature ESS .............................................................................................. Engineered Safeguards System FAA ............................................................................................... Federal Aviation Adm inistration FBI ................................................................................................ Federal Bureau of Investigation FEMA ............................................................................ Federal Em ergency Management Agency FPRR .............................................................................................. Fire Protection Review Report FSAR ................................................................................................. Final Safety Analysis Report G E ................................................................................................................... General Emergency HCTL .......................................................................................... Heat Capacity Tem perature Limit HPCI ............................................................................................ High Pressure Coolant Injection IC ...................................................................................................................... Initiating Condition IPEEE ............................. Individual Plant Exam ination of External Events (Generic Letter 88-20)ISFSI ......................................................................... Independent Spent Fuel Storage Installation Keff ...................................................................................... Effective Neutron M ultiplication Factor LCO ................................................................................................ Lim iting Condition of Operation EP-RM-004 Revision [X]Page 20 of 295 LER ............................................................................................................. Licensee Event Report LOCA ...................................................................................................... Loss of Coolant Accident LRW ..................................................................................................................... Liquid Radwaste LRW ................................................................................................................ Light W ater Reactor M PC ................................................ M axim um Perm issible Concentration/M ulti-Purpose Canister M PH ........................................................................................................................ M iles Per Hour M SIV ................................................................................................... M ain Steam Isolation Valve M SL ...................................................................................................................... M ain Steam Line m R, m Rem , m rem , m REM ............................................................ m illi-Roentgen Equivalent Man MW .................................................................................................................................. Megawatt NEI ............................................................................................................ Nuclear Energy Institute NESP ................................................................................ National Environmental Studies Project NPP ................................................................................................................ Nuclear Power Plant NRC ............................................................................................ Nuclear Regulatory Com m ission NSSS ............................................................................................. Nuclear Steam Supply System NORAD ................................................................ North American Aerospace Defense Com m and (NO)UE ............................................................................................. Notification of Unusual Event O BE ................................................................................................... Operating Basis Earthquake OCA .......................................................................................................... Owner Controlled Area ODCM/ODAM ..................................................... Off-site Dose Calculation (Assessment) M anual O RO ............................................................................................... Offsite Response Organization PA ........................................................................................................................... Protected Area PRA/PSA .................................. Probabilistic Risk Assessment / Probabilistic Safety Assessm ent PW R ..................................................................................................... Pressurized W ater Reactor PSIG ............................................................................................ Pounds per Square Inch Gauge R ..................................................................................................................................... R o e n tg e n RB ......................................................................................................................... Reactor Building RCIC .............................................................................................. Reactor Core Isolation Cooling RCS ......................................................................................................... Reactor Coolant System Rem , rem , REM .................................................................................... Roentgen Equivalent Man RETS ...................................................................... Radiological Effl uent Technical Specifications RHR .......................................................................................................... Residual Heat Rem oval RM .................................................................................................................... Recovery Manager RPS ..................................................................................................... Reactor Protection System RPV ......................................................................................................... Reactor Pressure Vessel RRC .............................................................................................................. Reactor Recirculation RW CU ....................................................................................................... Reactor W ater Cleanup SAR ............................................................................................................. Safety Analysis Report SBO ...................................................................................................................... Station Blackout SCBA ................................................................................... Self-Contained Breathing Apparatus EP-RM-004 Revision [X]Page 21 of 295 SDHR ..................................................................................... Supplem ental Decay Heat Rem oval SGTS ......................................................................................... Stand-By Gas Treatm ent System SPDS ........................................................................................ Safety Param eter Display System SRO ......................................................................................................... Senior Reactor Operator SSE .................................................................................................... Safe Shutdow n Earthquake SSES ................................................................................... Susquehanna Steam Electric Station SW ........................................................................................................................... Service W ater TEDE ............................................................................................ Total Effective Dose Equivalent TAF .................................................................................................................... Top of Active Fuel TRM ............................................................................................ Technical Requirem ents M anual TSC ........................................................................................................ Technical Support Center EP-RM-004 Revision [X]Page 22 of 295 6.0 Susquehanna-TO-NEI 99-01 Rev. 6 EAL CROSS-REFERENCE This cross-reference is provided to facilitate association and location of a Susquehanna EAL within the NEI 99-01 IC/EAL identification scheme. Further information regarding the development of the Susquehanna EALs based on the NEI guidance can be found in the EAL Comparison Matrix.NEI 99-01 Rev. 6 Susquehanna 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 RA3.2 AA3 2 RSI.1 AS1 1 RS1.2 AS1 2 RS1.3 ASI 3 RS2.1 AS2 1 RG1.1 AG1 1 RG1.2 AG1 2 RG1.3 AGI 3 RG2.1 AG2 1 CU1.1 Cui 1 EP-RM-004 Revision [X]Page 23 of 295 NEI 99-01 Rev. 6 Susquehanna EAL IC Example EAL CU1.2 CUl 2 CU2.1 CU2 1 CU3.1 CU4 1 CU4.1 CU3 1 CU4.2 CU3 2 CU4.1 CU4 1 CU5.1 CU5 1,2,3 CA1.1 CA1 1 CA1.2 CAI 2 CA2.1 CA2 1 CA4.1 CA3 1,2 CA6.1 CA6 1 CS1.1 CS1 1 CS1.2 CSl 2 CS1.3 CSl 3 CG1.1 CG1 1 CG1.2 CG1 2 FA1.1 FA1 1 FS1.1 FSl 1 FG1.1 FG1 1 HUI.1 HUI 1,23 HU2.1 HU2 1 HU3.1 HU3 1 HU3.2 HU3 2 HU3.3 HU3 3 EP-RM-004 Revision [X]Page 24 of 295 NEI 99-01 Rev. 6 Susquehanna EAL IC Example EAL HU3.4 HU3 4 N/A HU3 5 HU4.1 HU4 1 HU4.2 HU4 2 HU4.3 HU4 3 HU4.4 HU4 4 HU7.1 HU7 1 HAl.1 HAI 1,2 HA5.1 HA5 I HA6.1 HA6 1 HA7.1 HA7 1 HSI.1 HSI I HS6.1 HS6 1 HS7.1 HS7 1 HGI.1 HG1 1 HG7.1 HG7 1 SuI.1 Sul 1 SU3.1 SU2 1 SU4.1 SU3 1 SU4.2 SU3 2 SU5.1 SU4 1,2,3 SU6.1 SU5 1 SU6.2 SU5 2 SU7.1 SU6 1,2,3 SAI.1 SAl 1 EP-RM-004 Revision [X]Page 25 of 295 NEI 99-01 Rev. 6 Susquehanna EAL IC Example EAL SA3.1 SA2 1 SA6.1 SA5 1 SA8.1 SA9 1 SS1.1 SS1 1 SS2.1 SS8 1 SS6.1 SS5 1 SG1.1 SG1 1 SG1.2 SG8 1 EUI.1 E-HU1 1 EP-RM-004 Revision [X]Page 26 of 295 7.0 ATTACHMENTS 7.1 Attachment 1, Emergency Action Level Technical Bases 7.2 Attachment 2, Fission Product Barrier Matrix and Bases 7.3 Attachment 3, Safe Shutdown Room/Areas Tables R-2 & H-2 Bases 7.4 Attachment 4, Table R -Abnormal Rad Levels / Rad Effluents (Form EP-RM-004-R) 7.5 Attachment 5, Table E -ISFSI (Form EP-RM-004-E) 7.6 Attachment 6, Table H -Hazards (Form EP-RM-004-H) 7.7 Attachment 7, Table S -System Malfunctions (Form EP-RM-004-S) 7.8 Attachment 8, Table F -Fission Product Barrier Degradation (Form EP-RM-004-F) 7.9 Attachment 9, Table C -Cold Shutdown/Refueling System Malfunctions (Form EP-RM-004-C) Attachment 1 EP-RM-004 Revision [X]Page 27 of 295 ATTACHMENT 1 Emergency Action Level Technical Bases Cateciory 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 the 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. Radiologqical 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.

Page 1 of 195 Attachment 1 EP-RM-004 Revision [X]Page 28 of 295 Category: Subcategory: Initiating Condition: EAL: R -Abnormal Rad Levels / Rad Effluent 1 -Radiological Effluent Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem child thyroid CDE RGI.1 General Emergency Gaseous effluent > Table R-1 column "GE" for -> 15 min. (Notes 1, 2, 3, 4)Note 1: The ED/RM 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 RAI.1, RSI .1 and RGI.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 (Note 4)Release Point Monitor [ GE SAE Alert UE Plant Vent 00630 1.9E+09 1.9E+08 1.9E+07 4.0E+06 0 pCi/min pCi/min pCi/min pCi/min (site total) (site total) (site total) (site total)LRW RR-06433 ............- 2 x hi alarm RR-D12-1(2) SW/SDHR 1(2)R606 ---- 2 x hi alarm 4 Mode Applicability: All Definition(s): None Page 2 of 195 Attachment 1 EP-RM-004 Revision [X]Page 29 of 295 Susquehanna Basis: This EAL address gaseous radioactivity releases, that for whatever reason, cause effluent radiation monitor readings corresponding to EMERGENCY PLAN BOUNDARY doses that exceed either (ref. 1):* 1000 mrem TEDE* 5000 mrem CDE Child Thyroid The SITE BOUNDARY is irregularly shaped and therefore would result in dose projections and protective action recommendations that can vary significantly depending on plume direction and affected sector. Using dose projections calculated using the EPB (i.e. EXCLUSION AREA)provides a more consistent approach to Public Protective Action Recommendations since the EPB is more consistently defined in all directions. The EPB is at or within the SITE BOUNDARY in all compass sectors.The EPA PAGs are expressed in terms of the projected sum of the effective dose equivalent (EDE) from external radiation and the committed effective dose equivalent (CEDE) incurred from inhalation of radioactive materials, or as the committed dose equivalent (CDE) to the thyroid. For the purpose of these IC/EALs, the projected dose quantity Total Effective Dose Equivalent (TEDE), as defined in 10 CFR 20 is used in lieu of "the projected sum of... EDE...and CEDE...", with CEDE considering significant dose from inhaled radionuclides during the early phase of the event. The EPA protective action guidance provides for the use of adult thyroid dose conversion factors. However, the Commonwealth of Pennsylvania requires the use of child thyroid CDE for purposes of comparison of projected thyroid CDE to the PAG for thyroid CDE.The monitor reading threshold for RGI.1 was determined as described in ref. 1.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 child Thyroid). For multi-release point gaseous releases, classification should be based on dose assessment that considers the total site release rate.The SPING monitors the radioactive effluent from the Units 1 and 2 Turbine Building and Reactor Building Ventilation Stacks and the Standby Gas Treatment System Exhaust Vent. All five collectively are the Plant Vent on Table R-1. The SPING system is normally aligned to be operated from Brother Control Terminals 0C630 in the Control Room, using 0C677 as a backup in the TSC. Three Post-Accident Vent Stack Sampling Systems (PAVSSS) have been installed as backup to the SPING Units. They are used following an accident involving fuel degradation if the SPING monitoring capabilities are lost. Control Terminal CT-1 with System Operator Console CT-1 B interrogates each of the SPING and PAVSSS Radiation Monitors for particulate, iodine, and noble gases and informs the operator of changes in operational status, alarm condition, or system parameters within seconds of their occurrence. (ref. 2)NEI 99-01 Basis: This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude will require implementation of protective actions for the public.Page 3 of 195 Attachment 1 EP-RM-004 Revision [X]Page 30 of 295 I)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.Susquehanna Basis Reference(s):

1. NEP Technical Basis 02-005 Rev. #2 Noble Gas Release Rate Limits for EALs 2. FSAR 18.1.30 Accident-Monitoring Instrumentation

3. NEI 99-01 AG1 Page 4 of 195 Attachment 1 EP-RM-004 Revision [X]Page 31 of 295 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 child thyroid CDE EAL: RG1.2 General Emergency Dose assessment using actual meteorology indicates doses > 1,000 mrem TEDE or 5,000 mrem child thyroid CDE at or beyond the EMERGENCY PLAN BOUNDARY (Notes 3, 4)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, RSI.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): EMERGENCY PLAN BOUNDARY (EPB) -Same as the Exclusion Area Boundary, i.e., that area around SSES within a radius of 1800 feet determined in accordance with 10 CFR 100.11 Susquehanna Basis: Dose assessments are performed by computer-based methods (ref. 2).The EMERGENCY PLAN BOUNDARY (EPB) is used in assessing dose effects to the public rather than the SITE BOUNDARY. The EPB is at or within the SITE BOUNDARY in all compass sectors. The SSES dose projection model (MIDAS) utilizes the EPB when performing dose calculations (ref. 5).The SITE BOUNDARY is irregularly shaped and therefore would result in dose projections and protective action recommendations that can vary significantly depending on plume direction and affected sector. Using dose projections calculated using the EPB (i.e. Exclusion Area) provides a more consistent approach to Public Protective Action Recommendations since the EPB is more consistently defined in all directions. The EPB is at or within the SITE BOUNDARY in all compass sectors.The EPA PAGs are expressed in terms of the projected sum of the effective dose equivalent (EDE) from external radiation and the committed effective dose equivalent (CEDE) incurred from inhalation of radioactive materials, or as the committed dose equivalent (CDE) to the thyroid. For the purpose of these IC/EALs, the projected dose quantity Total Effective Dose Equivalent (TEDE), as defined in 10 CFR 20 is used in lieu of "the projected sum of... EDE...and CEDE...", with CEDE considering significant dose from inhaled radionuclides during the early Page 5 of 195 Attachment 1 EP-RM-004 Revision [X]Page 32 of 295 phase of the event. The EPA protective action guidance provides for the use of adult thyroid dose conversion factors. However, the Commonwealth of Pennsylvania requires the use of child thyroid CDE for purposes of comparison of projected thyroid CDE to the PAG for thyroid CDE.Since dose assessment is based on actual meteorology, whereas the monitor reading RG 1.1 is not, the results from these assessments may indicate that the classification is not warranted. For this reason, emergency implementing procedures call for the timely performance of dose assessments using actual meteorology and release information. If the results of these dose assessments are available when the classification is made (e.g., initiated at a lower classification level), the dose assessment results override the monitor reading RGI.1. Classification should not be delayed pending the results of these dose assessments. NEI 99-01 Basis: This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude will require implementation of protective actions for the public.Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. The TEDE dose is set at the EPA PAG of 1,000 mrem while the 5,000 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.Susquehanna Basis Reference(s):

1. Susquehanna LLC, Susquehanna Steam Electric Stations Emergency Plan, Section 7.1.1, Off Site Dose Calculations

2. EP-RM-005 SSES MIDAS-NU User Manual 3. NEI 99-01 AG1 Page 6 of 195 Attachment 1 EP-RM-004 Revision [X]Page 33 of 295 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 child thyroid CDE EAL: RG1.3 General Emergency Field survey results indicate EITHER of the following at or beyond the EMERGENCY PLAN BOUNDARY: " Closed window dose rates > 1,000 mR/hr expected to continue for 60 min.* Analyses of field survey samples indicate child thyroid CDE > 5,000 mrem for 60 min. of inhalation.(Notes 1, 2)Note 1: The ED/RM 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): EMERGENCY PLAN BOUNDARY (EPB) -Same as the Exclusion Area Boundary, i.e., that area around SSES within a radius of 1800 feet determined in accordance with 10 CFR 100.11 Susquehanna Basis: The EMERGENCY PLAN BOUNDARY (EPB) is used in assessing dose effects to the public rather than the SITE BOUNDARY. The EPB is at or within the SITE BOUNDARY in all compass sectors. The SSES dose projection model (MIDAS) utilizes the EPB when performing dose calculations (ref. 1).The SITE BOUNDARY is irregularly shaped and therefore would result in dose projections and protective action recommendations that can vary significantly depending on plume direction and affected sector. Using dose projections calculated using the EPB (i.e. EXCLUSION AREA)provides a more consistent approach to Public Protective Action Recommendations since the EPB is more consistently defined in all directions. The EPB is at or within the SITE BOUNDARY in all compass sectors.The EPA PAGs are expressed in terms of the projected sum of the effective dose equivalent (EDE) from external radiation and the committed effective dose equivalent (CEDE) incurred from inhalation of radioactive materials, or as the committed dose equivalent (CDE) to the thyroid. For the purpose of these IC/EALs, the projected dose quantity Total Effective Dose Page 7 of 195 Attachment 1 EP-RM-004 Revision [X]Page 34 of 295 Equivalent (TEDE), as defined in 10 CFR 20 is used in lieu of "the projected sum of...EDE...and CEDE...", with CEDE considering significant dose from inhaled radionuclides during the early phase of the event. The EPA protective action guidance provides for the use of adult thyroid dose conversion factors. However, the Commonwealth of Pennsylvania requires the use of child thyroid CDE for purposes of comparison of projected thyroid CDE to the PAG for thyroid CDE.NEI 99-01 Basis: This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude will require implementation of protective actions for the public.Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. The TEDE dose is set at the EPA PAG of 1,000 mrem while the 5,000 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.Classification based on effluent monGitor readings assumes that a rclcase path to the eirnent is established. if thc effluent flow past an effluen-Pt mnonitor is known to have stopped due to actions to isolate the release path, then the efflu ent monitor reading is no loger9 VAlid forF cla;s~sific.a;tion-purpaoses. Susquehanna Basis Reference(s):

1. EP-RM-005 SSES MIDAS-NU User Manual 2. NEI 99-01 AG1 Page 8 of 195 Attachment 1 EP-RM-004 Revision [X]Page 35 of 295 Category: Subcategory:

Initiating Condition: EAL: R -Abnormal Rad Levels / Rad Effluent 1 -Radiological Effluent Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem child thyroid CDE RSI.1 Site Area Emergency Gaseous effluent > Table R-1 column "SAE" for > 15 min. (Notes 1, 2, 3, 4)Note 1: The ED/RM 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 RAI.1, RSI.1 and RGI.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 (Note 4)Release Point Monitor GE SAE Alert UE U)Plant Vent 0C630 1.9E+09 1.9E+08 1.9E+07 4.OE+06 0 pCi/min pCi/min pCi/min pCi/min (noble gas) 0C677 (site total) (site total) (site total) (site total)LRW RR-06433 ---- ----.... 2 x hi alarm RR-D12-1(2) RHRSW A/B ---- ----... 2 x hi alarm o- 1 (2)R606 1(2) SW/SDHR RD1(2)R0 ---- ----.. 2 x hi alarm Mode Applicability: All Definition(s): None Page 9 of 195 Attachment 1 EP-RM-004 Revision [X]Page 36 of 295 Susquehanna Basis: This EAL address gaseous radioactivity releases, that for whatever reason, cause effluent radiation monitor readings corresponding to EMERGENCY PLAN BOUNDARY (EPB) doses that exceed either (ref. 1):* 100 mremTEDE* 500 mrem CDE Child Thyroid The SITE BOUNDARY is irregularly shaped and therefore would result in dose projections and protective action recommendations that can vary significantly depending on plume direction and affected sector. Using dose projections calculated using the EPB (i.e. EXCLUSION AREA)provides a more consistent approach to Public Protective Action Recommendations since the EPB is more consistently defined in all directions. The EPB is at or within the SITE BOUNDARY in all compass sectors.The EPA PAGs are expressed in terms of the projected sum of the effective dose equivalent (EDE) from external radiation and the committed effective dose equivalent (CEDE) incurred from inhalation of radioactive materials, or as the committed dose equivalent (CDE) to the thyroid. For the purpose of these IC/EALs, the projected dose quantity Total Effective Dose Equivalent (TEDE), as defined in 10 CFR 20 is used in lieu of "the projected sum of... EDE... and CEDE...", with CEDE considering significant dose from inhaled radionuclides during the early phase of the event. The EPA protective action guidance provides for the use of adult thyroid dose conversion factors. However, the Commonwealth of Pennsylvania requires the use of child thyroid CDE for purposes of comparison of projected thyroid CDE to the PAG for thyroid CDE.The monitor reading threshold for RSI.1 was determined as described in ref. 1.The column "SAE" gaseous effluent release values in Table R-1 correspond to calculated doses of 10% of the EPA Protective Action Guidelines (TEDE or CDE child Thyroid). For multi-release point gaseous releases, classification should be based on dose assessment that considers the total site release rate.The SPING monitors the radioactive effluent from the Units 1 and 2 Turbine Building and Reactor Building Ventilation Stacks and the Standby Gas Treatment System Exhaust Vent. All five collectively are the Plant Vent on Table R-1. The SPING system is normally aligned to be operated from Brother Control Terminals 0C630 in the Control Room, using 0C677 as a backup in the TSC. Three Post-Accident Vent Stack Sampling Systems (PAVSSS) have been installed as backup to the SPING Units. They are used following an accident involving fuel degradation if the SPING monitoring capabilities are lost. Control Terminal CT-1 with System Operator Console CT-1 B interrogates each of the SPING and PAVSSS Radiation Monitors for particulate, iodine, and noble gases and informs the operator of changes in operational status, alarm condition, or system parameters within seconds of their occurrence. (ref. 2)NEI 99-01 Basis: This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to 10% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public.Page 10 of 195 Attachment 1 EP-RM-004 Revision [X]Page 37 of 295 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 AG4RG 1.Susquehanna Basis Reference(s):

1. NEP Technical Basis 02-005 Rev. #2 Noble Gas Release Rate Limits for EALs 2. FSAR 18.1.30 Accident-Monitoring Instrumentation

3. NEI 99-01 AS1 Page 11 of 195 Attachment 1 EP-RM-004 Revision [X]Page 38 of 295 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 child thyroid CDE EAL: RS1.2 Site Area Emergency Dose assessment using actual meteorology indicates doses > 100 mrem TEDE or 500 mrem child thyroid CDE at or beyond the EMERGENCY PLAN BOUNDARY (Notes 3, 4)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 RAI.1, RSI.1 and RGI.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): EMERGENCY PLAN BOUNDARY (EPB) -Same as the Exclusion Area Boundary, i.e., that area around SSES within a radius of 1800 feet determined in accordance with 10 CFR 100.11 Susquehanna Basis: Dose assessments are performed by computer-based methods (ref. 2).The EMERGENCY PLAN BOUNDARY (EPB) is used in assessing dose effects to the public rather than the SITE BOUNDARY. The EPB is at or within the SITE BOUNDARY in all compass sectors. The SSES dose projection model (MIDAS) utilizes the EPB when performing dose calculations (ref. 5).The SITE BOUNDARY is irregularly shaped and therefore would result in dose projections and protective action recommendations that can vary significantly depending on plume direction and affected sector. Using dose projections calculated using the EPB (i.e. Exclusion Area) provides a more consistent approach to Public Protective Action Recommendations since the EPB is more consistently defined in all directions. The EPB is at or within the SITE BOUNDARY in all compass sectors.The EPA PAGs are expressed in terms of the projected sum of the effective dose equivalent (EDE) from external radiation and the committed effective dose equivalent (CEDE) incurred from inhalation of radioactive materials, or as the committed dose equivalent (CDE) to the thyroid. For the purpose of these IC/EALs, the projected dose quantity Total Effective Dose Equivalent (TEDE), as defined in 10 CFR 20 is used in lieu of "the projected sum of...EDE...and CEDE...", with CEDE considering significant dose from inhaled radionuclides during the early Page 12 of 195 Attachment 1 EP-RM-004 Revision [X]Page 39 of 295 phase of the event. The EPA protective action guidance provides for the use of adult thyroid dose conversion factors. However, the Commonwealth of Pennsylvania requires the use of child thyroid CDE for purposes of comparison of projected thyroid CDE to the PAG for thyroid CDE.Since dose assessment is based on actual meteorology, whereas the monitor reading RS1.1 is not, the results from these assessments may indicate that the classification is not warranted. For this reason, emergency implementing procedures call for the timely performance of dose assessments using actual meteorology and release information. If the results of these dose assessments are available when the classification is made (e.g., initiated at a lower classification level), the dose assessment results override the monitor reading RSI.1. Classification should not be delayed pending the results of these dose assessments. NEI 99-01 Basis: This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to 10% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public.Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. The TEDE dose is set at 10% of the EPA PAG of 1,000 mrem while the 500 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.Escalation of the emergency classification level would be via IC AG4RG1.Susquehanna Basis Reference(s):

1. Susquehanna LLC, Susquehanna Steam Electric Stations Emergency Plan, Section 7.1.1, Off Site Dose Calculations

2. EP-RM-005 SSES MIDAS-NU User Manual 3. NEI 99-01 AS1 Page 13 of 195 Attachment 1 EP-RM-004 Revision [X]Page 40 of 295 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 child thyroid CDE EAL: RS1.3 Site Area Emergency Field survey results indicate EITHER of the following at or beyond the EMERGENCY PLAN BOUNDARY: " Closed window dose rates> 100 mR/hr expected to continue for -> 60 min." Analyses of field survey samples indicate child thyroid CDE > 500 mrem for 60 min. of inhalation.(Notes 1, 2)Note 1: The ED/RM 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): EMERGENCY PLAN BOUNDARY (EPB) -Same as the Exclusion Area Boundary, i.e., that area around SSES within a radius of 1800 feet determined in accordance with 10 CFR 100.11.Susquehanna Basis: The EMERGENCY PLAN BOUNDARY (EPB) is used in assessing dose effects to the public rather than the SITE BOUNDARY. The EPB is at or within the SITE BOUNDARY in all compass sectors. The SSES dose projection model (MIDAS) utilizes the EPB when performing dose calculations (ref. 1).The SITE BOUNDARY is irregularly shaped and therefore would result in dose projections and protective action recommendations that can vary significantly depending on plume direction and affected sector. Using dose projections calculated using the EPB (i.e. EXCLUSION AREA)provides a more consistent approach to Public Protective Action Recommendations since the EPB is more consistently defined in all directions. The EPB is at or within the SITE BOUNDARY in all compass sectors The EPA PAGs are expressed in terms of the projected sum of the effective dose equivalent (EDE) from external radiation and the committed effective dose equivalent (CEDE) incurred from inhalation of radioactive materials, or as the committed dose equivalent (CDE) to the thyroid. For the purpose of these IC/EALs, the projected dose quantity Total Effective Dose Equivalent (TEDE), as defined in 10 CFR 20 is used in lieu of "the projected sum of...EDE...and Page 14 of 195 Attachment 1 EP-RM-004 Revision [X]Page 41 of 295 CEDE...", with CEDE considering significant dose from inhaled radionuclides during the early phase of the event. The EPA protective action guidance provides for the use of adult thyroid dose conversion factors. However, the Commonwealth of Pennsylvania requires the use of child thyroid CDE for purposes of comparison of projected thyroid CDE to the PAG for thyroid CDE.Since dose assessment is based on actual meteorology, whereas the monitor reading RSI. 1 is not, the results from these assessments may indicate that the classification is not warranted. For this reason, emergency implementing procedures call for the timely performance of dose assessments using actual meteorology and release information. If the results of these dose assessments are available when the classification is made (e.g., initiated at a lower classification level), the dose assessment results override the monitor reading RSI.1. Classification should not be delayed pending the results of these dose assessments. NEI 99-01 Basis: This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to 10% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public.Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. The TEDE dose is set at 10% of the EPA PAG of 1,000 mrem while the 500 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.ClaSSifiation based on effluent monRitor readings assumes that a release path to h eirnent is established. if the effluent flow past an effluent monitor is known to have stopped due to action to isolate the release path, then the effluent monRitor reading is no longer Valid for classification purposes-. Escalation of the emergency classification level would be via IC AG4RG1.Susquehanna Basis Reference(s):

1. EP-RM-005 SSES MIDAS-NU User Manual 2. NEI 99-01 AS1 Page 15 of 195 Attachment 1 EP-RM-004 Revision [X]Page 42 of 295 Category: Subcategory:

Initiating Condition: R -Abnormal Rad Levels / Rad Effluent 1 -Radiological Effluent Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem child thyroid CDE EAL: RAI.1 Alert Gaseous effluent > Table R-1 column "Alert" for > 15 min. (Notes 1, 2, 3, 4)Note 1: The ED/RM 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 RAI.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 (Note 4)Release Point Monitor GE SAE Alert UE oPlant Vent 00630 1.9E+09 1.9E+08 1.9E+07 4.OE+06 0 pCi/min pCi/min pCi/min pCi/min (site total) (site total) (site total) (site total)LRW RR-06433 ---- ---- ---- 2 x hi alarm RR-D12-1(2) RHRSW A/B ............ 2 x hi alarm o- 1 (2)R606 1(2) SW/SDHR D1(2)R0 .......-- 2 x hi alarm Mode Applicability: All Definition(s): None Page 16 of 195 Attachment 1 EP-RM-004 Revision [X]Page 43 of 295 Susquehanna Basis: This EAL address gaseous radioactivity releases, that for whatever reason, cause effluent radiation monitor readings corresponding to EMERGENCY PLAN BOUNDARY (EPB) doses that exceed either (ref. 1):* 10mremTEDE

50 mrem CDE Child Thyroid The SITE BOUNDARY is irregularly shaped and therefore would result in dose projections and protective action recommendations that can vary significantly depending on plume direction and affected sector. Using dose projections calculated using the EPB (i.e. Exclusion Area) provides a more consistent approach to Public Protective Action Recommendations since the EPB is more consistently defined in all directions.

The EPB is at or within the SITE BOUNDARY in all compass sectors.The EPA PAGs are expressed in terms of the projected sum of the effective dose equivalent (EDE) from external radiation and the committed effective dose equivalent (CEDE) incurred from inhalation of radioactive materials, or as the committed dose equivalent (CDE) to the thyroid. For the purpose of these IC/EALs, the projected dose quantity Total Effective Dose Equivalent (TEDE), as defined in 10 CFR 20 is used in lieu of "the projected sum of...EDE...and CEDE...", with CEDE considering significant dose from inhaled radionuclides during the early phase of the event. The EPA protective action guidance provides for the use of adult thyroid dose conversion factors. However, the Commonwealth of Pennsylvania requires the use of child thyroid CDE for purposes of comparison of projected thyroid CDE to the PAG for thyroid CDE.The monitor reading threshold for RA1.1 was determined as described in ref. 1.The column "ALERT" gaseous effluent release value 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). For multi-release point gaseous releases, classification should be based on dose assessment that considers the total site release rate.The SPING monitors the radioactive effluent from the Units 1 and 2 Turbine Building and Reactor Building Ventilation Stacks and the Standby Gas Treatment System Exhaust Vent. All five collectively are the Plant Vent on Table R-1. The SPING system is normally aligned to be operated from Brother Control Terminals 0C630 in the Control Room, using 0C677 as a backup in the TSC. Three Post-Accident Vent Stack Sampling Systems (PAVSSS) have been installed as backup to the SPING Units. They are used following an accident involving fuel degradation if the SPING monitoring capabilities are lost. Control Terminal CT-1 with System Operator Console CT-1B interrogates each of the SPING and PAVSSS Radiation Monitors for particulate, iodine, and noble gases and informs the operator of changes in operational status, alarm condition, or system parameters within seconds of their occurrence. (ref. 2)NEI 99-01 Basis: This IC addresses a release of gaseous or liquid radioactivity that results in projected or actual offsite doses greater than or equal to 1% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude represent an actual or potential substantial degradation of the level of safety of the plant as indicated by a radiological release that significantly exceeds regulatory limits (e.g., a significant uncontrolled Page 17 of 195 Attachment 1 EP-RM-004 Revision [X]Page 44 of 295 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 AS4RS1.Susquehanna Basis Reference(s):

1. NEP Technical Basis 02-005 Rev. #2 Noble Gas Release Rate Limits for EALs 2. FSAR 18.1.30 Accident-Monitoring Instrumentation

4. NEI 99-01 AA1 Page 18 of 195 Attachment 1 EP-RM-004 Revision [X]Page 45 of 295 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 child thyroid CDE EAL: RA1.2 Alert Dose assessment using actual meteorology indicates doses > 10 mrem TEDE or 50 mrem child thyroid CDE at or beyond the EMERGENCY PLAN BOUNDARY (Notes 3, 4)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 RAI.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): EMERGENCY PLAN BOUNDARY (EPB) -Same as the Exclusion Area Boundary, i.e., that area around SSES within a radius of 1800 feet determined in accordance with 10 CFR 100.11.Susquehanna Basis: Dose assessments are performed by computer-based methods (ref. 2).The EMERGENCY PLAN BOUNDARY (EPB) is used in assessing dose effects to the public rather than the SITE BOUNDARY. The EPB is at or within the SITE BOUNDARY in all compass sectors. The SSES dose projection model (MIDAS) utilizes the EPB when performing dose calculations (ref. 5).The SITE BOUNDARY is irregularly shaped and therefore would result in dose projections and protective action recommendations that can vary significantly depending on plume direction and affected sector. Using dose projections calculated using the EPB (i.e. Exclusion Area) provides a more consistent approach to Public Protective Action Recommendations since the EPB is more consistently defined in all directions. The EPB is at or within the SITE BOUNDARY in all compass sectors.The EPA PAGs are expressed in terms of the projected sum of the effective dose equivalent (EDE) from external radiation and the committed effective dose equivalent (CEDE) incurred from inhalation of radioactive materials, or as the committed dose equivalent (CDE) to the thyroid. For the purpose of these IC/EALs, the projected dose quantity Total Effective Dose Equivalent (TEDE), as defined in 10 CFR 20 is used in lieu of "the projected sum of...EDE...and CEDE...", with CEDE considering significant dose from inhaled radionuclides during the early phase of the event. The EPA protective action guidance provides for the use of adult thyroid Page 19 of 195 Attachment 1 EP-RM-004 Revision [X]Page 46 of 295 dose conversion factors. However, the Commonwealth of Pennsylvania requires the use of child thyroid CDE for purposes of comparison of projected thyroid CDE to the PAG for thyroid CDE.Since dose assessment is based on actual meteorology, whereas the monitor reading RA1.1 is not, the results from these assessments may indicate that the classification is not warranted. For this reason, emergency implementing procedures call for the timely performance of dose assessments using actual meteorology and release information. If the results of these dose assessments are available when the classification is made (e.g., initiated at a lower classification level), the dose assessment results override the monitor reading RAI.1. Classification should not be delayed pending the results of these dose assessments. NEI 99-01 Basis: This IC addresses a release of gaseous or liquid radioactivity that results in projected or actual offsite doses greater than or equal to 1% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releasesof 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 AS4RS1.Susquehanna Basis Reference(s):

1. Susquehanna LLC, Susquehanna Steam Electric Stations Emergency Plan, Section 7.1.1, Off Site Dose Calculations

2. EP-RM-005 SSES MIDAS-NU User Manual 3. NEI 99-01 AA1 Page 20 of 195 Attachment 1 EP-RM-004 Revision [X]Page 47 of 295 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 child 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 child thyroid CDE at or beyond the EMERGENCY PLAN BOUNDARY for 60 min. of exposure (Notes 1, 2)Note 1: The ED/RM 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): EMERGENCY PLAN BOUNDARY (EPB) -Same as the Exclusion Area Boundary, i.e., that area around SSES within a radius of 1800 feet determined in accordance with 10 CFR 100.11.Susquehanna Basis: RA1.3 addresses uncontrolled releases that are detected by sample analyses, particularly on unmonitored pathways, e.g., spills of radioactive liquids into storm drains, heat exchanger leakage in river water systems, etc.For a radiological liquid release, the calculated effluent concentration from a chemistry sample is compared to the emergency action level. Shift Management utilizes emergency response procedures to notify risk counties and to obtain river water samples.The Susquehanna station incorporates features intended to control the release of radioactive effluents to the environment. Further, there are administrative controls established to prevent unintentional releases, or control and monitor intentional releases. These controls are located in the Technical Requirements Manual (TRM). The occurrence of extended, uncontrolled radioactive releases to the environment is indicative of degradation in these features and/or controls.EAL RA1.3 addresses uncontrolled releases that are detected by sample analyses, particularly on unmonitored pathways, e.g., spills of radioactive liquids into storm drains, heat exchanger leakage in river water systems, etc. This EAL reflects the concern that releases in excess of the referenced offsite dose values represent an uncontrolled situation and hence a potential degradation in the level of safety. Although the calculated dose is very low, it is the degradation in plant control as indicated by the failure to terminate the release that is of primary concern.Page 21 of 195 Attachment 1 EP-RM-004 Revision [X]Page 48 of 295 The EMERGENCY PLAN BOUNDARY is referenced in EAL RA1.3 because this EAL is based upon liquid release limits from the plant. The release limits are contained in the TRM and are in turn based upon calculation methodology specified in the ODCM. The ODCM utilizes the EMERGENCY PLAN BOUNDARY to establish plant release limits.NEI 99-01 Basis: This IC addresses a release of gaseous or liquid radioactivity that results in projected or actual offsite doses greater than or equal to 1% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude represent an actual or potential substantial degradation of the level of safety of the plant as indicated by a radiological release that significantly exceeds regulatory limits (e.g., a significant uncontrolled release).Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or -appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. The TEDE dose is set at 1% of the EPA PAG of 1,000 mrem while the 50 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.ClaGsification based on effluent monRitor Feadings assumes that a release path to the environent is established. if the effluent flow past an effluent moenitor is known to have stopped due to actions to isolate the release path, then the effluent moneitor reading is no longer valid for claGsifiGation purpEscalation of the emergency classification level would be via IC AS4RS1.Susquehanna Basis Reference(s):

1. ODCM-QA-003 Effluent Monitor Setpoints 2. NEI 99-01 AA1 Page 22 of 195 Attachment 1 EP-RM-004 Revision [X]Page 49 of 295 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 child thyroid CDE EAL: RA1.4 Alert Field survey results indicate EITHER of the following at or beyond the EMERGENCY PLAN BOUNDARY: " Closed windowdose rates> 10 mR/hr expected to continue for > 60 min." Analyses of field survey samples indicate child thyroid CDE > 50 mrem for 60 min. of inhalation.(Notes 1,2)Note 1: The ED/RM 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): EMERGENCY PLAN BOUNDARY,(EPB) -Same as the Exclusion Area Boundary, i.e., that area around SSES within a radius of 1800 feet determined in accordance with 10 CFR 100.11 Susquehanna Basis: The EMERGENCY PLAN BOUNDARY (EPB) is used in assessing dose effects to the public rather than the SITE BOUNDARY. The EPB is at or within the SITE BOUNDARY in all compass sectors. The SSES dose projection model (MIDAS) utilizes the EPB when performing dose calculations (ref. 1).The SITE BOUNDARY is irregularly shaped and therefore would result in dose projections and protective action recommendations that can vary significantly depending on plume direction and affected sector. Using dose projections calculated using the EPB (i.e. Exclusion Area) provides a more consistent approach to Public Protective Action Recommendations since the EPB is more consistently defined in all directions. The EPB is at or within the SITE BOUNDARY in all compass sectors.The EPA PAGs are expressed in terms of the projected sum of the effective dose equivalent (EDE) from external radiation and the committed effective dose equivalent (CEDE) incurred from inhalation of radioactive materials, or as the committed dose equivalent (CDE) to the thyroid. For the purpose of these IC/EALs, the projected dose quantity Total Effective Dose Page 23 of 195 Attachment 1 EP-RM-004 Revision [X]Page 50 of 295 Equivalent (TEDE), as defined in 10 CFR 20 is used in lieu of "the projected sum of...EDE...and CEDE...", with CEDE considering significant dose from inhaled radionuclides during the early phase of the event. The EPA protective action guidance provides for the use of adult thyroid dose conversion factors. However, the Commonwealth of Pennsylvania requires the use of child thyroid CDE for purposes of comparison of projected thyroid CDE to the PAG for thyroid CDE.NEI 99-01 Basis: This IC addresses a release of gaseous or liquid radioactivity that results in projected or actual offsite doses greater than or equal to 1% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude represent an actual or potential substantial degradation of the level of safety of the plant as indicated by a radiological release that significantly exceeds regulatory limits (e.g., a significant uncontrolled release).Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. The TEDE dose is set at 1% of the EPA PAG of 1,000 mrem while the 50 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.Escalation of the emergency classification level would be via IC ASI-RS1.Susquehanna Basis Reference(s):

1. EP-RM-005 SSES MIDAS-NU User Manual 2. NEI 99-01 AA1 Page 24 of 195 Attachment 1 EP-RM-004 Revision [X]Page 51 of 295 Category: Subcategory:

Initiating Condition: R -Abnormal Rad Levels / Rad Effluent 1 -Radiological Effluent Release of gaseous or liquid radioactivity greater than 2 times the TRM limits for 60 minutes or longer EAL: RUI.1 Unusual Event Gaseous or liquid effluent > Table R-1 column "UE" for -> 60 min. (Notes 1, 2, 3)Note 1: The ED/RM 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 1 Monitor [ GE SAE Alert UE Plant Vent 00630 1.9E+09 1.9E+08 1.9E+07 4.OE+06 4 pCi/min pCi/min pCi/min pCi/min (noble gas) 0C677 (site total) (site total) (site total) (site total)LRW RR-06433 ---- 2 x hi alarm RR-D12-" 1(2) RHRSW A/B ............ 2 x hi alarm Cr 1(2)R606 RR-D12-1(2) SW/SDHR 1(2)R604 ----.... 2 x hi alarm Mode Applicability: All Definition(s): None Susquehanna Basis: The EMERGENCY PLAN BOUNDARY is used in RUI.1 because it is based upon release limits from the plant. The release limits are based upon the Plant Technical Requirements Manual and in turn based upon calculation methodology specified in the ODCM. The ODCM utilizes the Page 25 of 195 Attachment 1 EP-RM-004 Revision [X]Page 52 of 295 EMERGENCY PLAN BOUNDARY to establish plant release limits. EAL RUI.1 is an indication of degradation in the level of safety of the plant.This EAL represents radioactivity releases, that for whatever reason, cause liquid effluent radiation monitor readings to exceed two times the Technical Requirements Manual limit and releases are not terminated within 60 minutes. This alarm setpoint may be associated with a planned batch release, or a continuous release path. In either case, the setpoint is established by the ODCM to warn of a release that is not in compliance with the applicable TRM release limit. Indexing the EAL threshold to the ODCM setpoints in this manner ensures that the EAL threshold will never be less than the setpoint established by a specific discharge permit.For a radiological liquid release, the calculated effluent concentration from a chemistry sample is compared to the emergency action level. Shift Management utilizes emergency response procedures to notify risk counties and to obtain river water samples.Gaseous Releases The column "UE" gaseous release value in Table R-1 represents two times the appropriate TRM release rate limits (ref. 1-3).The SPING monitors the radioactive effluent from the Units 1 and 2 Turbine Building and Reactor Building Ventilation Stacks and the Standby Gas Treatment System Exhaust Vent. All five collectively are the Plant Vent on Table R-1. The SPING system is normally aligned to be operated from Brother Control Terminals 0C630 in the Control Room, using 0C677 as a backup in the TSC. Three Post-Accident Vent Stack Sampling Systems (PAVSSS) have been installed as backup to the SPING Units. They are used following an accident involving fuel degradation if the SPING monitoring capabilities are lost. Control Terminal CT-1 with System Operator Console CT-1B interrogates each of the SPING and PAVSSS Radiation Monitors for particulate, iodine, and noble gases and informs the operator of changes in operational status, alarm condition, or system parameters within seconds of their occurrence. (ref. 4)Liquid Releases The column "UE" liquid release values in Table R-1 represent two times the appropriate TRM release rate limits associated with the specified release point (ref. 5-9).LRW A sample pump takes a portion of the LRW effluent line flow and passes it through a scintillation detector (RE-06433). The contents of EDST, the LRW sample tanks, or the laundry drain sample tank can be lined up for release. Release flow is through two isolation valves, HV-06432A1 and A2 to the cooling tower blowdown line.The isolation valves will close on: " Low Sample Flow (< 0.5 gpm)* Low Blowdown Flow (< 5500 gpm)* High radiation (calculated for each release based on sample)* Rad Monitor downscale (calculated)" Rad Monitor inoperable Page 26 of 195 Attachment 1 EP-RM-004 Revision [X]Page 53 of 295" RHRSW The Residual Heat Removal (RHR) Service Water RMS detects primary coolant leakage into the RHR Service Water during RHR Heat Exchanger operation. Local indication of RHR Service Water Loop'A' radiation is provided by RITS-1 1216A and RHR Service Water Loop 'B' by RITS-11216B. Output is also sent to Control Room alarms and Radiation Recorder RR-D12-1R606 on Panel 1C600. The following RHR Service Water RMS high radiation annunciators are located on Control Room Panel 1C601: RHR SWA HI RADIATION (AR-109-F01) is actuated at a setpoint determined by Chemistry." RHR SW B HI RADIATION (AR-113-F01) is actuated at a setpoint determined by Chemistry." SW/SDHR Service Water/Supplemental Decay Heat Removal RMS detects radioactive material in-leakage to the service water system from the spent fuel pool heat exchangers. During unit outages when SDHR is placed in service, the SDHR Detector is connected to the Radiation Monitoring Unit in place of the Service Water Detector.The Service Water/Supplemental Decay Heat Removal RMS annunciator SERVICE WATER EFFLUENT HI RADIATION (AR-123-D04) setpoint is variable as determined by Chemistry Group.NEI 99-01 Basis: This IC addresses a potential decrease in the level of safety of the plant as indicated by a low-level radiological release that exceeds regulatory commitments for an extended period of time (e.g., an uncontrolled release). It includes any gaseous or liquid radiological release, monitored or un-monitored, including those for which a radioactivity discharge permit is normally prepared.Nuclear power plants incorporate design features intended to control the release of radioactive effluents to the environment. Further, there are administrative controls established to prevent unintentional releases, and to control and monitor intentional releases. The occurrence of an extended, uncontrolled radioactive release to the environment is indicative of degradation in these features and/or controls.Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. 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.EAL-#-l This EAL addresses normally occurring continuous radioactivity releases from monitored gaseous or liquid effluent pathways. =AL #2 This EAL addresses radioactivity releasce that cause effluent radiation monitor r,4eadings to eXceed 2 timer, the liit es,--tabshed Page 27 of 195 Attachment 1 EP-RM-004 Revision [X]Page 54 of 295 by a radioactiVity dfischarge permit. This EAL= will typically be aSSo dated with planned bac releases from non continuous release pathways (e.g., radwaste, waste gas).-EAL= #3 This EAL addressos uncontrolled gaseous or liquid releases that are detected byT sample analyses or evRironmental on unmnRitored pathways (e.gi., spills o radioactiVe liquids into storm drains, heat exchanger leakage in river water systems, ecEAL Bases Escalation of the emergency classification level would be via IC AAI-RAI.Susquehanna Basis Reference(s):

1. TRM 3.3.4 TRM Post-Accident Monitoring Instrumentation

2. TRM 3.11.2 Gaseous Effluents 3. Susquehanna Calculation EC ENVR 1041 Airborne Effluent Limiting Site Release Rate &Plant Vent Effluent Monitor Setpoints 4. FSAR 18.1.30 Accident-Monitoring Instrumentation

5. OP-179-002 Process Radiation Monitoring System 6. FSAR 9.2 Water Systems 7. TRM 3.11.1.4 Liquid Radwaste Effluent Monitoring Instrumentation

8. TRM 3.11.1.5 Radioactive Liquid Process Monitoring Instrumentation

9. ON-069-001 Abnormal Rad Release Liquid 10. SSES Offsite Dose Calculation Manual 11. NEI 99-01 AU1 Page 28 of 195 Attachment 1 EP-RM-004 Revision [X]Page 55 of 295 Category:

R -Abnormal Rad Levels / Rad Effluent Subcategory: 1 -Radiological Effluent Initiating Condition: Release of gaseous or liquid radioactivity greater than 2 times the TRM 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 TRM limits for > 60 min. (Notes 1, 2)Note 1: The ED/RM 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 Susquehanna Basis: For a radiological liquid release, the calculated effluent concentration from a chemistry sample is compared to the emergency action level. Shift Management utilizes emergency response procedures to notify risk counties and to obtain river water samples.Limits associated with liquid and gaseous radioactive effluents are contained in the Technical Requirements Manual (TRM). The methodology for calculation of offsite dose or release rates to ensure compliance with the applicable TRM limits is outlined in the ODCM.EAL RU1.2 addresses radioactivity releases, that for whatever reason, cause effluent radiation monitor readings to exceed two times the applicable TRM limit and are not terminated within 60 minutes. The effluent monitor alarm setpoints are established by the ODCM to warn of a release that is not in compliance with the applicable TRM limit (ref.1). The ODCM establishes a methodology for determining effluent radiation monitor setpoints. The ODCM specifies default source terms and, for gaseous releases, prescribes the use of pre-determined annual average meteorology in the most limiting downwind sector for showing compliance with the regulatory commitments. The fundamental basis of this IC is NOT a dose or dose rate, but rather the degradation in the level of safety of the plant implied by the uncontrolled release EAL RU1.2 includes any release for which a radioactivity discharge permit was not prepared or not applicable, or a release that exceeds the conditions (e.g., minimum dilution flow, maximum discharge flow, alarm setpoints, etc.) on the applicable permit. Although the calculated dose is very low, it is the degradation in plant control as indicated by the failure to terminate the release that is of primary concern.Page 29 of 195 Attachment 1 EP-RM-004 Revision [XI Page 56 of 295 NEI 99-01 Basis: This IC addresses a potential decrease in the level of safety of the plant as indicated by a low-level radiological release that exceeds regulatory commitments for an extended period of time (e.g., an uncontrolled release). It includes any gaseous or liquid radiological release, monitored or un-monitored, including those for which a radioactivity discharge permit is normally prepared.Nuclear power plants incorporate design features intended to control the release of radioactive effluents to the environment. Further, there are administrative controls established to prevent unintentional releases, and to control and monitor intentional releases. The occurrence of an extended, uncontrolled radioactive release to the environment is indicative of degradation in these features and/or controls.Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. Classq-ific-ation based on effluent mon)itor readings assumes that a release path to the-eirnent 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 classifiation pur~poses-. Releases should not be prorated or averaged. For example, a release exceeding 4 times release limits for 30 minutes does not meet the EAL. EAL #01 This EAL addresses normally occurr~ing continuouIs radioactivity releases from monitored gaseous or liquid effluent pathways.E.AL 92 -This EAL. addresses radioactivity releases that cause effluent radiation monitor readings to exceed 2 times the limit established by a radioactivity discharge permit. This EAL will typically be associated with planned batch releases from non continuous release pathways (e.g., radwaste, waste gas).EAL #3 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 AA4RAI.Susquehanna Basis Reference(s):

1. ODCM-QA-003 Effluent Monitor Setpoints 2. NEI 99-01 AU1 Page 30 of 195 Attachment 1 EP-RM-004 Revision [X]Page 57 of 295 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 spent fuel racks for 60 minutes or longer EAL: RG2.1 General Emergency Spent fuel pool level CANNOT BE RESTORED to at least 0.5 ft. above the top of the spent fuel racks for > 60 min. (Note 1)Note 1: The ED/RM should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Mode Applicability: All Definition(s): None Susquehanna Basis: Each fuel storage pool is designed to maintain the water level in the pool above the top of active fuel providing cooling for the fuel rods. Technical Specifications require greater than or equal to 22 ft. of water be maintained over the top of irradiated fuel assemblies seated in the spent fuel storage racks at all times (ref. 1). In the event of loss of fuel pool inventory, Operations will assess multiple indications in accordance with ON-FPC-101(201) and AOP-081-001 (ref. 3, 4).Post-Fukushima order EA-12-051 required the installation of reliable SFP level indication capable of identifying normal SPF level 22.75 feet above the top of the fuel racks (Level 1 or el.817'-1 "), SFP level 10 ft. above the top of the fuel racks (Level 2 or el. 804'-4") and SFP level at the top of the fuel racks (Level 3 or el. 794'-10" which for SSES is 0.5 feet on instrument above the top of the fuel racks). Each spent fuel pool is equipped with primary and backup guided wave radar probes to measure pool level. The range is continuous from the high pool level elevation (23.25 feet on instrument or el. 817'-7") to the top of the spent fuel racks (0.0 feet on instrument or el. 794'-4") (ref. 2).Each new SFP LI system provides two alarms to the associated Unit's Control Room Benchboard. The first alarm identifies a low water level condition at 20 ft. (instrument) above the spent fuel rack (elevation 814'-4"). The second alarm identifies a low low water level condition at 10 ft. (instrument) above the spent fuel rack (Level 2 or elevation 804'-4"). These alarms are intended to alert the control room operators of the loss of SFP inventory so actions are taken to provide make-up as soon as possible (ref. 2).Page 31 of 195 Attachment 1 EP-RM-004 Revision [X]Page 58 of 295 NEI 99-01 Basis: This IC 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. Susquehanna Basis Reference(s): I. Technical Specifications 3.7.7 Spent Fuel Storage Pool Water Level 2. PLA-6980 Enclosure 1 Susquehanna Units 1 & 2 Overall Integrated Plan with Regard to Reliable Spent Fuel Pool Instrumentation

3. ON-FPC-101(201)

Loss of Fuel Pool Cooling 4. AOP-081-001 Fuel Handling Abnormal Operating Procedure 5. NEI 99-01 AG2 Page 32 of 195 Attachment 1 EP-RM-004 Revision [XI Page 59 of 295 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 < 0.5 ft. above the top of the spent fuel racks Mode Applicability: All Definition(s): None Susquehanna Basis: Each fuel storage pool is designed to maintain the water level in the pool above the top of active fuel providing cooling for the fuel rods. Technical Specifications require greater than or equal to 22 ft. of water be maintained over the top of irradiated fuel assemblies seated in the spent fuel storage racks at all times (ref. 1). In the event of loss of fuel pool inventory, Operations will assess multiple indications in accordance with ON-FPC-1 01 (201) and AOP-081-001 (ref. 3, 4).Post-Fukushima order EA-12-051 required the installation of reliable SFP level indication capable of identifying normal SPF level 22.75 feet above the top of the fuel racks (Level 1 or el.817'-1"), SFP level 10 ft. above the top of the fuel racks (Level 2 or el. 804'-4") and SFP level at the top of the fuel racks (Level 3 or el. 794'-10" which for SSES is 0.5 feet on instrument above the top of the fuel racks). Each spent fuel pool is equipped with primary and backup guided wave radar probes to measure pool level. The range is continuous from the high pool level elevation (23.25 feet on instrument or el. 817'-7") to the top of the spent fuel racks (0.0 feet on instrument or el. 794'-4") (ref. 2).Each new SFP LI system provides two alarms to the associated Unit's Control Room Benchboard. The first alarm identifies a low water level condition at 20 ft. (instrument) above the spent fuel rack (elevation 814'-4"). The second alarm identifies a low low water level condition at 10 ft. (instrument) above the spent fuel rack (Level 2 or elevation 804'-4"). These alarms are intended to alert the control room operators of the loss of SFP inventory so actions are taken to provide make-up as soon as possible (ref. 2).NEI 99-01 Basis: This IC 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 AG-1-RG1 or AG2RG2.Page 33 of 195 Attachment 1 EP-RM-004 Revision [X]Page 60 of 295 EAL tT-his EAL cecalates from AU2 in that the loss, of level, in the affected portion of the REFUEI -NGATHWAY, is of sufficient .agnitude to have resulted in uncover.. Of irradiated fuel. Indications of irradiated fuel uncover; mnay include direct or indirect visual obsengation (e.g., reports from personn~el Or caeaiaes), as well as significant changes in water and radiation levels, or other plant paarameteirs.. Computational aids may also be used (e g., a boil Off curve). Classification of an event using thi-s E=AL soul be based on the totalit' of available indicatfions, reports and observations. WXhile an area radiation monitor could detect an inrGease in a dose rate due to a lowering of water level in some portion of the REFUELING PATHWAY, the reading mnay not be a reliable indication of whether oe not the fuel rs actually unrcvered. To the degree possible, readings should be considered in combination ether available iRdicatieRs of invento,'- loss.A drop in water level above irradiated fuel within the reactor vessel may be classified accordance Recognition Categwr, C during the Cold Shutdown and Refueling modes.Susquehanna Basis Reference(s):

1. Technical Specifications 3.7.7 Spent Fuel Storage Pool Water Level 2. PLA-6980 Enclosure 1 Susquehanna Units 1 & 2 Overall Integrated Plan with Regard to Reliable Spent Fuel Pool Instrumentation

3. ON-FPC-101(201)

Loss of Fuel Pool Cooling 4. AOP-081-001 Fuel Handling Abnormal Operating Procedure 5. NEI 99-01 AA2;f'Page 34 of 195 Attachment 1 EP-RM-004 Revision [X]Page 61 of 295 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.Susquehanna Basis: This EAL applies to all instances of irradiated fuel handling, including those that are not directly in support of a reactor refueling outage. (e.g. unit-to-unit fuel shuffles, dry fuel storage canister loading, etc.).Each fuel storage pool is designed to maintain the water level in the pool above the top of active fuel providing cooling for the fuel rods. Technical Specifications require greater than or equal to 22 feet of water be maintained over the top of irradiated fuel assemblies seated in the spent fuel storage racks at all times (ref. 1). In the event of loss of fuel pool inventory, Operations will assess multiple indications in accordance with ON-FPC-101(201) and AOP-081-001 (ref. 3, 4).NEI 99-01 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 (See nek!p6F Net-es. These events present radiological safety challenges to plant personnel and are precursors to a release of radioactivity to the environment. As such, they represent an actual or potential substantial degradation of the level of safety of the plant.This IC applies t fuel that iS lieRned for d,' 6torage uip to tho point that the loaded storage cask is sealed. Once sealed, damage to a loaded cask causing loSs Of the CONFINEMVENT BOUNDARY is classified in accorFdance with IC E= HIJI Escalation of the emergencY would be based on either Recg0nition Categor; A or C IQ&.EAL44 This EAL escalates from AU2-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. Page 35 of 195 Attachment 1 EP-RM-004 Revision [X]Page 62 of 295 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.EAL--#2 This EAL addresses a release of radioactive mnaterial caused by mechanical damage to irradiated fuel. Damnaging events may iRclude 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 obse"atio. S of a potential fuel dama event (e.g., a fuel handling accident)-. EAL#3 Spent fuel pel, water level at this value is within the Iower end- of the level range reGessar, to prevent significant dose consequences from direct gamnma radiation to personnel pertorming operations in the vicinity of the spent fuel pool. This condition reflects a sign;ificant loss of spent fuel pool water invento,' and thus it is also a precursor to a loss of the abilit' to adequately cool the irradiated fuel assembles stored in the pool.Escalation of the emergency classification level would be via ICe A&S--RS1 9eAS2-(see-Susquehanna Basis Reference(s):

1. Technical Specifications 3.7.7 Spent Fuel Storage Pool Water Level 2. PLA-6980 Enclosure I Susquehanna Units 1 & 2 Overall Integrated Plan with Regard to Reliable Spent Fuel Pool Instrumentation

3. ON-FPC-101(201)

Loss of Fuel Pool Cooling 4. AOP-081-001 Fuel Handling Abnormal Operating Procedure 5. NEI 99-01 AA2 Page 36 of 195 Attachment 1 EP-RM-004 Revision [X]Page 63 of 295 Category: R -Abnormal Rad Levels / Rad Effluent Subcategory: 2 -Irradiated Fuel Event Initiating Condition: Significant lowering of water level above, or damage to, irradiated fuel EAL: RA2.2 Alert Damage to irradiated fuel resulting in a release of radioactivity AND Any of the following radiation monitor indications:

Refuel Floor High Exhaust (> 18 mR/hr)" Refuel Floor Wall Exhaust (> 21 mR/hr)" Channel 14 Spent Fuel Pool Area Criticality Monitor (> 100 mR/hr)* Channel 15 Refueling Floor Area (> 80 mR/hr)* Channel 42 Refueling Floor Area (> 80 mR/hr)* Channel 47 (Ul) /44 (U2) Spent Fuel Pool Area Criticality Monitor (> 100 mR/hr)" Channel 49 Refueling Floor High Range Monitor (on scale)Mode Applicability:

All Definition(s): None Susquehanna Basis: The Reactor Building ventilation process monitoring system isolates Zone 3 HVAC on high exhaust radiation. Zone 3 exhaust can be monitored at 1C600 (2C600) (ref. 4):* RR D12 1R605 (2R605), Refuel Floor Wall Exhaust Radiation Monitor* RR D12 1R607 (2R607), Refuel Floor High Exhaust Radiation Monitor The listed radiation monitors and specified alarm setpoints/ indications are those associated with a fuel handling accident or damaged spent fuel (ref. 1, 3, 5).NEI 99-01 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-(see-Devei) A -ete). 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.Page 37 of 195 Attachment 1 EP-RM-004 Revision [X]Page 64 of 295 This IC applies to irraddiatted fu-lel that is, licensed for dr; storage up to the point that the loaded storage cask is scaled. Once sealed, damage to a loaded cask causing loss of the CONFINEMENT BOUNDARY is classified in accordance with IC E HUI.Escalation of the emergency would be based on either Recognition Category A-R_or C ICs.EAL #This EAL escalates from AU2 On that the loss of level, inthe affected porti.onof the REFUE~L ING P 'A THWAY, is Of sufficient magnitude to have reasulted-in uncoever,' of irradiated fuel Inicatonsof irradiated fuel uncGover; may include dierect or ind-irecAt visa osrvto (e.g., reports from personnel Or camnera images), as well as significant changes in water and radiatfion levels, or other plant parameters. Comnputational aids may also be uised (e.g., a boil Off curve). Classification of an event using this , EALh should be based on the totality of available*indications, reports and observations. While an; area radiation moniRtor could detect an inrGease in a dose rate due to a lowering of Water level in oRf the REFUELING PATHWAY, the reading may not be a reliable indicr0ation of whether o not the fuel is a ,tually uncovered. To the degree possible, reading should be considered in combination with other a;;aiable indications of inventor' loss.A drop in water level above irradiated fuel within the reactorvessel May be classife in accor.Pr d-Ance Recognition Categor,' C duOrig the Cold Shutdown.F andr- Refueling moedes.This EAL addresses a release of radioactive material caused by mechanical damage to irradiated fuel. Damaging events may include the dropping, bumping or binding of an assembly, or dropping a heavy load onto an assembly. A rise in readings on radiation monitors should be considered in conjunction with in-plant reports or observations of a potential fuel damaging event (e.g., a fuel handling accident).EAL

3Spent fuel peol water level at this "value is '-within the cowertend of the level Fange necessarf to prevent significant dese consequences ftrm dFrect gamma radiation to personnel pe~fGFMiR9 operaýtions in the- vicinity of the spent fuel pool. This condition reflects a significant loss of spent fuel pool water inventor,'

and thus it s also a precursor to a; loss of the ability to adequately GAool the raite ulasebe stored in the Peek Escalation of the emergency classification level would be via ICe A4-1RS1e -AS2 (seeA Susquehanna Basis Reference(s):

1. OP-179(279)-001 Area Radiation Monitoring System 2. PLA-6980 Enclosure 1 Susquehanna Units 1 & 2 Overall Integrated Plan with Regard to Reliable Spent Fuel Pool Instrumentation

3. AOP-081-001 Fuel Handling Abnormal Operating Procedure 4. EO-000-104 Secondary Containment Control 5. AR-101(201)-001

6. NEI 99-01 AA2 Page 38 of 195 Attachment I EP-RM-004 Revision [X]Page 65 of 295 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 < 10 ft. above the top of the spent fuel racks Mode Applicability: All Definition(s): None Susquehanna Basis: Each fuel storage pool is designed to maintain the water level in the pool above the top of active fuel providing cooling for the fuel rods. Technical Specifications require greater than or equal to 22 ft. of water be maintained over the top of irradiated fuel assemblies seated in the spent fuel storage racks at all times (ref. 1). In the event of loss of fuel pool inventory, Operations will assess multiple indications in accordance with ON-FPC-101(201) and AOP-081-001 (ref. 3, 4).Post-Fukushima order EA-12-051 required the installation of reliable SFP level indication capable of identifying normal SPF level 22.75 feet above the top of the fuel racks (Level 1 or el.81 7'-1 "), SFP level 10 ft. above .the top of the fuel racks (Level 2 or el. 804'-4") and SFP level at the top of the fuel racks (Level 3 or el. 794'-10" which for SSES is 0.5 feet on instrument above the top of the fuel racks). Each spent fuel pool is equipped with primary and backup guided wave radar probes to measure pool level. The range is continuous from the high pool level elevation (23.25 feet on instrument or el. 817'-7") to the top of the spent fuel racks (0.0 feet on instrument or el. 794'-4") (ref. 2).Each new SFP LI system provides two alarms to the associated Unit's Control Room Benchboard. The first alarm identifies a low water level condition at 20 ft. (instrument) above the spent fuel rack (elevation 814'-4"). The second alarm identifies a low low water level condition at 10 ft. (instrument) above the spent fuel rack (Level 2 or elevation 804'-4"). These alarms are intended to alert the control room operators of the loss of SFP inventory so actions are taken to provide make-up as soon as possible (ref. 2).NEI 99-01 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 see Devefew Notes). These events present radiological safety challenges to plant personnel and are precursors to a release of radioactivity to the environment. As such, they represent an actual or potential substantial degradation of the level of safety of the plant.This IC applie6 to irradiated fuel that is lieRnsed for dry torage up to the point that the loaded Storage cask i6 sealed. Once sealed, damage to a lo-aded- cask c-ausing loss ofeh CONFINEMENT BOUNDARY is classified in accordance With IC E H'J!.Page 39 of 195 Attachment 1 EP-RM-004 Revision [X]Page 66 of 295 Escalation of the cmergency Would be based On eithor Recognition Categor,' A Or C lGs.EAL #1 S Ths;F EAL escPralates, from AU2 in that the lo6s of level, in the affected portion ofth REFUELING PATHWAY, is Of sufficient magnitude to have resulted i, unoer; Of irradiated fuel. Irdicatieor Of irradiated fuel Unwovery may itnlude diret or idecmayt visual obsesi atiOn (e.g., reports from personnRel or camera images), as well as Sfignificant changes in water And radiatior devels, or ether plant parameters. Computatinal aids mhay also be used (e.g., a bee!Off GUR'e). Classification of an event using thiG EAL should be based on the totality of available indications, reports and obserwations. While aRn area adiation monito" rcoud detect an increase in a dose rate due to a onf woatwer level in some porton of the REFUELING PAThWAY, the reading may not be a reliable irndication f whether or not the fuel is actually uirnermed. To the degree possible, readings ould be onsidered ini combinathieon with ether available indicationr s of inventofrc 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.EAL #2 This EAL addresses a release of radioactive material caused by mechanical damage to irradiated fuel. Damaging events oay inhlude the dropping, bumping or binding of an assembly, or dropping a heavy load onto an assembly. A rise in readings On-Radiation rmnRitos should be considered in conjunction With in plant reports or obserwatiens of a potential fuel damagin event (e.g., a fuel handling acident).EAL #3 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 ICs ASI--RSle9F A92-(see-S Dev!peF Noes)e.EAL #T-his EAL escalates from AUJ2 in that the loss of level, in the affected portion Of the REFUELING ~ ~ .PAHAief suffiient magnitude to have resulted in; uncover,' ofirradiated fuel. Indications6 of irradiated fuel uncover,' may include direct Or indirect visual Gbserwation-(e.g., reports from personnel Or camera images), as well as significant changes on water and radiation levels, or other plant parameters. Computational aids May also be used (e.g., a boil of GUR'e). Classification of an event using this EAL sheuld be based on the totality of available indications, repor-ts and obserwatiens. IWhfffile a;n a;rea radiation monGitor could detect an inrGease in a dose rate. duo to a lowering of ate leelin some portion of the REFUELING PATHWAY, the reading9 may not be a reliable indication of whether Or not the fuel is actually uncovered. To9 the degree Possible, readings should be considered in combination with other available indications Of inventor,' loss.A drop in water level above irradiated fuel within the reactor vessel mnay be classified i Page 40 of 195 Attachment 1 EP-RM-004 Revision [X]Page 67 of 295 accordance RecGn9ition Category C during tho Cold Shutdo-AWn and Refueling modos.Susquehanna Basis Reference(s):

1. Technical Specifications 3.7.7 Spent Fuel Storage Pool Water Level 2. PLA-6980 Enclosure I Susquehanna Units 1 & 2 Overall Integrated Plan with Regard to Reliable Spent Fuel Pool Instrumentation

3. ON-FPC-101(201)

Loss of Fuel Pool Cooling 4. AOP-081-001 Fuel Handling Abnormal Operating Procedure 5. NEI 99-01 AA2 Page 41 of 195 Attachment 1 EP-RM-004 Revision [X]Page 68 of 295 Category: Subcategory: Initiating Condition: EAL: R -Abnormal Rad Levels I Rad Effluent 2 -Irradiated Fuel Event Unplanned loss of water level above irradiated fuel RU2.1 Unusual Event UNPLANNED water level drop in the REFUELING PATHWAY as indicated by any of the following on EITHER unit:* Fuel Pool Water Low Level alarm" Skimmer Surge Tank Low Level alarm" Visual observation of a water level drop below a fuel pool skimmer surge tank inlet" Observation of water draining down the outside wall of primary containment AND UNPLANNED rise in area radiation levels as indicated by any of the following radiation monitors:* Channel 14 Spent Fuel Pool Area Criticality Monitor" Channel 15 Refueling Floor Area* Channel 42 Refueling Floor Area 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.Susquehanna Basis: Each fuel storage pool is designed to maintain the water level in the pool above the top of active fuel providing cooling for the fuel rods. Technical Specifications require greater than or equal to 22 feet of water be maintained over the top of irradiated fuel assemblies seated in the spent fuel storage racks at all times (ref. 1). In the event of loss of fuel pool inventory, Operations will assess multiple indications in accordance with ON-FPC-101 (201) and AOP-081-001 (ref. 4, 6).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 RPV and spent fuel pool.Page 42 of 195 Attachment I EP-RM-004 Revision [X]Page 69 of 295 Control room alarms associated with elevated refuel floor area radiation levels include (ref. 4, 5): " REFUELING FLOOR AREA HI RADIATION (AR-101-D05)

SPENT FUEL POOL AREA HI RADIATION (AR-101-E05)

The listed ARMs are the normal range monitors that detect increasing area radiation due to a lack of shielding in the REFUELING PATHWAY (ref. 6).NEI 99-01 Basis: 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 AA2RA2.Susquehanna Basis Reference(s):

1. Technical Specifications 3.7.7 Spent Fuel Storage Pool Water Level 2. OP-179(279)-001 Area Radiation Monitoring System 3. AOP-081-001 Fuel Handling Abnormal Operating Procedure 4. EO-000-104 Secondary Containment Control 5. SSES-FSAR Table 12.3-7 Area Radiation Monitoring System 6. ON-FPC-101(201)

Loss of Fuel Pool Cooling 7. NEI 99-01 AU2 Page 43 of 195 Attachment 1 EP-RM-004 Revision [X]Page 70 of 295 Category: Subcategory: Initiating Condition: EAL: R -Abnormal Rad Levels / Rad Effluent 3 -Area Radiation Levels Radiation levels that IMPEDE access to equipment necessary for normal plant operations, cooldown or shutdown RA3.1 Alert Dose rates > 15 mR/hr in any of the following areas:* Main Control Room" Radwaste Control Room" Both the Central Alarm Station (CAS) and Secondary Alarm Station (SAS)Mode Applicability: All Definition(s): None Susquehanna Basis: Central Alarm Station (CAS) and Secondary Alarm Station (SAS) are included in this EAL because of their importance in permitting access to areas required to assure safe plant operations (ref. 1). Both are included in this EAL because either security station can effectively permit access to areas required to assure safe plant operations. It is not the intent of this EAL that there be continuous radiation monitoring in the CAS or SAS. However, if a radiological release is in progress and there are indications that the release may affect the CAS and SAS and dose rates in both areas are determined by manual radiological survey to be greater than 15 mR/hr, an Alert shall be declared.NEI 99-01 Basis: This IC addresses elevated radiation levels in certain plant rooms/areas sufficient to preclude or impede personnel from performing actions necessary to maintain normal plant operation, or to perform a normal plant cooldown and shutdown. As such, it represents an actual or potential substantial degradation of the level of safety of the plant. The Emergency Director/Recovery Manager should consider the cause of the increased radiation levels and determine if another IC may be applicable.For EAL #2, an Alert dGclaration is warranted if entry into the affected room/area is, o9r may be, ProedullY required duriRn the plant operating mode in effect at the time of the elevated radiation levels. The emergency classificatiORon, isnt contingent upon4 whether entry is actually necessar,' at the time of the inrGeased radiation levels. Access should be considered as impeded if extraordinary measures are necessar,' to facilitate entr,' ot personnel into the affected rooMlarea (e.g., installing temporary shielding reurn se Of non admigiStrative limits).Page 44 of 195 r"-UU1=O MA Ct" CA V" rl "0 Attachment 1 EP-RM-004 Revision [X]Page 71 of 295 An emergency declaration is not Warranted if any of the following conditions apply.* The plant is in an operating moede different than the mode specified for the affected FrGmCarea (i.e., entry is not required during the .perating mode On effect at the time o.the elevated radiation levels). For example, the plant is in Mode 1 when the radiation inceas ocurs, and the procedures used for nermal operation, cooldoWn and shutdown do not require entry into the affeted room until Mode-.,* The i emreased radiation levels are a result of a planed acgtivity that incuesgo compensator; measures Which address the temporyinceiblt of a room Or area (e.g., radiography, spent filter or resin transfer, etc.,).aThe action for which roomMarea cntr,' is requirFed is of an administrative Or record keepin nature (e.g., normal rounds or routine inspections).

The access control measures, are of a Gonsewative Or precautionar, nature, and would not actually prevent or impede a required action.Escalation of the emergency classification level would be via Recognition Category AýR, C or F ICs.Susquehanna Basis Reference(s):

1. AR-1650683

2. NEI 99-01 AA3 Page 45 of 195 Attachment 1 EP-RM-004 Revision [X]Page 72 of 295 Category: Subcategory:

Initiating Condition: R -Abnormal Rad Levels / Rad Effluent 3- Area Radiation Levels Radiation levels that IMPEDE access to equipment necessary for normal plant operations, cooldown or shutdown EAL: RA3.2 Alert An UNPLANNED event results in radiation levels that prohibit or IMPEDE access to any Table R-2 rooms or areas (Note 5)Note 5: If the equipment in the listed room or area was already inoperable or out-of-service before the event occurred, then no emergency classification is warranted. Table R-2 Safe Operation & Shutdown Areas Elevation Unit I Area(s) ** Unit 2 Area(s) ** Mode(s)670' RB 27 RB 32 3,4,5 683' RB 27, 28, 29 RB 32, 33, 34 3,4,5 703' RB 28, 29 RB 33, 34 3,4,5 719' RB 25, 29 RB 30, 34 3,4,5 749' RB 25, 29 RB 32, 33 3,4,5** See Chart I for location of plant areas Chart 1- Plant Area Key Plan 70 LOWLEVEL WATER TREATMENT RADWASTE H1481 ACIDAND I E I CHLORINE 53 50 BUILDING PUMPHOUSE IUNIT02 l l IUNIT , ITURBINE BLDG. I TURI Hl SPRAY POND VALVE VAULT ESSW 561 PUMPHOUSE RADWASTE r#1 BINE BLDG.TURB.16 115 1 14 1 13 14 3 2 1 38 3"/I E TUR 38 1 37 1 20 1 91 18 1171 8 1 7 6 5'I V-:/24 23 1 22 21 12 11 10 9 40 39 42 41--1-1 COND. STORAGE-INTAKE STRUCTURE 36 44 35 43 i---COND&REF STORAGE ,- DIESEL GENERATOR I E #2 E REACTOR REACTOR rlE DIESELRO GENERATOR Page 46 of 195 Attachment 1 EP-RM-004 Revision [X]Page 73 of 295 Mode Applicability: 3 -Hot Shutdown Definition(s): IMPEDE(D) -Personnel access to a room or area is hindered to an extent that extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., requiring use of protective measures such as temporary shielding, SCBAs or beyond Emergency Plan RWP dose extensions that are not routinely employed to access the room/area). 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.Susquehanna Basis: If the equipment in the listed room or area was already inoperable, or out-of-service, before the event occurred, then no emergency should be declared since the event will have no adverse impact beyond that already allowed by Technical Specifications at the time of the event.The list of plant areas in Table R-2 specify those areas that contain equipment which require a manual/local action as specified in operating procedures used for normal plant operation, cooldown and shutdown. Areas in which actions of a contingent or emergency nature would be performed (e.g., an action to address an off-normal or emergency condition such as emergency repairs, corrective measures or emergency operations) are not included. In addition, the list specifies the plant mode(s) during which entry would be required for each room or area. See Chart 1 for the specific locations of areas listed in Table R-2. See Attachment 3 for more details of how the Table R-2 was developed (ref. 1).NEI 99-01 Basis: This IC addresses elevated radiation levels in certain plant rooms/areas sufficient to preclude or impede personnel from performing actions necessary to maintain normal plant operation, or to perform a normal plant cooldown and shutdown. As such, it represents an actual or potential substantial degradation of the level of safety of the plant. The Emergency Director/Recovery Mana.qer should consider the cause of the increased radiation levels and determine if another IC may be applicable. For EAL-#a2RA3.2, an Alert declaration is warranted if entry into the affected room/area is, or may be, procedurally required during the plant operating mode in effect at the time of the elevated radiation levels. The emergency classification is not contingent upon whether entry is actually necessary at the time of the increased radiation levels. Access should be considered as impeded if extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., installing temporary shielding, requiring use of non-routine protective equipment, requesting an extension in dose limits beyond normal administrative limits).An emergency declaration is not warranted if any of the following conditions apply:* The plant is in an operating mode different than the mode specified for the affected room/area (i.e., entry is not required during the operating mode in effect at the time of Page 47 of 195 Attachment I EP-RM-004 Revision [X]Page 74 of 295 the elevated radiation levels). For example, the plant is in Mode I when the radiation increase occurs, and the procedures used for normal operation, cooldown and shutdown do not require entry into the affected room until Mode 4." The increased radiation levels are a result of a planned activity that includes compensatory measures which address the temporary inaccessibility of a room or area (e.g., radiography, spent filter or resin transfer, etc.)." The action for which room/area entry is required is of an administrative or record keeping nature (e.g., normal rounds or routine inspections)." The access control measures are of a conservative or precautionary nature, and would not actually prevent or impede a required action." If the equipment in the listed room or area was already inoperable, or out-of-service, before the event occurred, then no emergency should be declared since the event will have no adverse impact beyond that already allowed by Technical Specifications at the time of the event.Escalation of the emergency classification level would be via Recognition Category A R, C or F ICs.Susquehanna Basis Reference(s):

1. Attachment 3 Safe Operation

& Shutdown Areas Tables R-2 & H-2 Bases 2. NEI 99-01 AA3 Page 48 of 195 Attachment 1 EP-RM-004 Revision [X]Page 75 of 295 Category C -Cold Shutdown / Refueling System Malfunction EAL Group: Cold Conditions (RCS temperature _ 200 0 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 (4 -Cold Shutdown, 5 -Refueling, D -Defueled). The events of this category pertain to the following subcategories:

1. RPV Level Reactor Pressure Vessel water level is directly related to the status of adequate core cooling and, therefore, fuel clad integrity.

2. Loss of Essential AC Power Loss of essential plant electrical power can compromise plant safety system operability including decay heat removal and emergency core cooling systems which may be necessary to ensure fission product barrier integrity.

This category includes loss of onsite and offsite power sources for 4.16 kV ESS buses.3. Loss of Vital DC Power Loss of emergency plant electrical power can compromise plant safety system operability including decay heat removal and emergency core cooling systems which may be necessary to ensure fission product barrier integrity. This category includes loss of power to or degraded voltage on the 125 VDC vital buses.4. RCS Temperature Uncontrolled or inadvertent temperature or pressure increases are indicative of a potential loss of safety functions.

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.

Page 49 of 195 Attachment 1 EP-RM-004 Revision [X]Page 76 of 295 Category: Subcategory: Initiating Condition: C -Cold Shutdown / Refueling System Malfunction 1 -RPV Level Loss of RPV inventory affecting fuel clad integrity with Containment challenged EAL: CGI.1 General Emergency RPV level < -161 in. (TAF) for > 30 min. (Note 1)AND Any Containment Challenge indication, Table C-2 Note 1: The ED/RM 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)* PC hydrogen concentration

> 6%* UNPLANNED rise in PC pressure* Exceeding one or more Secondary Containment Control Max Safe Radiation Levels (EO-000-104 Table 9) that can be read in the Control Room (Table C-6)Page 50 of 195 Attachment 1 EP-RM-004 Revision [X]Page 77 of 295 Table C-6 Max Safe Reactor Building Radiation Limits Max Safe Rad RB Area ARM Channel Lxmit Elevation (ft) ARM Number Description LRmiR (R/HR)818 49 Refuel Floor Area 10 749 52 RWCU Recirc PP Access 10 54 Fuel Pool PP Area 719 50 CRD North 10 51 CRD South 670 53 Remote Shutdown Room 10 645 48 HPCI PP & Turbine Room 10 57 RCIC PP & Turbine Room 55 RHR A C PP Room 56 RHR B D PP Room Mode Applicability: 4 -Cold Shutdown, 5 -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 Susquehanna, CONTAINMENT CLOSURE is established per NDAP-QA-0309 (ref. 4) for Primary Containment OR is established per NDAP-QA-0321 (ref. 5) for Secondary Containment. 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.Susquehanna Basis: When RPV level drops below -161 in., core uncovery starts to occur (ref. 1).Four conditions are associated with a challenge to Primary Containment (PC) integrity: 0 CONTAINMENT CLOSURE is not established. Page 51 of 195 Attachment 1 EP-RM-004 Revision [X]Page 78 of 295 In the early stages of a core uncovery event, it is unlikely that hydrogen buildup due to a core uncovery could result in an explosive mixture of dissolved gases in the Primary Containment. However, Primary Containment monitoring and/or sampling should be performed to verify this assumption and a General Emergency declared if it is determined that an explosive mixture exists. An explosive mixture can be formed when hydrogen gas concentration in the Primary Containment atmosphere is greater than 6%by volume in the presence of oxygen (>5%) (ref. 2). In Cold Shutdown and Refueling modes it is assumed that the Primary Containment is de-inerted.

Any unplanned increase in PC pressure in the Cold Shutdown or Refueling mode indicates a potential loss of CONTAINMENT CLOSURE capability.

Unplanned Primary Containment pressure increases indicates CONTAINMENT CLOSURE cannot be assured and the Primary Containment cannot be relied upon as a barrier to fission product release.* Secondary Containment radiation monitors that can be read in the Control Room should provide indication of increased release that may be indicative of a challenge to CONTAINMENT CLOSURE. The Max Safe radiation levels are indicative of problems in the secondary containment that are spreading. The locations into which the primary system discharge is of concern correspond to the areas addressed in EO-000-104, Secondary Containment Control, (ref. 3).NEI 99-01 Basis: 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 RGS/FeaGto YesseIRPV 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 Page 52 of 195 Attachment 1 EP-RM-004 Revision [X]Page 79 of 295 indications to assess whether or not containment is challenged. in EAL 2.-btThe 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 (reactor vessel.R. S [P.. r or RPV [1,4r) level may be caused 4y instrumentation and/or power failures, or water lovel dropping below the -rnge of available if water level cannot be monitored, operator, may deterFmine that an inventor loss is occurring by obsoF~ing chane in su and/or tank levels. Sump and/or tanklel changes must be evaluated againSt other potential sources of water flow to ensure the" ar-e indicative of leakage from the (reactor ves...oRCS [DP/,'R. or RPV [BWRD.Thisese EALs 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. Susquehanna Basis Reference(s):

1. EO-000-101 RPV Control 2. EP-DS-001 Containment Combustible Gas Control 3. EO-000-104 Secondary Containment Control 4. NDAP-QA-0309 Primary Containment Access and Control 5. NDAP-QA-0321 Secondary Containment Integrity Control 6. NEI 99-01 CG1 Page 53 of 195 Attachment 1 EP-RM-004 Revision [X]Page 80 of 295 Category: Subcategory:

Initiating Condition: C -Cold Shutdown / Refueling System Malfunction 1 -RPV Level Loss of RPV inventory affecting fuel clad integrity with Containment challenged EAL: CG1.2 General Emergency RPV level cannot be monitored for - 30 min. (Note 1)AND UNPLANNED increase in any Table C-1 sump or tank level due to a loss of RPV inventory of sufficient magnitude to indicate core uncovery AND Any Containment Challenge indication, Table C-2 Note 1: The ED/RM 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* Drywell equipment drain tank* Drywell sumps* Reactor Building sump* LRW collection tanks* Main condenser hotwell* Suppression pool* Visual observation Table C-2 Containment Challenge Indications

CONTAINMENT CLOSURE not established (Note 6)* PC hydrogen concentration

> 6%* UNPLANNED rise in PC pressure* Exceeding one or more Secondary Containment Control Max Safe Radiation Levels (EO-000-104 Table 9) that can be read in the Control Room (Table C-6)Page 54 of 195 Attachment 1 EP-RM-004 Revision [X]Page 81 of 295 Table C-6 Max Safe Reactor Building Radiation Limits Max Safe Rad RB Area ARM Channel Limit Elevation (ft) ARM Number Description RimRt (R/HR)818 49 Refuel Floor Area 10 749 52 RWCU Recirc PP Access 10 54 Fuel Pool PP Area 719 50 CRD North 10 51 CRD South 670 53 Remote Shutdown Room 10 645 48 HPCI PP & Turbine Room 10 57 RCIC PP & Turbine Room 55 RHR A C PP Room 56 RHR B D PP Room Mode Applicability: 4 -Cold Shutdown, 5- Refueling Definition(s): UNPLANNED-. A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.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 Susquehanna, CONTAINMENT CLOSURE is established per NDAP-QA-0309 (ref. 7) for Primary Containment OR is established per NDAP-QA-0321 (ref. 8) for Secondary Containment. Susquehanna Basis: In Cold Shutdown mode, the RCS will normally be intact and standard RPV level monitoring means are available. RPV level in the Refueling mode is normally monitored using the Shutdown Range or temporary RPV shutdown level transmitter (ref. 1).If RPV level monitoring capability is unavailable, the RPV inventory loss must be detected by the leakage indications listed in Table C-1. Level increases must be evaluated against other Page 55 of 195 Attachment 1 EP-RM-004 Revision [X]Page 82 of 295 potential sources of leakage such as cooling water sources inside the drywell to ensure they are indicative of RPV leakage. Rise in drywell equipment drain tank level and drywell floor sump level is the normal method of monitoring and calculating leakage from the RPV (ref. 2). A Reactor Building sump level rise may also be indicative of RCS inventory losses external to the Primary Containment from systems connected to the RPV. With RHR System operating in the Shutdown Cooling mode, an unexplained rise in suppression pool water level could be indicative of RHR valve misalignment or leakage (ref. 3, 4). If the make-up rate to the RPV unexplainably rises above the pre-established rate, a loss of RPV inventory may be occurring even if the source of the leakage cannot be immediately identified. Visual observation of leakage from systems connected to the RCS that cannot be isolated could be indicative of a loss of RPV inventory. Four conditions are associated with a challenge to Primary Containment (PC) integrity:

CONTAINMENT CLOSURE is not established.
In the early stages of a core uncovery event, it is unlikely that hydrogen buildup due to a core uncovery could result in an explosive mixture of dissolved gases in the Primary Containment.

However, Primary Containment monitoring and/or sampling should be performed to verify this assumption and a General Emergency declared if it is determined that an explosive mixture exists. An explosive mixture can be formed when hydrogen gas concentration in the Primary Containment atmosphere is greater than 6%by volume in the presence of oxygen (>5%) (ref. 5). In Cold Shutdown and Refueling modes it is assumed that the Primary Containment is de-inerted." Any unplanned increase in PC pressure in the Cold Shutdown or Refueling mode indicates a potential loss of CONTAINMENT CLOSURE capability. Unplanned Primary Containment pressure increases indicates CONTAINMENT CLOSURE cannot be assured and the Primary Containment cannot be relied upon as a barrier to fission product release.* Secondary Containment radiation monitors should provide indication of increased release that may be indicative of a challenge to CONTAINMENT CLOSURE. The Max Safe radiation levels are indicative of problems in the secondary containment that are spreading. The locations into which the primary system discharge is of concern correspond to the areas addressed in EO-000-104, Secondary Containment Control, (ref. 6).NEI 99-01 Basis: 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 RG-,eaGtef vesseeRPV level cannot be restored, fuel damage is probable.With CONTAINMENT CLOSURE not established, there is a high potential for a direct and Page 56 of 195 Attachment 1 EP-RM-004 Revision [X]Page 83 of 295 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. in EAL 2.b-,tThe 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 (roactor vessel.R. S [PIl 4'R] or RPV fB.WR@ 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 Ve6 RCS RPV [81qr).Thisese EALs 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. Susquehanna Basis Reference(s):

1. IC 180 005 Installation and Removal of Temporary Unit 1 RPV Shutdown Level Transmitter

2. ON-100(200)-005 Excess Drywell Leakage Identification

3. OP-149(249)-002 RHR Operation in Shutdown Cooling Mode 4. ON-149(249)-001 Loss of RHR Shutdown Cooling Mode 5. EP-DS-001 Containment Combustible Gas Control 6. EO-000-104 Secondary Containment Control 7. NDAP-QA-0309 Primary Containment Access and Control 8. NDAP-QA-0321 Secondary Containment Integrity Control 9. NEI 99-01 CG1 Page 57 of 195 Attachment 1 EP-RM-004 Revision [X]Page 84 of 295 Category:

C -Cold Shutdown / Refueling System Malfunction Subcategory: 1 -RPV Level Initiating Condition: Loss of RPV inventory affecting core decay heat removal capability EAL: CSI.1 Site Area Emergency CONTAINMENT CLOSURE not established AND RPV level < -129 in. (Level 1)Mode Applicability: 4 -Cold Shutdown, 5 -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 Susquehanna, CONTAINMENT CLOSURE is established per NDAP-QA-0309 (ref. 3) for Primary Containment OR is established per NDAP-QA-0321 (ref. 4) for Secondary Containment. Susquehanna Basis: RPV level is normally monitored using the instruments in Figure C-1 (ref. 1, 2).When RPV level decreases to -129 in., RPV water level is below the low-low-low ECCS actuation setpoint (Level 1) (ref. 1).The magnitude of this loss of water indicates that makeup systems have not been effective and may not be capable of preventing further RPV water level decrease 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.NEI 99-01 Basis: This IC addresses a significant and prolonged loss of (FeaeteG-FssefRCS 'PWR}.eo-r-V[BWR]) wenteFylevel 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 RGSkeaete vesselRPV 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. Page 58 of 195 Attachment 1 EP-RM-004 Revision [X]Page 85 of 295 The difference in the specified RCS!reactor vesse!RPV levels of EAL-s-4-ICS1.1 and 2 7 bCS1.2 reflect the fact that with CONTAINMENT CLOSURE established, there is a lower probability of a fission product release to the environment. Ir EAL 3.a, the 30 mdirute citeioner is tied to a readily recognizable event star time (i.e., the total loss of ability to mon-itor level), and allows sufficient time to mnRitor, assess aNRd cG-elato Scactow and plant o-nditions to determine if core urcovei; has actually occurred (i.e., te actount foravariuN s aRCident prGriession and instrumentation ucAsesainties Itaow Malo suientt for pesormatne of actiognc ta terminate leakage, reouver inventor' ConGtrolmakeup equipment andSor reStore level f on csitoig.The inabilit' to mnucito (reatoir vessel/RIS [PWR] eo RPV [BheR]) level may be causedb i nstrumentation and/orIe power fasiues, or water level dropping below the range of availal N a io)f water level cannot be monitored, operators ma" deterie that an inventorA loss is occurring by .obsorwing change in supnd/or tank levels. Sump and/or tank lee changes mAust be evaluated against other potential sour-e-s-of water flow to ensure theyar indicative of leakage from the (reactor vessel/RCS [PVVR] Or RPV [B3WR]).-These-ThisEAL-s 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 CG 1 or AG-4RGI1.Susquehanna Basis Reference(s): 1.- M-142 P&ID Nuclear Boiler Vessel Instrumentation, Sheets 1, 2 2. ON-145(245)-004 RPV Water Level Anomaly 3. NDAP-QA-0309 Primary Containment Access and Control 4. NDAP-QA-0321 Secondary Containment Integrity Control 5. NEI 99-01 CS1 Page 59 of 195 Attachment 1 EP-RM-004 Revision [X]Page 86 of 295 Figure C-1 RPV Levels (ref. 1, 2)500" ui/ -180" 180" 658.50" MAIN STEAM SHUTDOWN UNE URANGE UPSET RANGE NARROW WIDE RANGE RANGE EXTENDED 599.00", 0" 0" 0" 517.00" 4 -150" -110" -150"$T NFUEL HEIGHT ABOVl INDICATION ZONE LEVEL VESSEL ZERO (INCHES)RANGE (NHCHES)8 581.5 +54-318" 7 566.5 +39 6 562.5 +35 152" 5 562.5 +35 INTRUME 4 557.5 +30 TAP 3 540.5 +13 2 489.5 -38 1 398.5 -129 0 324.5 -203 Page 60 of 195 Attachment 1 EP-RM-004 Revision [X]Page 87 of 295 Category: C -Cold Shutdown I Refueling System Malfunction Subcategory: 1 -RPV Level Initiating Condition: Loss of RPV inventory affecting core decay heat removal capability EAL: CS1.2 Site Area Emergency CONTAINMENT CLOSURE established AND RPV level < -161 in. (TAF)Mode Applicability: 4 -Cold Shutdown, 5 -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 Susquehanna, CONTAINMENT CLOSURE is established per NDAP-QA-0309 (ref. 2) for Primary Containment OR is established per NDAP-QA-0321 (ref. 3) for Secondary Containment. Susquehanna Basis: When RPV level drops to the top of active fuel (TAF) (an indicated RPV level of -161 in.), core uncovery starts to occur (ref. 1).NEI 99-01 Basis: This IC addresses a significant and prolonged loss of (reactor vesse.. R. S [P.R] .r RPV[-rWR]) inventor; wevel 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 RGS!ea4te vesselRPV 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 EAL-s-1-CS1.1 and 2.bCSl.2 reflect the fact that with CONTAINMENT CLOSURE established, there is a lower probability of a fission product release to the environment. In E.AL 2-a, the 30 minute criterion is tied to a r-eadily recognizable event start tim:e (i.e. the total loss of ability to moenitor level), and allows sufficient timne to moenitor, assess and correlate reactor and plant conditions to determine if core uncoerey has, actually occurred (i.e., to account Page 61 of 195 Attachment 1 EP-RM-004 Revision [X]Page 88 of 295 for various accident prors ionad instrumentation Unce~tainties). it also allows sufficfient time for pe~fermanco of actions16 to erint leakage, ccerientor; contrOl/makeup equipment and/or restore leveol nitoring.The in-blity to monitao (reator vesselRaS [PbyR] or RPV [LettR]) level may be Dcausedby iStrumentation and/or- power failures, or water level dropping below the rntge of availabe instrumentatioR. if water level canort be mn nitored, operatorss ay determwne that An ent ls9ala ios ofurrig by bsergncg clhanges in sump anwor tank levels. Sump and/RG tank level changes must be evaluated agairnst other potetial sources of water flow to ensure they-are indicative of leakage fE0ro the (reactor vessel!RS [PVR] Or RPV [BCRr).These-ThisEALs 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 OGI or AG4RG1.Susquehanna Basis Reference(s):

1. EO-000-.102 RPV Control 2. NDAP-QA-0309 Primary Containment Access and Control 3. NDAP-QA-0321 Secondary Containment Integrity Control 4. NEI 99-01 CS1 Page 62 of 195 Attachment 1 EP-RM-004 Revision [X]Page 89 of 295 C -Cold Shutdown / Refueling System Malfunction 1 -RPV Level Loss of RPV inventory affecting core decay heat removal capability Category: Subcategory:

Initiating Condition: EAL: CS1.3 Site Area Emergency RPV level cannot be monitored for 30 min. (Note 1)AND UNPLANNED increase in any Table C-1 sump or tank level due to a loss of RPV inventory of sufficient magnitude to indicate core uncovery Note 1: The ED/RM should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Table C-1 Sumps & Tanks* Drywell equipment drain tank* Drywell sumps* Reactor Building sump* LRW collection tanks* Main condenser hotwell* Suppression pool* Visual observation Mode Applicability: 4 -Cold Shutdown, 5 -Refueling Definition(s): UNPLANNED-. A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.Susquehanna Basis: In Cold Shutdown mode, the RCS will normally be intact and standard RPV level monitoring means are available. RPV level in the Refueling mode is normally monitored using the Shutdown Range or temporary RPV shutdown level transmitter (ref. 1).If RPV level monitoring capability is unavailable, the RPV inventory loss must be detected by the leakage indications listed in Table C-1. Level increases must be evaluated against other potential sources of leakage such as cooling water sources inside the drywell to ensure they are indicative of RPV leakage.Rise in drywell equipment drain tank level and drywell floor sump level is the normal method of monitoring and calculating leakage from the RPV (ref. 2). A Reactor Building sump level rise may also be indicative of RCS inventory losses external to the Primary Containment from Page 63 of 195 Attachment 1 EP-RM-004 Revision [X]Page 90 of 295 systems connected to the RPV. With RHR System operating in the Shutdown Cooling mode, an unexplained rise in suppression pool water level could be indicative of RHR valve misalignment or leakage (ref. 3, 4). If the make-up rate to the RPV unexplainably rises above the pre-established rate, a loss of RPV inventory may be occurring even if the source of the leakage cannot be immediately identified. Visual observation of leakage from systems connected to the RCS that cannot be isolated could be indicative of a loss of RPV inventory. NEI 99-01 Basis: This IC addresses a significant and prolonged loss of (reactor vessel!R.S [BWR]) 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 RGS1eaetGF Ve66elRPV level cannot be restored, fuel damage is probable.Outage/shutdoWn contingencY plans typically providc for Fe establiShing Or Yerif~ing CONTAINMENT CLOSURE folloWinRg a loss of heat removal or RCS inventorY contFol funrtions. The diffEcrLne in the specified RiStreadtor vossel levels of EALs 1o.b and 2.b seflcct the fact that with CONTAINMENT CLOSURE established, there is a lower probability of a fissio product release to thenirmnt... E. I -e-7t~he 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 (reactor vesselý"RS [P.IR] o-r RPV [-VWR] 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 (FeaGtoFvesse RCS [PVWR] or RPV [BVVR]).These-This EALs addresses concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal; SECY 91-283, Evaluation of Shutdown and Low Power Risk Issues; NUREG-1449, Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the United States;and NUMARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management. Escalation of the emergency classification level would be via IC CGI or AG4RG1 Susquehanna Basis Reference(s):

1. IC 180 005 Installation and Removal of Temporary Unit 1 RPV Shutdown Level Transmitter

2. ON-100(200)-005 Excess Drywell Leakage Identification

3. OP-149(249)-002 RHR Operation in Shutdown Cooling Mode 4. ON-149(249)-001 Loss of RHR Shutdown Cooling Mode Page 64 of 195 Attachment 1 EP-RM-004 Revision [X]Page 91 of 295 5. NEI 99-01 CS1 Page 65 of 195 Attachment 1 EP-RM-004 Revision [X]Page 92 of 295 C -Cold Shutdown / Refueling System Malfunction I -RPV Level Loss of RPV inventory Category: Subcategory:

Initiating Condition: EAL: CA1.1 Alert Loss of RPV inventory as indicated by RPV level < -38 in. (Level 2)Mode Applicability: 4 -Cold Shutdown, 5 -Refueling Definition(s): None Susquehanna Basis: The threshold RPV level of -38 in. is the low-low ECCS actuation setpoint (ref. 1). RPV level is normally monitored using the instruments in Figure C-1 (ref. 1, 2).When reactor vessel water level drops to -38 in. high pressure steam-driven injection sources HPCI (ECCS) and RCIC receive an initiation signal (ref. 1, 2). Although these systems cannot restore RCS inventory in the cold condition, the Level 2 actuation setpoint is operationally significant and is indicative of a loss of RCS inventory significantly below the low level scram setpoint specified in CUI.1.NEI 99-01 Basis: 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-#4, a lowering of water level below (cite specific 38 in. indicates that operator actions have not been successful in restoring and maintaining [PWR-G o-RPV level. The heat-up rate of the coolant will increase as the available water inventory is reduced. A continuing decrease in water level will lead to core uncovery.Although related, this EAL4 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 Residual Heat Removal suction point). An increase in RCS temperature caused by a loss of decay heat removal capability is evaluated under IC GA3CA4.For EAL #2, the inability to monitor (reactor vessel!RCS [PWR] or RPV [B3WR]) level may be caused by instrumentation and!or power failures, Or water level dropping below the range of available intumnato. if water level cannot be monOitored, operators mnay determine th-at An inventory loss isocurn byG obeP'ig change insmpad/or tank levels. Sump and/or tank level changes must be evaluated against other po-te-ntfial soeurces of water flow to ensure they are indicative of leakage from the (reactor Ycssel!RCS [PWVR] or RPV [BWR])-.The 15- minute duration for the loss of level indication Was chosen because iishalf of the EAL duration specified in C- G-21 Page 66 of 195 Attachment 1 EP-RM-004 Revision [X]Page 93 of 295 If the (Feactor v..seIRCS [PWR] or RPV [BWRr)-i ;i..eiy water level continues to lower, then escalation to Site Area Emergency would be via IC CS1.Susquehanna Basis Reference(s):

1. M-142 P&ID Nuclear Boiler Vessel Instrumentation, Sheets 1, 2 2. ON-145(245)-004 RPV Water Level Anomaly 3. NEI 99-01 CA1 Page 67 of 195 Attachment 1 EP-RM-004 Revision [X]Page 94 of 295 Figure C-1 RPV Levels (ref. 1, 2)VESSEL TOP 670,s" STEAM DRYER ASSEMBLY "z STEAM DRYER SEAL SKIRT *-SHUTDOWN RANGE INSTRUMENT ZERO 527.6" TAF 366.3" LEELN-ENLAUR BAF 216.3" VESSEL ZERO-- -INDICATION (INCHES)+54+39+35+35+30+13-38-129-203 Page 68 of 195 Attachment I EP-RM-004 Revision [X]Page 95 of 295 C -Cold Shutdown / Refueling System Malfunction 1 -RPV Level Loss of RPV inventory Category: Subcategory:

Initiating Condition: EAL: CA1.2 Alert RPV level cannot be monitored for > 15 min. (Note 1)AND UNPLANNED increase in any Table C-1 sump or tank levels due to a loss of RPV inventory Note 1: The ED/RM should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Table C-1 Sumps & Tanks" Drywell equipment drain tank* Drywell sumps* Reactor Building sump" LRW collection tanks* Main condenser hotwell* Suppression pool* Visual observation Mode Applicability: 4 -Cold Shutdown, 5 -Refueling Definition(s): UNPLANNED-. A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.Susquehanna Basis: In Cold Shutdown mode, the RCS will normally be intact and standard RPV level monitoring means are available. RPV level in the Refueling mode is normally monitored using the Shutdown Range or temporary RPV shutdown level transmitter (ref. 1).In this EAL, all water level indication is unavailable for greater than 15 minutes and the RPV inventory loss must be detected by the leakage indications listed in Table C-1. Level increases must be evaluated against other potential sources of leakage such as cooling water sources inside the drywell to ensure they are indicative of RPV leakage. Rise in drywell equipment drain tank level and drywell floor sump level is the normal method of monitoring and calculating leakage from the RPV (ref. 2). A Reactor Building sump level rise may also be indicative of RCS inventory losses external to the Primary Containment from systems connected to the RPV. With Page 69 of 195 Attachment I EP-RM-004 Revision [X]Page 96 of 295 RHR System operating in the Shutdown Cooling mode, an unexplained rise in suppression pool water level could be indicative of RHR valve misalignment or leakage (ref. 3, 4). If the make-up rate to the RPV unexplainably rises above the pre-established rate, a loss of RPV inventory may be occurring even if the source of the leakage cannot be immediately identified. Visual observation of leakage from systems connected to the RCS that cannot be isolated could be indicative of a loss of RPV inventory. NEI 99-01 Basis: 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 EAL #1, a lowering of water lovel below (satc specific level) indicates that operator actions have not been SUccessful in festoring and maintaining (reactor vesselR ran or RPV [aber) water level. The heat up rate of the olant well inrease as the avalable at inventor'y is reduced. A oc ntinuing decrease in water level will lead to core unc/over'. a Although related, E:AL #1 is concerned with the loss of RCS inventor,' and Rot th potential concgurent effects On systemS needed for decay heat removal (e.g., loss of a Residual Heat Removal suction point). An inrGease OR RCS temperature caused by a loss of decay heat removal capability is evaluated under [C CA3.For this EAL-#2, the inability to monitor (reactor v.sse.R.S [PW,7R or RPV ,-evel 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 (reac~tor ve~sscliRS [PVVR] or RPV-fBWRý). The 15-minute duration for the loss of level indication was chosen because it is half of the EAL duration specified in IC CS1I If the (FreaGtGr veose!RCS [PLVAR4] RPV fB3AR]) inventory level continues to lower, then escalation to Site Area Emergency would be via IC CSI.Susquehanna Basis Reference(s):

1. IC 180 005 Installation and Removal of Temporary Unit 1 RPV Shutdown Level Transmitter

2. ON-100(200)-005 Excess Drywell Leakage Identification

3. OP-149(249)-002 RHR Operation in Shutdown Cooling Mode 4. ON-149(249)-001 Loss of RHR Shutdown Cooling Mode 5. NEI 99-01 CAI Page 70 of 195 Attachment 1 EP-RM-004 Revision [X]Page 97 of 295 Category:

C -Cold Shutdown / Refueling System Malfunction Subcategory: I -RPV Level Initiating Condition: UNPLANNED loss of RPV inventory for 15 minutes or longer EAL: CUI.1 Unusual Event UNPLANNED loss of reactor coolant results in RPV level less than a required lower limit for -15 min. (Note 1)Note 1: The ED/RM should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Mode Applicability: 4 -Cold Shutdown, 5 -Refueling Definition(s): UNPLANNED-. A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.Susquehanna Basis: Figure C-1 illustrates the elevations of the RPV level instrument ranges (ref. 1).With the plant in Cold Shutdown, RPV water level is normally maintained above the RPV low level scram setpoint of +13 in. (ref. 1, 2). However, if RPV level is being controlled below the RPV low level scram setpoint, or if level is being maintained in a designated band in the reactor vessel it is the inability to maintain level above the low end of the designated control band due to a loss of inventory resulting from a leak in the RCS that is the concern.With the plant in Refueling mode, RPV water level is normally maintained at or above the reactor vessel flange (Technical Specifications 3.9.6 requires at least 22 ft of water above the top of the reactor vessel flange in the refueling cavity during refueling operations). The RPV flange is at an indicated level of 217.5 in. as indicated on the Shutdown Range RPV water level instrument (ref. 3).EAL RU2.1 may also be applicable based on increasing radiation levels due to loss of inventory in the REFULING PATHWAY.NEI 99-01 Basis: 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 (reGt. . [P-WR Gr-RPV [BW44LJ 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 Page 71 of 195 Attachment 1 EP-RM-004 Revision [X]Page 98 of 295 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 (reactor vs,,Is!RCQ [PLKR] or RPV [-9WR@level can change several times during the course of a refueling outage as different plant configurations and system lineups are implemented. This EAL is met if the minimum level, specified for the current plant conditions, cannot be maintained for 15 minutes or longer. The minimum level is typically specified in the applicable operating procedure but may be specified in another controlling document.The 15-minute threshold duration allows sufficient time for prompt operator actions to restore and maintain the expected water level. This criterion excludes transient conditions causing a brief lowering of water level.EAL= #2 addrcsses a condition where all means to detorminRe (reactor vesseliRCS [P'A'R]or RPV IBlWRD level have been lost. In this condition, operators mnay determine that a M nvontor~ loss is occurring by ebserwing chan~geinsp andio tank levels. Sump and/or tank feVel changes must be eva!luated agains6t other-poatential3 sourc-'Fes of water flow to ensure- they are indicative of leakage from the (reactor vessel/ROS [PWR] or RPV [B'A'R]).Continued loss of RCS inventory may result in escalation to the Alert emergency classification level via either IC CA1 or CA3.Susquehanna Basis Reference(s):

1. M-142 P&ID Nuclear Boiler Vessel Instrumentation, Sheets 1, 2 2. ON-145(245)-004 RPV Water Level Anomaly 3. GO-100(200)-006 Cold Shutdown, Refueling and Defueled 5. NEI 99-01 CUl Page 72 of 195 Attachment 1 EP-RM-004 Revision [X]Page 99 of 295 Figure C-1 RPV Levels (ref. 1, 2)500"180" 658.50" MAIN STEAM SHUTDOWN LINE URANGE UPSET NARROW WIDE E RANGE RANGE EXTENDED 60" RANGE 599.00" 0" 60 0" 0" 517.00"-150" -110" -150" 366" FUEL HEIGHT ABOVE INDICATION ZONE LEVEL VESSEL ZERO (INCHES)RANGE (INCHES (581.5 +54-310" 7 566.5 +39 6 562.5 +35152" 5 562.5 +35 INTRUME 4 557.5 +30 TAP 3 540.5 +13 2 489.5 -38 1 398.5 -129 0 324.5 -203 Page 73 of 195 Attachment 1 EP-RM-004 Revision [X]Page 100 of 295 C -Cold Shutdown / Refueling System Malfunction 1 -RPV Level UNPLANNED loss of RPV inventory for 15 minutes or longer Category: Subcategory:

Initiating Condition: EAL: CU1.2 Unusual Event RPV water level cannot be monitored AND UNPLANNED increase in any Table C-1 sump or tank levels due to a loss of RPV inventory Table C-1 Sumps & Tanks* Drywell equipment drain tank* Drywell sumps* Reactor Building sump* LRW collection tanks* Main condenser hotwell* Suppression pool* Visual observation Mode Applicability: 4 -Cold Shutdown, 5 -Refueling Definition(s): UNPLANNED-. A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.Susquehanna Basis: Note: The 15 minute criteria in the IC only applies to EAL CU1.1. If the conditions specified in this EAL last longer than 15 minutes, CA1.2 applies.In Cold Shutdown mode, the RCS will normally be intact and standard RPV level monitoring means are available. RPV level in the Refueling mode is normally monitored using the Shutdown Range or temporary RPV shutdown level transmitter (ref. 1).In this EAL, all water level indication is unavailable and the RPV inventory loss must be detected by the leakage indications listed in Table C-1. Level increases must be evaluated against other potential sources of leakage such as cooling water sources inside the drywell to ensure they are indicative of RPV leakage. Rise in drywell equipment drain tank level and drywell floor sump level is the normal method of monitoring and calculating leakage from the RPV (ref. 2). A Reactor Building sump level rise may also be indicative of RCS inventory losses Page 74 of 195 Attachment 1 EP-RM-004 Revision [X]Page 101 of 295 external to the Primary Containment from systems connected to the RPV. With RHR System operating in the Shutdown Cooling mode, an unexplained rise in suppression pool water level could be indicative of RHR valve misalignment or leakage (ref. 3, 4). If the make-up rate to the RPV unexplainably rises above the pre-established rate, a loss of RPV inventory may be occurring even if the source of the leakage cannot be immediately identified. Visual observation of leakage from systems connected to the RCS that cannot be isolated could be indicative of a loss of RPV inventory. NEI 99-01 Basis: 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 vesselRCS r er--RPV [BVVR])-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.EAL #1 rFecognizes that the min .i.u required (reacto vessel!R.S [P-rK or RPV[81,44R]) level can change several times during tho course of a refueling outage as different plant configurations and 6ystem lineups are implemented. This E.AL is Met if the mninimumn level, specified for the current plant conditions, cannot be maintained for 15 mnfinutes or longer. The miniimum level is typically specified in the applicable operating procedure but may be specife in another controlling decument.The 15 minute threshold duration allows suffiient time for perompt operator actions to restoee and maintain the expected water level. This criterion excludes transient conditions causn brief lawering of water level.This EAL addresses a condition where all means to determine (reactor ves.el.R. S , o RPV--BVRA level have been lost. In this condition, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the (reactor vesseliRCS [PWR]or RPVL B..R-..Continued loss of RCS inventory may result in escalation to the Alert emergency classification level via either IC CA1 or GA3CA4.Susquehanna Basis Reference(s):

1. IC 180 005 Installation and Removal of Temporary Unit 1 RPV Shutdown Level Transmitter

2. ON-100(200)-005 Excess Drywell Leakage Identification

3. OP-149(249)-002 RHR Operation in Shutdown Cooling Mode 4. ON-149(249)-001 Loss of RHR Shutdown Cooling Mode 5. NEI 99-01 CUI Page 75 of 195 Attachment 1 EP-RM-004 Revision [X]Page 102 of 295 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 essential buses for 15 minutes or longer EAL: CA2.1 Alert Loss of ALL offsite and ALL onsite AC power capability to ALL 4.16 kV ESS buses on EITHER unit for> 15 min. (Note 1)Note 1: The ED/RM should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Mode Applicability: 4 -Cold Shutdown, 5 -Refueling, D -Defueled Susquehanna Basis: The Class 1E 4.16 kV system supplies all the Engineered Safety Feature (ESF) loads and other loads that are needed for a safe and orderly plant shutdown, and for keeping the plant in a safe shutdown condition. See Figure C-2 (ref. 1, 2) The eight Class 1E 4.16 kV ESS Buses 1(2)A through 1(2)D receive power from either the four ESS 13.8/4.16 kV transformers or the diesel generators (A, B, C, D and additional diesel generator E). Buses 1A-1D supply Unit 1 and common loads and Buses 2A-2D supply Unit 2 loads. This configuration prevents a loss of all ESS Buses for one unit in the event one of the ESS Transformers is lost.During normal plant operation, ESS Transformer 101 supplies preferred power to ESS Bus IA and 2A and is an alternate power supply to ESS bus 1D and 2D. ESS Transformer 111 supplies preferred power to ESS Bus 1C and 2C, and is an alternate power supply to ESS bus 1 B and 2B. ESS Transformer 201 supplies preferred power to ESS Bus 1D and 2D, and is an alternate power supply to ESS bus 1A and 2A. ESS Transformer 211 supplies preferred power to ESS Bus 1 B and 2B, and is an alternate power supply to ESS bus 1C and 2C.On a loss of a preferred power source, the bus rapidly transfers to the alternate power source to maintain component power. If both the preferred and alternate power sources are lost, the associated standby diesel generator connects to the ESS bus. (ref. 2-6)This cold condition EAL is equivalent to the hot condition loss of all offsite AC power EAL SSI. 1.NEI 99-01 Basis: This IC addresses a total loss of AC power that compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink.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 Page 76 of 195 Attachment 1 EP-RM-004 Revision [X]Page 103 of 295 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 AS-1-RS1.Susquehanna Basis Reference(s):

1. FSAR Section 8.2 Offsite Power System 2. FSAR Section 8.3 Onsite Power System 3. Technical Specifications 3.8.2 AC Sources -Shutdown 4. Technical Specifications 3.8.8 Distribution System -Shutdown 5. ON-104 (204)-001 Units 1(2) Response to Loss of All Offsite Power 6. EO-100 (200)-030 UNIT 1(2) Response to Station Blackout 7. NEI 99-01 CA2 Page 77 of 195 Attachment 1 EP-RM-004 Revision [X]Page 104 of 295 Figure C-2 ESS 13.8/4.16 kV Transformers and Distribution (ref. 1)Page 78 of 195 Attachment I EP-RM-004 Revision [X]Page 105 of 295 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 essential buses for 15 minutes or longer EAL:[U2.1 Unusual Event AC power capability to ALL 4.16 kV ESS buses on EITHER unit 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 ED/RM should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Mode Applicability: 4 -Cold Shutdown, 5 -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. Susquehanna Basis: The Class IE 4.16 kV system supplies all the Engineered Safety Feature (ESF) loads and other loads that are needed for a safe and orderly plant shutdown, and for keeping the plant in a safe shutdown condition. See Figure C-2 (ref. 1, 2) The eight Class 1E 4.16 kV ESS Buses 1(2)A through 1(2)D receive power from either the four ESS 13.8/4.16 kV transformers or the diesel generators (A, B, C, D and additional diesel generator E). Buses 1A-1D supply Unit 1 and common loads and Buses 2A-2D supply Unit 2 loads. This configuration prevents a loss of all ESS Buses for one unit in the event one of the ESS Transformers is lost.During normal plant operation, ESS Transformer 101 supplies preferred power to ESS Bus 1A and 2A and is an alternate power supply to ESS bus 1D and 2D. ESS Transformer 111 supplies preferred power to ESS Bus 1 C and 2C, and is an alternate power supply to ESS bus 1B and 2B. ESS Transformer 201 supplies preferred power to ESS Bus ID and 2D, and is an Page 79 of 195 Attachment 1 EP-RM-004 Revision [X]Page 106 of 295 alternate power supply to ESS bus 1A and 2A. ESS Transformer 211 supplies preferred power to ESS Bus 1B and 2B, and is an alternate power supply to ESS bus 1C and 2C.On a loss of a preferred power source, the bus rapidly transfers to the alternate power source to maintain component power. If both the preferred and alternate power sources are lost, the associated standby diesel generator connects to the ESS bus. (ref. 2-6)This cold condition EAL is equivalent to the hot condition EAL SAI.1.NEI 99-01 Basis: 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-aRd-E-QPs, and capable of supplying required power to an essential-merqency bus. Some examples of this condition are presented below.* A loss of all offsite power with a concurrent failure of all-but-ne division of emergency power sources (e.g., an-onsite diesel generators)." 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 tFah4 division of emergency buses being back-fed from an offsite power source.Fifteen minutes was selected as a threshold to exclude transient or momentary losses of power.The subsequent loss of the remaining single power source would escalate the event to an Alert in accordance with IC CA2.Susquehanna Basis Reference(s):

1. FSAR Section 8.2 Offsite Power System 2. FSAR Section 8.3 Onsite Power System 3. Technical Specifications 3.8.2 AC Sources -Shutdown 4. Technical Specifications 3.8.8 Distribution System -Shutdown 5. ON-1 04 (204)-001 Units 1(2) Response to Loss of All Offsite Power 6. EO-100 (200)-030 UNIT 1(2) Response to Station Blackout 7. NEI 99-01 CU2 Page 80 of 195 Attachment 1 EP-RM-004 Revision [X]Page 107 of 295 Figure C-2 ESS 13.8/4.16 kV Transformers and Distribution (ref. 1)Page 81 of 195 Attachment 1 EP-RM-004 Revision [X]Page 108 of 295 Category:

C -Cold Shutdown / Refueling System Malfunction Subcategory: 3 -Loss of Vital DC Power Initiating Condition: Loss of vital DC power for 15 minutes or longer EAL: CU3.1 Unusual Event< 105 VDC bus voltage indications on Technical Specification required 125 VDC buses for > 15 min. (Note 1)Note 1: The ED/RM should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Mode Applicability: 4 -Cold Shutdown, 5 -Refueling Definition(s): None Susquehanna Basis: The Class 1E Battery Banks are 1(2)D610 (Channel A), 1(2)D620 (Channel B), 1(2)D630 (Channel C), and 1(2)D640 (Channel D). Each bank consists of 60 cells connected in series.Each cell produces a nominal voltage of 2.06 VDC resulting in a total battery bank terminal voltage of 123.6 VDC. All battery banks are designed to supply power to its load center for four hours in the event of a loss of power from its battery charger (ref. 1-3).105 VDC is the minimum design voltage limit (ref. 4).Indicated voltage for the vital 125 VDC main distribution buses is local only. Local voltage indication is available for each bus based on dispatching a field operator in accordance with Control Room alarm response procedure AR-1(2)06-001 (A12,B12,C12,D12). Field observation of indicated voltage constitutes the point in time when availability of indications to plant operators that an emergency action level has been, or may be, exceeded.This EAL is the cold condition equivalent of the hot condition loss of DC power EAL SS2.1.NEI 99-01 Basis 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 TFain AlDivision I is out-of-service (inoperable) for scheduled outage maintenance work and Train 8Division II is in-service (operable), then a loss of Vital DC power affecting T--aiR-BDivision II Page 82 of 195 Attachment 1 EP-RM-004 Revision [X]Page 109 of 295 would require the declaration of an Unusual Event. A loss of Vital DC power to TFain-ADivision I 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 GA-3CA4, or an IC in Recognition Category AR.Susquehanna Basis Reference(s):

1. FSAR Section 8.3.2 DC Power Systems 2. Susquehanna Drawing No. E107159, Sheet 1, "Single Line Meter & Relay Diagram 125 VDC, 250 VDC & 120 VAC Systems" 3. Technical Specifications 3.8.5 DC Sources -Shutdown 4. ON-102(202)-610, -620, -630, -640 Loss of 125V DC 5. AR-1(2)06-001 Main Turbine/Generator, Computer HVAC, Instrument AC, 24V DC, 125V DC, 250V DC Panel 2C651 6. NEI 99-01 CU4 Page 83 of 195 Attachment 1 EP-RM-004 Revision [X]Page 110 of 295 C -Cold Shutdown / Refueling System Malfunction 4 -RCS Temperature Inability to maintain the plant in cold shutdown Category: Subcategory:

Initiating Condition: EAL: CA4.1 Alert UNPLANNED increase in RCS temperature to > 200OF for > Table C-3 duration (Note 1)OR UNPLANNED RPV pressure increase > 10 psig due to loss of decay heat removal capability Note 1: The ED/RM should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.Table C-3: RCS Heat-up Duration Thresholds RCS Status CONTAINMENT CLOSURE Heat-up Duration Status INTACT N/A 60 min.*established 20 min.*Not fNTACT not established 0 min.* If an RCS heat removal system is in operation within this time frame and RCS temperature is being reduced, the EAL is not applicable. Mode Applicability: 4 -Cold Shutdown, 5- Refueling Definition(s): CONTAINMENT CLOSURE -The 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 Susquehanna, CONTAINMENT CLOSURE is established per NDAP-QA-0309 (ref. 3) for Primary Containment OR is established per NDAP-QA-0321 (ref. 4) for Secondary Containment. 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).UNPLANNED -. A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.Page 84 of 195 Attachment 1 EP-RM-004 Revision [X]Page 111 of 295 Susquehanna Basis: In the absence of reliable RCS temperature indication caused by the loss of decay heat removal capability, classification should be based on the RPV pressure increase criteria when the RCS is intact in Mode 4 or based on time to boil data when in Mode 5 or the RCS is not intact in Mode 4.Available methods of determining RCS temperature can be found in Operations surveillance procedures (ref. 1, 2).10 psig is the lowest pressure increase increment that can be reasonably read in the control room (ref. 2): NEI 99-01 Basis: 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 in-ventory is reduced (e.g., mid loop operation in; PWRs). 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 eO-is-at redu.ed inventorY [PWrI, 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 Primary Containment or Reactor Building atmosphere and subsequently to the environment, and 2) there is reduced reactor coolant inventory above the top of irradiated fuel.E-AL-#2The 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 AS4RS1.Susquehanna Basis Reference(s):

1. SO-100(200)-011 Reactor Vessel Temperature and Pressure Recording 2. SO-100(200)-006 Shiftly Surveillance Operating Log 3. NEI 99-01 CA3 Page 85 of 195 Attachment I EP-RM-004 Revision [X]Page 112 of 295 Category:

C -Cold Shutdown / Refueling System Malfunction Subcategory: 4 -RCS Temperature Initiating Condition: UNPLANNED increase in RCS temperature EAL: CU4.1 Unusual Event UNPLANNED increase in RCS temperature to > 200OF due to loss of decay heat removal capability Mode Applicability:. 4 -Cold Shutdown, 5 -Refueling Definition(s): UNPLANNED-. A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.Susquehanna Basis: In the absence of reliable RCS temperature indication caused by a loss of decay heat removal capability, classification should be based on EAL CU4.2 should RPV level indication be subsequently lost.Available methods of determining RCS temperature can be found in Operations surveillance procedures (ref. 2, 3).NEI 99-01 Basis: 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 Director/Recovery Manager should also refer to IC QA4CA4.A momentary UNPLANNED excursion above the Technical Specification cold shutdown temperature limit when the heat removal function is available does not warrant a classification. EAL-#4-This EAL involves a loss of decay heat removal capability, or an addition of heat to the RCS in excess of that which can currently be removed, such that reactor coolant temperature cannot be maintained below the cold shutdown temperature limit specified in Technical Specifications. During this condition, there is no immediate threat of fuel damage because the core decay heat load has been reduced since the cessation of power operation. During an outage, the level in the reactor vessel will normally be maintained at or above the reactor vessel flange. Refueling evolutions that lower water level below the reactor vessel flange are carefully planned and controlled. A loss of forced decay heat removal at reduced inventory may result in a rapid increase in reactor coolant temperature depending on the time after shutdown.Page 86 of 195 Attachment 1 EP-RM-004 Revision [X]Page 113 of 295 EAli #2 reflects a ou ndition whee there has been a signiftant loss o f C badon capability nepessan' to monitor RCS pcoditiens and operators would be unable to mtnitor key parameters nnessar; to assure core decay heat removal. During this condition, thero is n i mmedieate threat of fuel damage because the core decay heat load has boen reduced since the cessation of power operation. Fifteen minutes was selected as a threshold to eXclude transGient or Moeneta~' losses of Escalation to Alert would be via IC CAl based on an inventory loss or IC GA3-CA4_based on exceeding plant config uratio n-specifi c time criteria.Susquehanna Basis Reference(s):

1. Technical Specifications Table 1.1-1 2. SO-100(200)-011 Reactor Vessel Temperature and Pressure Recording 3. SO-100(200)-006 Shiftly Surveillance Operating Log 4. NEI 99-01 CU3 Page 87 of 195 Attachment 1 EP-RM-004 Revision [X]Page 114 of 295 Category:

C -Cold Shutdown /.Refueling System Malfunction Subcategory: 4 -RCS Temperature Initiating Condition: UNPLANNED increase in RCS temperature EAL: CU4.2 Unusual Event Loss of ALL RCS temperature and RPV level indication for > 15 min. (Note 1)Note 1: The ED/RM should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Mode Applicability: 4 -Cold Shutdown, 5 -Refueling Definition(s): None Susquehanna Basis: Available methods of determining RCS temperature can be found in Operations surveillance procedures (ref. 4, 5).RPV level is normally monitored using the instruments in Figure C-2 (ref. 1, 2).NEI 99-01 Basis: This ,C-EAL addresses an UNPLANN.ED inc.rease in RCS temperature above the TGchnical Specificartion cold s,-hutd, ot n e ... t ..err limt,49 or the inability to determine RCS temperature and RPV level, and represents a potential degradation of the level of safety of the plant. If the RCS is not intact and CONTAINMENT CLOSURE is not established during this event, the Emergency Director/Recovery Manager should also refer to IC GA3CA4.A mrnenetar; UNPLANNED excursion above the Technical Specification cold shutdown temperature limit when the heat removal fiuGtion is available does net warant a clssifiation. EAL #1 irvelves a loss of decay heat removal capability, oF an addition of heat to the RCS in excess-,R of tha-t Which can currently be removed, such that reactor coolant temperature cannot be maintained belowA. thlencold shutdown temperaturfe limit specified in Technica 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. DurngFR an outage, the level in the reactor vessel will normally be maintained above the reactor vessel flange. Refueling evolutions that lower water level below the reactor vesse flange arc carefully planned and controlled. A loss of forced decay heat rem oval at reduced i nventor; may result in a rapid inrGease in reactor coolant temperature depending on the time after shutdown.EAI=-#2This 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 Page 88 of 195 Attachment 1 EP-RM-004 Revision [X]Page 115 of 295 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 GA2-CA4 based on exceeding plant configuration-specific time criteria.Susquehanna Basis Reference(s):

1. M-142 P&ID Nuclear Boiler Vessel Instrumentation, Sheets 1, 2 2. ON-145(245)-004 RPV Water Level Anomaly 3. Technical Specifications Table 1.1-1 4. SO-1 00(200)-011 Reactor Vessel Temperature and Pressure Recording 5. SO-100(200)-006 Shiftly Surveillance Operating Log 6. NEI 99-01 CU3 Page 89 of 195 Attachment 1 EP-RM-004 Revision [X]Page 116 of 295 Figure C-2 RPV Levels (ref. 1, 2)0O0 s00" 180" 180" 658.50", MAIN STEAM SHUTDOWN LUNE RANGE UPSET RANGE-NARROW WIDE E AANGE EXTENDED 599.00" 60" RANGE 0" 0" 517.00"-150" -110" -150" FUEL HEIGHTABOVE INDICATION RNGE L VESSEL ZERO (INCHES)PAE(INCHES)

____8 581.5 +54-310" 7 566.5 +39 y 6 562.5 +35 152" 5 562.5 +35 JET PUMP INSTRUMENT 4 557.5 +30 TAP 3 540.5 +13 2 489.5 -38 1 398.5 -129 0 324.5 -203 Page 90 of 195 Category: Subcategory: Initiating Condition: EAL: Attachment 1 EP-RM-004 Revision [X]Page 117 of 295 C -Cold Shutdown / Refueling System Malfunction 5 -Loss of Communications Loss of all onsite or offsite communications capabilities CU5.1 Unusual Event Loss of ALL Table C-4 onsite communication methods OR Loss of ALL Table C-4 ORO communication methods OR Loss of ALL Table C-4 NRC communication methods Table C-4 Communication Methods System Onsite ORO NRC UHF Radio X Plant PA System X Dedicated Conference Lines X Commercial Telephone Systems X X X Cellular Telephone X X FTS-2001 (ENS) X X Mode Applicability: 4 -Cold Shutdown, 5 -Refueling, D -Defueled Definition(s): None Susquehanna Basis: Onsite/offsite communications include one or more of the systems listed in Table C-4 (ref. 1, 2, 3).UHF Radio Onsite portable radio communication systems are described in the Susquehanna SES Physical Security Plan and in the Susquehanna SES Emergency Plan. Four UHF channels, each Page 91 of 195 Attachment 1 EP-RM-004 Revision [X]Page 118 of 295 consisting of two frequencies for duplex operation through one of five in-plant repeaters, provide onsite portable radio communications. Operations is assigned two channels; one channel is assigned to Unit 1 and one to Unit 2. Operators in the plant on rounds and on specific assignments are equipped with handheld two-way radios.Plant PA System The plant PA system is an intra-plant public address providing the following functions:

A 5-channel page-talk handset intercom system for on-site communications between plant locations.
Broadcast accountability and fire alarms designed to warn personnel of emergency conditions.

The system consists of telephone handsets, amplifiers and loudspeakers located at various selected areas throughout the plant.Dedicated Conference Lines (Centrex Three (3) diqit dialinQ)The Dedicated Conference Lines are those normally used to communicate with several offsite agencies at one time (e.g., 191 conference line).Commercial Telephone Systems Two independent telecommunications networks exist to provide primary and backup telephone communications between ERFs and offsite agencies.Plant Cellular Telephone Cell phones can be utilized to perform both ORO and NRC communications. FTS 2001 (ENS)This system is for NRC offsite communications but may also be used to perform ORO notifications. This EAL is the cold condition equivalent of the hot condition EAL SU7.1.NEI 99-01 Basis: This IC addresses a significant loss of on-site or offsite communications capabilities. While not a direct challenge to plant or personnel safety, this event warrants prompt notifications to 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.).EAL-#4The first EAL condition addresses a total loss of the communications methods used in support of routine plant operations. EAL-#2The 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 (see Notes) the Commonwealth of Pennsylvania, Luzerne and Columbia County EOCs-Page 92 of 195 Attachment 1 EP-RM-004 Revision [X]Page 119 of 295 EA[--#3The third EAL addresses a total loss of the communications methods used to notify the NRC of an emergency declaration. Susquehanna Basis Reference(s):

1. EP-RM-007 Emergency Telephone Instructions and Directory 2. SSES Emergency Plan Section 8 3. FSAR Section 9.5.2 4. NEI 99-01 CU5 Page 93 of 195 Attachment 1 EP-RM-004 Revision [X]Page 120 of 295 Category: Subcategory:

Initiating Condition: EAL: C -Cold Shutdown / Refueling System Malfunction 6 -Hazardous Event Affecting Safety Systems Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode CA6.1 Alert The occurrence of any Table C-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 C-5 Hazardous Events* Seismic event (earthquake)

Internal or external FLOODING event* High winds or tornado strike* FIRE* EXPLOSION 9 Other events with similar hazard characteristics as determined by the Shift Manager Mode Applicability:

4 -Cold Shutdown, 5 -Refueling Definition(s): EXPLOSION -A rapid, violent and catastrophic failure of a piece of equipment due to combustion, chemical reaction or overpressurization. A release of steam (from high energy lines or components) or an electrical component failure (caused by short circuits, grounding, arcing, etc.) should not automatically be considered an explosion. Such events require a post-event inspection to determine if the attributes of an explosion are present.Page 94 of 195 Attachment 1 EP-RM-004 Revision [X]Page 121 of 295 FIRE -Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred but is NOT required if large quantities of smoke and heat are observed.FLOODING -A condition where water is entering a room or area faster than installed equipment is capable of removal, resulting in a rise of water level within the room or area.SAFETY SYSTEM -A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 1OCFR50.2): 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. Susquehanna Basis:* The significance of a seismic event is discussed under EAL HU2.1 (ref. 1, 2)." Internal FLOODING may be caused by events such as component failures, equipment misalignment, or outage activity mishaps (ref. 3, 4, 5).* Seismic Category I structures are analyzed to withstand a sustained, design wind velocity of at least 80 mph. (ref. 6, 7).* Areas containing functions and systems required for safe shutdown of the plant are identified by Fire Zone in the fire response procedure (ref. 8, 9).* 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.NEI 99-01 Basis: 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.EAL-I -.4-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.-AL-4.b.2The 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 Page 95 of 195 Attachment 1 EP-RM-004 Revision [X]Page 122 of 295 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 AS4RS1.Susquehanna Basis Reference(s):

1. ON-000-002 Severe Weather / Natural Phenomena 2. FSAR Section 3.7 Seismic Design 3. ON-169(269)-001 Flooding in Turbine Building 4. ON-169(269)-002 Flooding in Reactor Building 5. FSAR Section 3.4 Water Level (Flood) Design 6. FSAR Section 3.3 Wind and Tornado Loadings 7. FSAR Section 3.5 Missile Projection

8. SSES-FPRR Section 6.2 Fire Area Description

9. ON-013-001 Response to Fire 9. NEI 99-01 CA6 Page 96 of 195 Attachment 1 EP-RM-004 Revision [X]Page 123 of 295 Category H -Hazards and Other Conditions Affecting Plant Safety EAL Group: ANY (EALs in this category are applicable to any plant condition, hot or cold.)Hazards are non-plant, system-related events that can directly or indirectly affect plant operation, reactor plant safety or personnel safety.1. Security Unauthorized entry attempts into the Protected Area, bomb threats, sabotage attempts, and actual security compromises threatening loss of physical control of the plant.2. Seismic Event Natural events such as earthquakes have potential to cause plant structure or equipment damage of sufficient magnitude to threaten personnel or plant safety.3. Natural or Technology Hazard Other natural and non-naturally occurring events that can cause damage to plant facilities include tornados, FLOODING, hazardous material releases and events restricting site access warranting classification.

4. Fire Fires can pose significant hazards to personnel and reactor safety. Appropriate for classification are fires within the site Protected Area or which may affect operability of equipment needed for safe shutdown 5. Hazardous Gas Toxic, corrosive, asphyxiant or flammable gas leaks can affect normal plant operations or preclude access to plant areas required to safely shutdown the plant.6. Control Room Evacuation If the Control Room must be evacuated, additional support for monitoring and controlling plant functions is necessary through the emergency response facilities.

7. ED/RM Judqment The EALs defined in other categories specify the predetermined symptoms or events that are indicative of emergency or potential emergency conditions and thus warrant classification.

While these EALs have been developed to address the full spectrum of possible emergency conditions which may warrant classification and subsequent implementation of the Emergency Plan, a provision for classification of emergencies based on operator/management experience and judgment is still necessary. The EALs of this category provide the Emergency Director/Recovery Manager the latitude to classify emergency conditions consistent with the established classification criteria based upon Emergency Director/Recovery Manager judgment.Page 97 of 195 Attachment 1 EP-RM-004 Revision [X]Page 124 of 295 Category: Subcategory: Initiating Condition: EAL: H -Hazards I -Security HOSTILE ACTION resulting in loss of physical control of the facility HGI.1 General Emergency A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by the Security Shift Supervision AND EITHER of the following has occurred: Any of the following safety functions cannot be controlled or maintained" Reactivity" RPV water level" RCS heat removal OR Damage to spent fuel has occurred or is IMMINENT Mode Applicability: All Definition(s): HOSTILE ACTION -An act toward Susquehanna 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 Susquehanna. 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 -Area within the station inner security fence (PROTECTED AREA Barrier)designated to implement the requirements of 10 CFR 73.Susquehanna Basis: Security Shift Supervision 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 SSES Physical Security Plan (Safeguards) information (ref. 1).This IC does not apply to a HOSTILE ACTION directed at an ISFSI PROTECTED AREA; such an attack should be assessed using IC HA1.Page 98 of 195 Attachment 1 EP-RM-004 Revision [X]Page 125 of 295 If the plant equipment necessary to maintain the safety functions can be controlled from another location, then the EAL is not met.Loss of SAS and/or CAS does not impact equipment needed for safety functions. NEI 99-01 Basis: 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 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.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-fand Independent Spent Fuel Storage Instalation Securiy 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 Susquehanna Physical Security Plan (ref.1).Susquehanna Basis Reference(s):

1. SSES Physical Security Plan 2. ON-000-010 Security Event 3. NEI 99-01 HG1 Page 99 of 195 Attachment 1 EP-RM-004 Revision [X]Page 126 of 295 Category:

H -Hazards Subcategory: 1 -Security Initiating Condition: HOSTILE ACTION within the PROTECTED AREA EAL: HSI.1 Site Area Emergency A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by the Security Shift Supervision Mode Applicability: All Definition(s): HOSTILE ACTION -An act toward Susquehanna 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 Susquehanna. 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 -Area within the station inner security fence (PROTECTED AREA Barrier)designated to implement the requirements of 10 CFR 73.Susquehanna Basis: Security Shift Supervision 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 SSES Physical Security Plan (Safeguards) information (ref. 1).NEI 99-01 Basis: This IC addresses the occurrence of a HOSTILE ACTION within the PROTECTED AREA. This event will require rapid response and assistance due to the possibility for damage to plant equipment. Timely and accurate communications between Security Shift Supervision and the Control Room is essential for proper classification of a security-related event.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-jand independent Spent Fue! Sterage In-st-allation SeGcudty Pro gF9MJ.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 Orcianization Page 100 of 195 Attachment 1 EP-RM-004 Revision [X]Page 127 of 295 (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 HAl. It also does not apply to incidents that are accidental events, acts of civil disobedience, or otherwise are not a HOSTILE ACTION perpetrated by a HOSTILE FORCE. Examples include the crash of a small aircraft, shots from hunters, physical disputes between employees, etc.Reporting of these types of events is adequately addressed by other EALs, or the requirements of 10 CFR § 73.71 or 10 CFR § 50.72.Emergency plans and implementing procedures are public documents; therefore, EALs should not incorporate Security-sensitive information. This includes information that may be advantageous to a potential adversary, such as the particulars concerning a specific threat or threat location. Security-sensitive information should be contained in non-public documents such as the Susquehanna Physical Security SafeguaFd&Plan (ref. 1).Escalation of the emergency classification level would be via IC HG1.Susquehanna Basis Reference(s):

1. SSES Physical Security Plan 2. ON-O00-010 Security Event 3. NEI 99-01 HS1 Page 101 of 195 Attachment I EP-RM-004 Revision [X]Page 128 of 295 Category:

H -Hazards Subcategory: 1 -Security Initiating Condition: HOSTILE ACTION within the OWNER CONTROLLED AREA or airborne attack threat within 30 minutes EAL: HAI.1 Alert A HOSTILE ACTION is occurring or has occurred within the OWNER CONTROLLED AREA as reported by the Security Shift Supervision OR A validated notification from NRC of an aircraft attack threat within 30 min. of the site Mode Applicability: All Definition(s): HOSTILE ACTION -An act toward Susquehanna 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 Susquehanna. 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 -Includes the area within the expanded security perimeter, i.e., the areas that are bordered by the Vehicle Barriers System. The OWNER CONTROLLED AREA also includes the Monitored OWNER CONTROLLED AREA (MOCA) as defined in Security Procedures. Susquehanna Basis: Security Shift Supervision 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 SSES Physical Security Plan (Safeguards) information (ref. 1).NEI 99-01 Basis: This IC addresses the occurrence of a HOSTILE ACTION within the OWNER CONTROLLED AREA or notification of an aircraft attack threat. This event will require rapid response and assistance due to the possibility of the attack progressing to the PROTECTED AREA, or the need to prepare the plant and staff for a potential aircraft impact.Timely and accurate communications between the Security Shift Supervision and the Control Room is essential for proper classification of a security-related event.Page 102 of 195 Attachment 1 EP-RM-004 Revision [X]Page 129 of 295 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 Fewl Storage lInst-allation-S2ecuity ProGgram.As time and conditions allow, these events require a heightened state of readiness by the plant staff and implementation of onsite protective measures (e.g., evacuation, dispersal or sheltering). The Alert declaration will also heighten the awareness of Offsite Response Organizations (OROs), allowing them to be better prepared should it be necessary to consider further actions.This IC does not apply to incidents that are accidental events, acts of civil disobedience, or otherwise are not a HOSTILE ACTION perpetrated by a HOSTILE FORCE. Examples include the crash of a small aircraft, shots from hunters, physical disputes between employees, etc.Reporting of these types of events is adequately addressed by other EALs, or the requirements of 10 CFR § 73.71 or 10 CFR § 50.72.EAL-#4-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.EAL-#2The 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 Physical Security Plan (ref. 1).Susquehanna Basis Reference(s):

1. SSES Physical Security Plan 2. ON-000-010 Security Event 3. NEI 99-01 HA1 Page 103 of 195 Attachment 1 EP-RM-004 Revision [X]Page 130 of 295 Category:

H -Hazards Subcategory: 1 -Security Initiating Condition: Confirmed SECURITY CONDITION or threat EAL: HUI.1 Unusual Event A SECURITY CONDITION that does not involve a HOSTILE ACTION as reported by the Security Shift Supervision OR Notification of a credible security threat directed at the site OR A validated notification from the NRC providing information of an aircraft threat Mode Applicability: All Definition(s): 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 Susquehanna 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 Susquehanna. 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).Susquehanna Basis: This EAL is based on the SSES Physical Security Plan (ref. 1).Security Shift Supervision 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 SSES Physical Security Plan (Safeguards) information (ref. 1).NEI 99-01 Basis: This IC addresses events that pose a threat to plant personnel or SAFETY SYSTEM equipment, and thus represent a potential degradation in the level of plant safety. Security events which do not meet one of these EALs are adequately addressed by the requirements of 10 CFR § 73.71 Page 104 of 195 Attachment 1 EP-RM-004 Revision [X]Page 131 of 295 or 10 CFR § 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. Classification of these events will initiate appropriate threat-related notifications to plant personnel and Offsite Response Organizations. Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan-[and independent Spent Fuel Storage Installation SecUriy Pro gram.E-AL#-IThe first threshold references (site- Geeifle-the seeu4-iySecurity Shift Supervisionshift supe~visiei*4 because these are the individuals trained to confirm that a security event is occurring or has occurred. Training on security event confirmation and classification is controlled due to the nature of Safeguards and 10 CFR § 2.39 information. EAL-#aThe second threshold addresses the receipt of a credible security threat. The credibility of the threat is assessed in accordance with (site specific pr-.edu.e)the Susquehanna Physical Security Plan (ref. 1).EAL-#aThe 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 th.ee Susquehanna Physical Security Plan (ref., 1 )(sRto, peific pro.edure). 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 Susquehanna Physical Security Plan (ref. 1).Escalation of the emergency classification level would be via IC HA1.Susquehanna Basis Reference(s):

1. SSES Physical Security Plan 2. ON-000-010 Security Event 3. NEI 99-01 HU1 Page 105 of 195 Attachment 1 EP-RM-004 Revision [X]Page 132 of 295 Category:

H -Hazards and Other Conditions Affecting Plant Safety Subcategory: 2 -Seismic Event Initiating Condition: Seismic event greater than OBE levels EAL: HU2.1 Unusual Event Seismic event greater than OPERATING BASIS EARTHQUAKE (OBE) as indicated by seismic instrumentation in the Control Room recording level greater than an OBE Mode Applicability: All Definition(s): None Susquehanna Basis: Ground motion acceleration of 0.05g is the OPERATING BASIS EARTHQUAKE (OBE) for Susquehanna (ref. 1). Earthquake Monitoring Panel 0C696 provides indication of strong motion, OBE, or SAFE SHUTDOWN EARTHQUAKE (SSE) events for the Unit 1 Containment Foundation, Unit 2 Containment Foundation, and the ESW Pumphouse (as well as tape printout). Input from all six channels are recorded when a trigger initiates the system. A seismic event generally starts with an indication in the Control Room, Annunciator SEISMIC MON SYSTEM TRIGGERED (AR-016-001 window G06) on 0C653. The OBE is signaled by an LED illuminated green on the upper panel adjacent to the label, OPERATING BASIS EARTHQUAKE (ref. 2-4).If above seismic monitoring is inoperable, ensure on-shift Operations and Security crews are aware of the need to report vibratory ground motion (earthquakes) to the control room, and the need for control room staff to contact outside agencies to confirm level of earthquake per OP-099-002 and ON-000-002. 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. 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 Susquehanna. Provide the analyst with the following Susquehanna coordinates: 4105'2015 north latitude, 7608'566" west longitude (ref. 4). Alternatively, near real-time seismic activity can be accessed via the NEIC website: http://earthquake.usgs.govleqcenter/ NEI 99-01 Basis: This IC addresses a seismic event that results in accelerations at the plant site greater than those specified for an Operating Basis Earthquake (OBE). An earthquake greater than an OBE but less than a Safe Shutdown Earthquake (SSE) should have no significant impact on safety-related systems, structures and components; however, some time may be required for the plant Page 106 of 195 Attachment 1 EP-RM-004 Revision [X]Page 133 of 295 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 Director/Recovery Manager 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 SA-9SA8.Susquehanna Basis Reference(s):

1. TRM 3.3.2 Seismic Monitoring Instrumentation

2. FSAR Section 3.7a Seismic Design 3. ON-000-002 Natural Phenomena 4. OP-099-002 Seismic Monitoring System 5. NEI 99-01 HU2 Page 107 of 195 Attachment I EP-RM-004 Revision [X]Page 134 of 295 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 -Area within the station inner security fence (PROTECTED AREA Barrier)designated to implement the requirements of 10 CFR 73.Susquehanna Basis: If damage is confirmed visually or by other in-plant indications, the event may be escalated to an Alert under EAL CA6.1 or SA8.1.A tornado striking (touching down) within the Protected Area warrants declaration of an Unusual Event regardless of the measured wind speed at the meteorological tower. A tornado is defined as a violently rotating column of air in contact with the ground and extending from the base of a thunderstorm. NEI 99-01 Basis: This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant.EAL-#-IEAL HU3.1 addresses a tornado striking (touching down) within the PROTECTED AREA.EAL ff2 addresses flooding of a buildiRg From or area that results on operators isolating power to a SAFETY SYSTEM com~ponent duo to water level or other wetting conceFrn. Classification i also requlred if the water level or rolated wetting cau;es an automatic isolation of a SAFETY SYSTEM comnponent from its power source (e.g., a breaker or relay trip). TO Warrant classification, operability of the affected component must be rlequired by TlecGhniV iall fFr the current opeFating mnode.EAL #3 addresses a hazardous materials event originating at an offeite locnatien and Of sufficient magnitude to impede the movement of personnel within the PROTECTED AREA.EAL #4 addresses a hazardous event that causes an On site impedimnent to vehicle movement and significant enough to prohibit the plant staff fromn accessing the site using personal vehicles Examples of such an event include site floinemg caUsed by a hlrrFcane, heav rains up. river water releases, damA failure, etc., or an On site train derafilment blocking the access rod This EAL is not intended apply impediments uch as fo.g, snow, ice, Or vehicle breakdowns Or accidentS, but ratheF to more significant conditiGRG suc~h as the Hurrican Page 108 of 195 Attachment 1 EP-RM-004 Revision [X]Page 135 of 295 Androw strike on Turikey PiRnt iR 1992, the flooding around the Cooper Station dUFORg r h Midwest floods of 1993, Or the flooding alrund Ft. Calhoun Station in 2011 reAL.1 95 aaddresses (site specific description). Escalation of the emergency classification level would be based on ICs in Recognition Categories AR, F, S or C.Susquehanna Basis Reference(s):

1. NEI 99-01 HU3 Page 109 of 195 Attachment 1 EP-RM-004 Revision [X]Page 136 of 295 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. Susquehanna Basis: Refer to NPE 91 001, SSES IPE, to identify susceptible internal flooding areas (ref. 1).If the SAFETY SYSTEM component was operating at the time of isolation, EAL SA8.1 or CA6.1 may be applicable based on degraded SAFETY SYSTEM performance. This EAL addresses water entering a room or area faster than installed equipment is capable of removing but is not applicable to water spraying on equipment of a magnitude that does not meet the definition of FLOODING.NEI 99-01 Basis: This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant.EAL #1 addresses a tornado Ftriking (tuchinRg doWn) within the PROTECTEiD) AREA.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 caused by the FLOODING. Classification is also required if the water level or related wetting caused by the FLOODING causes an automatic isolation of a SAFETY SYSTEM component from its Page 110 of 195 Attachment 1 EP-RM-004 Revision [X]Page 137 of 295 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.EAL= #3 addrcssos a hzrosmaterials event originating at an offsite loca-tion-anRd of sgufficfient magritude to impede the n monvment of withiR the PROTECTED AREA.EAL #4 addresses a hazardo~us event that causes an; en site impediment to vehicle moevement 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 hurcane,h .r up river water releases, dam failure, etc., or an on site train derailment blocking the access rod This EAL is not itended apply to routine impediments such as fog, snOW, ice, r Vehicle bhreakdownsVR or accidents, but rather to mor~e significant conditions SUch; as the Hurrican~e Andrew strike on Turkey Point in 1992, the flooding arnd the oeeper Station during th Midwest floods of 1993, or the flooding around Ft. Calhoun Station in 2011-.EAL #6 addresses (site specific deGcription). Escalation of the emergency classification level would be based on ICs in Recognition Categories AR, F, S or C.Susquehanna Basis Reference(s):

1. NPE 91 001 SSES IPE 2. NEI 99-01 HU3 Page 111 of 195 Attachment I EP-RM-004 Revision [X]Page 138 of 295 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 measures such as temporary shielding, SCBAs or beyond Emergency Plan RWP dose extensions that are not routinely employed to access the room/area). PROTECTED AREA -Area within the station inner security fence (PROTECTED AREA Barrier)designated to implement the requirements of 10 CFR 73.Susquehanna Basis: As used here, the term "offsite" is meant to be areas external to the PROTECTED AREA.NEI 99-01 Basis: This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant.EAL #1 addresses a tornado striking (touching down) within the PROTECTED AREA.This EALi addresses floedig of a building room Or alsa that results in aperatnoe isolating power to a SAFETY SYSTEM component due to water level or other wetting concerns. Clas'sification is6 also required if the water lovel or related wettin case anautoatic isolation o-f a; SAFEFTY SYSTEM component fromn its power source (egabekro elay trip). To warrant classification, operability of the affected copnn mut be required by Technical Specific~ation for thoren oprting moede.EAL-#3This 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.EAL #4 addresses a hazardous event that causes an en site impediment t v.hicle movement and significant enough to prohibit the plant staff from accessing the cite using personal vehicles.Examples of SUch an event include site flo diAg caused by a hur.cne heavy rains, up river water releases, damn failure, etc., or an on site train derailment blocking the access red This EAL is not intended apply to routine impediments SUch as fog, BnOW, iGe, or vehicle beakdewns or accidents, but rather to more significant conditions such as the Hurric Page 112 of 195 Attachment 1 EP-RM-004 Revision [X]Page 139 of 295 Andrew strike on T-urkey Point inR 1992, the floodin~g around the Cooper Station during thee Midwest floopdsq of 19A93, or the flooding around Ft. Calhoun Station in 2011.E-=AL1 45 addresses (site specific description). Escalation of the emergency classification level would be based on ICs in Recognition Categories AR, F, S or C.Susquehanna Basis Reference(s):

1. NEI 99-01 HU3 Page 113 of 195 Attachment 1 EP-RM-004 Revision [X]Page 140 of 295 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 Susquehanna Basis: Events to consider are river flooding, hurricane, wind storms that block all of the multiple routes to get to the site.NEI 99-01 Basis: This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant.EAL #1 addresses a tornado striking (touching down) within the PROTECTEID AREA.This EAL addresses floodiRg of a building ro or , a'ra that results in operator-isolating power to a SAFETY SYSTEM ..ompenent due to water level or other wetting cOncerns. Class ifiation is, also required if the water level or related wettin cassa automatic isolatfion of a_ SAFFETY SYSTEM component from its power Source (e.g.q, a bhreakeir olr relay trip). To9 warrant classification, operability of the affected component must be requlired by Technical Specifcatiens for the curr.ent operatin. mode.EAL #3 addresses a hazardous materials event originating at anm ffeite l cation and of sufficient magnitude to impede the movement of personnel within the PROTECTED AREA.EAL44This 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.Page 114 of 195 Attachment 1 EP-RM-004 Revision [X]Page 141 of 295 Escalation of the emergency classification level would be based on ICs in Recognition Categories AR, F, S or C.Susquehanna Basis Reference(s):

1. NEI 99-01 HU3 Page 115 of 195 Attachment 1 EP-RM-004 Revision [X]Page 142 of 295 Category: Subcategory:

Initiating Condition: EAL: H -Hazards and Other Conditions Affecting Plant Safety 4 -Fire FIRE potentially degrading the level of safety of the plant 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 ED/RM should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Table H-1 Fire Areas* Control Structure* Diesel Generator Buildings" ESSW Pump House* Reactor Buildings* Turbine Buildings* ISFSI 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.Susquehanna 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. A single 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. If a fire is verified to be occurring by field report, the 15 minute time limit is from the original receipt of multiple fire detection alarms/indications or field confirmation of the single fire detection alarm.Page 116 of 195 Attachment 1 EP-RM-004 Revision [X]Page 143 of 295 A valid fire detection alarm and a valid fire suppression alarm in the same area are considered receipt of multiple fire alarms or indications. Independent fire detection or suppression alarms in Table H-1 areas that are in close proximity are treated as multiple indications of a fire.Table H-1 Fire Areas are based on the SSES Fire Protection Review Report (FPRR). Table H-1 Fire Areas include those structures containing functions and systems required for safe shutdown of the plant (SAFETY SYSTEMS). (ref. 1)NEI 99-01 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.EAL-#4-T4he-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 knowledge that a fires exists emergency declaration locsk starts at the time of a report from the field, receipt of multiple fire detection alarms or indications, or field confirmation of a single alarm or indication. The fire duration clock starts at the time that knowledge of a fire exists., that the initial alarm, , or report was received, and not the time that a subsequent vorif-catio-n. actsion war, pe~red.Similarly, the firo duration clock also starts at the time of receipt of the initial alarmn, indication Or This E=AL 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 ala~rm. U~pon receipt, operators Will take prompt actions to confirm the validity of a single fire a!arm. For EAL assessment purposes, the 30 minute clok starts at the time that the initial alarm was received, and not the time thata subsequent ecation an ,was ---i. pertrmed.A single fire alaFrm, absent o~ther indication(s) of a FIRE, may be indicative of equipment failure oasprous activation, and not an actual FIRE. ForF this reason, additional time is allowed to veiytevalidity of the alarm. The 30) m~inute peried is a reasonable amount of time to detrmneifan actual FIRE exists; however, after that time, and absent ifrtontoth contrar,', it is assumed that an; actual FIRE iinprogress. if an actual FIRE is verified by a report fromn the field, then EAL= #1 is immediately applicable, and the em~ergency must be declared if the FI=RE is not eodinguished within 1 5 minutes of the-report. if the alarm is verified to be due to an equipment failur or a .prius activation, and this verification occurs within 30 minutes of the receipt of the alam then. this EAL is Rot applicable and no emerrgency declaFrfaet is warra.t.d. In addition to a FIRE addressed by EAL #1 9r EAL #2, a FIRE within the plant PROTECTED AREA ;ot extinguished within 60 minutes may also potentially degrade the level of plant safety.Page 117 of 195 Attachment 1 EP-RM-004 Revision [X]Page 144 of 295 This basis extends to a FIRE occurring within the PROTECTED AREA of an ISFSI lcoated eutside the plant PROTECTED AREA. [Sentence for plantG with an !SF=SI outside the plant.P~et~ered AFea4 EAL #4 if a FIRE withiR the plant eoFr ISS [for plants with an outside the plant Protected Areal PROTECTED AREA is Of GUfficient size to r-equir aaepos by an effsite fire-fighting agency (e.g., a local town Fire [Depa~tment), then the le*e ofpatsfety is potentially degraded. The dispatch of an offsite firefighting agency to the site requires an emergency declaration only if it is needed to actively suppeot firefighting effo~ts because the fire is beyond the capability of the Fire Brigade to extinguish. Declaration is not necessary if the agencY resources are placed o stand by, or SUPPO~ting post extinuishment recov, orinetigation actienc.Basis Related DReuemet ,rmfl~erm AoveidoxRl, Appendix R to 10 CFR 50, states in pan:.-Criterion 3 of Appendix A to this pa~t specifies that "Structures, systems, and components impetant to safety shall be deigned and located to minimize, consisteRt with other safety requirements, the probability and effect Of fires and explosion. When considering the effects, of fire, those systems associated with achieving-and maintaining cafe shutdown conditions assume major to safe*t because damage to them caR lead to core damage resulting from loss Of coolant through boil off.Because fire may affect safe shutdown systems and because the less of function of systems' used to mitigate the consequences Of des-ign basis accide-nts n..de pest fire conditions does not peFr s impact pubIli safet the need to limit fire damage to systems required to achieve and maintain safe shutdeWn conditions is greater than the need to lenit fire damage to theoe systems requied to mitigate the consequences of design basis accidents. In additio, AppeRdix R to 10 CFiR 50, requires, amcong ether coniderations, the use of 1 hour fire barrier-s for the enclosure of cable and equipment and associated non safety circuits of one redundant train (G.2.G). As used in EAL #2, the 30 minutes to verify a single alaFrm 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 SARSA8.Susquehanna Basis Reference(s):

1. SSES FPRR Section 6.2 Fire Area Description

2. NEI 99-01 HU4 Page 118 of 195 Attachment 1 EP-RM-004 Revision [X]Page 145 of 295 H -Hazards and Other Conditions Affecting Plant Safety 4 -Fire FIRE potentially degrading the level of safety of the plant Category: Subcategory:

Initiating Condition: 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 (i.e., proved or disproved) within 30 min. of alarm receipt (Note 1)Note 1: The ED/RM should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Table H-1 Fire Areas* Control Structure* Diesel Generator Buildings* ESSW Pump House* Reactor Buildings" Turbine Buildings* ISFSI 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.Susquehanna Basis: The 30 minute requirement begins upon receipt of a single valid fire detection or fire suppression 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 within 30 minutes, classification shall be made based on EAL HU4.1.Page 119 of 195 Attachment 1 EP-RM-004 Revision [X]Page 146 of 295 Table H-1 Fire Areas are based on the SSES Fire Protection Review Report (FPRR). Table H-1 Fire Areas include those structures containing functions and systems required for safe shutdown of the plant (SAFETY SYSTEMS). (ref. 1)NEI 99-01 Basis: This IC addresses the magnitude and extent of FIRES that may be indicative of a potential degradation of the level of safety of the plant.The intent of the 15 minute duration is to size the FIRE and to discriminate against small FIRES that 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 f a suppresssion system, etc,.Upon receipt, operators will take prompt actiens to c"nfirM the validity ef an initial fre alarm, indicatien, or report. For EAL assessment purposes, the emergency declaration clock 6tarts at the time that the initial alarm, indication, or report was received, and net the time that a subsequent verification action war, Similarly, the fire duration clk o1k aIs1 st ats at the time of receipt of the initial alarm, indication or report-.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 EAL-#4HU4.1 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. EAt-#13 In addition to a FIRE addressed by EAL #1 or EAL #2, a FIRE within the plant PROTECTED AREA not extRnguished within 60 minutes may also potentially degrade the level of plant safety.This basis exten~ds to a F4RE= GGG.Rring w,'th,'n the PROTECTED0 AREA of an !SFS located outskde the plant PROTECTED AREA4. [Sentence for-plants with an !WS!S outside the plant P-Fete~ted AFea]EAL. #4 If a FIRE within the plant or ISS, [for plants with an ISFS! outside the plant Protected Area]PROTCTEDAREAis Of sufficent size to require a response by an off-site firefighting (e.g., a lcal town Fire Department), then the level of plant safey is potentially Page 120 of 195 Attachment 1 EP-RM-004 Revision [X]Page 147 of 295 degradcd. The dispatch of an offsite firefighting agency to the site requires an emergency RoIny if it is needed to actively suppoh t firefighting efforts because the fir' is beyond the Ecapabilit' of the Fire Brigade to extinguish. Declaration is not necessary if the agencGY resources are placed on staRd by, or supporting post etinguishment receF aGtieR&.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 EAL-#2HU4.2, 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 SA-gSA8.Susquehanna Basis Reference(s):

1. SSES FPRR Section 6.2 Fire Area Description

2. NEI 99-01 HU4 Page 121 of 195 Attachment 1 EP-RM-004 Revision [X]Page 148 of 295 Category: Subcategory:

Initiating Condition: EAL: H -Hazards and Other Conditions Affecting Plant Safety 4 -Fire FIRE potentially degrading the level of safety of the plant 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 ED/RM 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 -Area within the station inner security fence (PROTECTED AREA Barrier)designated to implement the requirements of 10 CFR 73.Susquehanna Basis: None NEI 99-01 Basis: This IC addresses the magnitude and extent of FIRES that may be indicative of a potential degradation of the level of safety of the plant.The intent of the 15 mirute duration is to size the FIRE and to discriminate against small FIRES that arc readily extinguished (e.g., smoldering waste paper basket). in addition to alarms, other indications of a FIRE could be a drop in fire mnain prossure, automatic activation of a suppr8ssion system, otG.Upon receipt, operators will take promnpt actions to confirm the validity of an initial fire alam indication, or report. For E=AL assessment purposes, the emergencGy declaration clock starts at the tome that the initial alarm, inRdication, Or report was received, and not the time that a subsequent verfification; action was pe~fermed. Similarly, the fire duration clock also starts at the time Of rFeGeipt of the iRitial alarm, indicatieR Or E-AL-#2-This EAL addresses of a fire alarm, and the existence of a FIRE is not verified (i.e., proved or disproved) within 30 mninutes of the alarmn. Upon receipt, operators will take prompt actions to eonfirm the validity of a single fire alarm. For E=AL assessment purposes, the Page 122 of 195 Attachment 1 EP-RM-004 Revision [XJ Page 149 of 295 30 minutoR cock starts at the tome that the initial alarm war-,' rGeeved, and ant the time th, subsequent .verficatien acton. was performed. A single fire alarm, absent other indication(6) of a FIRE, May be indicativo of equipment failuree or a spurious activation, and not an; actual FIRE. For this reason, addition~al time is allowed to verify the validity of the ala~rm. The 30 minute period is a reasonable amonGUt of time to determine if an actual FIRE exists; however, after that time, and absent information to the cOntrar.', it is; assu..med that an actUal FIRE is in progress.If an actual FIRE is verified by a report from the field, then EAL #1 is immediately applicable, and the emergency must be declared if the FIRE is net extinguished within 15 m~inutes Of theif the alarm is verified to be due to an equipment failure or.a .purius activation, and this verification occurs within 30 minuters of the receipt of the alam then this EAL is Rot applicable and no emergency declaration is warranted. In addition to a FIRE addressed by EAL #4-HU4.1 or JE.A#2HU4.2, a FIRE within the plant PROTECTED AREA not extinguished within 60-minutes may also potentially degrade the level of plant safety. This basis extends to a F4RE ecc-rRing within the PROTECTED AREA ef an 19SFS! located outside the plant PRO TECTE-D AREA. [Sentence for- plants with an ISFS( outsid the plant Protected Aree]if a FI=RE within the plant or- ISFS! [for- plants withf an ISFSI outside the plant Protected Area]PROTECTED AREA is Of sufficient size to require a respoe by an off-site firefighting agencY (e.g., a lecal town Fire tnRt), then the leel of plant safety is potentially degraded. The dspatch ef an cffsete firefightig th site equires an emergeRny declaration only if it needed to actively support firefighting effeots because the fire is beyond the apability of the Fire Brigade to extinguish. Declaration is no~t necessary if the agency resources are placedo stand by, or supporting post extinguishment reG or, i4nvestation actions.Or. Redt. aseds R. rn, fcren-dnts. Appendix R to 10 CF=R 50, states, in part;Criterion 3 of Appendix A to this part specifies that "'Structures, systems, and comRponentS impor-tant to safety shall be designed and located to miiie iosistent with other safety requirements, the probability and effect 9f fires adepoin.When considering the effects of fire, those systems associated with achieving and*maintaining safe shutdoWn conditions, assume major im;portance to safet because damage to them can lead to core damage resulting ferom loss of coolant through boil off.Because fire may affect Safe shutdown systemns and because the lorss of function ot*systems used to mitigate the consequences of design basis accidents under post fire conditions does not per se im~pact public safet, the need to limit Fie damage to systems required to achieve and maintain safe shutdown conditions is greater than the need to li~mit fire damage to those systems requir~ed to mnitigate the consequences of design basis accidents. in addition, Appendix R to 10 CFIR 50, requires, among other considerations, the use of 1 hour fire barriers for the enclosure Of cable and equipment and assocr-i-ated non safety ciF~rcuit of one Page 123 of 195 Attachment 1 EP-RM-004 Revision [X]Page 150 of 295 redundant tr-ain (G.2.c). As used in EAL #2, the 30 minutes to verif' a single alarm is well within this worst case 1 hour time poriod-.Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via IC CA6 or SA89.Susquehanna Basis Reference(s):

1. NEI 99-01 HU4 Page 124 of 195 Attachment 1 EP-RM-004 Revision [X]Page 151 of 295 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 -Area within the station inner security fence (PROTECTED AREA Barrier)designated to implement the requirements of 10 CFR 73.Susquehanna Basis: None NEI 99-01 Basis: This IC addresses the magnitude and extent of FIRES that may be indicative of a potential degradation of the level of safety of the plant.The intent of the 15 minute duration is to size the FIRE and to discrimin!ate against small FIRES th-atare readily e~ingwished (e.g., smoaldering waste paper basket). In addition to alarms, other indications of a FIRE could be a drop fin fire main pressure, automatic activation of a suppReio system, etc.-Upn acipt, operators will take prompt actions to confirm the validity of an initial fire alarm, indicatfion, or report. For EAL assessment purposes, the emnergency declaration clock starts at.the time that the initial alarm, indication, or report was received, and net the time that a subsequent verification action was pe~fermed. Similarly, the fire duration clock also starts at the time Of receipt of the initial alarmA, indication Or report-.E-AL2 This EAL addresses, receipt Of a single fire alarm, and the existence of a FIRE is not Verife (i.e., proved or disro;ved) within 30 Minutes of tlhe alarm. UpoR operaters will tak prompt actions to coRfirm the validity of a single fire a!arm. For EAL assessmnent 30 min~ute clock starts at the time that the initial alarm was received, and Rot the time tha subsequent verification action was, pe~rfemed. the Page 125 of 195 Attachment 1 EP-RM-004 Revision [X]Page 152 of 295 A single fire alarm, absent other indication(s) of a FIRE, may be indicative of equipment failure or a spurious activatien, and not an actual FIRE. F this reason, additional time is allowed to verif; the validit; of the alarr. The 30 minute us a reasonable amount of time to determ~ine if an actual FIRE exists; however, after that time, and absent information to theit is assumed that an actual FIRE is in prOgress.if an actual FIRE is verified by a report from the field, then EAL #1 is immediately apptiabe, and the emergency must be declared if the FIRE is net extinguished within 15. mpinutes of the repot. if the alarm is verified to be duo to an equipment failure or a spurius activation, and this veificatiGR occurs within 30 minutes of the receipt of the alam thnthsEL snt applicable and no emergency declaration is Wa~rranted. In addition to a FIRE addressed by EAL #1 or EAL #2, a FIRE within the plant PRCTectEDFm AREA not extinguished within 60 minutes may also potentially degrade the level of plant safety.dhis basis extendfs te a FiREf cgh Uing i thin the s ROTECTEsD AREA of an dSFrSi l ycate outside the plant PROTECpTrED AREA. [Sentence fo p fants With an ISFS! otheapab the poant Prete~ted A~ea]E-AL#4 If a FIRE within the plant or ISF=SI [for-plants with an ISF-Si outside the plant Protected Area]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.Basis-Related Requremen+a, ts from nnanrlh, V Appendix R to 10 CFR 50, states in part: Criterion 3 of Appendix A to this part specifies that "Structures, systems, and compon.ents important to safety shall be designed and located to minimize, ,onsistent with other safety requirements, the probability and effect of fiFesandeplosions When consARidering the effects of fire, those systems associated with achieving and maintaining safe shutdown conditions assume major impo.tance to safety because damage to them can lead to core damage resulting from less Of coolant through boil off.Because fire may affect safe shutdown systems and because the loss Of function 0t systems used to mitigate the conseque-ncesc-of d;e-sign; basis accidents under pest fire onn;diti9;nn does not per se impact public safety, the need to limit fire damage to systems to ahieve adnd maintain safe shutdown conditions is greater than the need to limit fire damage to these systems required to mitigate the consequences of design In addition, Appendix R to 10 CFR 50, reqluires, among otheFr considerations, the use of I hour fire ba-rriers fr the enclosure of cnable and equipment and associated non safety of one Page 126 of 195 Attachment 1 EP-RM-004 Revision [X]Page 153 of 295 ed-undant train (G.2.c). As used in EAL= #2, the 30 mninutes to verify a single alarm is well within thiG wors-Ft case 1 hour ti me period.Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via IC CA6 or SA80.Susquehanna Basis Reference(s):

1. NEI 99-01 HU4 Page 127 of 195 Attachment 1 EP-RM-004 Revision [X]Page 154 of 295 Category: Subcategory:

Initiating Condition: H -Hazards and Other Conditions Affecting Plant Safety 5 -Hazardous Gases Gaseous release IMPEDING access to equipment necessary for normal plant operations, cooldown or shutdown EAL: HA5.1 Alert Release of a toxic, corrosive, asphyxiant or flammable gas into any Table H-2 rooms or areas AND Entry into the room or area is prohibited or IMPEDED (Note 5)Note 5: If the equipment in the listed room or area was already inoperable or out-of-service before the event occurred, then no emergency classification is warranted. Table H-2 Safe Operation & Shutdown Areas Elevation Unit 1 Area(s) ** Unit 2 Area(s)** Mode(s)670' RB 27 RB 32 3,4,5 683' RB 27, 28, 29 RB 32, 33, 34 3,4,5 703' RB 28, 29 RB 33, 34 3,4,5 719' RB 25, 29 RB 30, 34 3,4,5 749' RB 25, 29 RB 32, 33 3,4,5 729' CS 12, 21 CS 12, 21 1, 2, 3, 4, 5, D** See Chart I for location of plant areas Chart 1- Plant Area Key Plan LOW LEVEL WATER TREATMENT RADWASTE F7-I CHLORINE 1 3.L2LL51 5 0 47 BUILDING PUMPHOUSE UNIT#2 ? 2 ?UNI1 TURBINE BLDG. TURBINE BLDG.B SPRAY POND VALVE VAULT ESSW H51 PUMPHOUSE RADWASTE TURBe 1 16 1 15 14 1 13 4 13 2 1 38 I37 I-I-4--4 --- I -I-- -t -I L......J , 1 20 1 19 18 1 17 8 17 6 524 23-I COND. STORAGE-'R IVER INTAKE STRUCTURE 22 121 1 12 11 10 9 40 39 42 41 i ._______ I -.9- .9 36135 I COND&REF STORAGE 341 3312912. 44 43 ,-DIESEL GENERATOR E62 ~ #1 REACTOR REACTOR E DIESEL 81 GENERATOR Page 128 of 195 Attachment 1 EP-RM-004 Revision [X]Page 155 of 295 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 measures such as temporary shielding, SCBAs or beyond Emergency Plan RWP dose extensions that are not routinely employed to access the room/area). Susquehanna Basis: If the equipment in the listed room or area was already inoperable, or out-of-service, before the event occurred, then no emergency should be declared since the event will have no adverse impact beyond that already allowed by Technical Specifications at the time of the event.The list of plant areas in Table H-2 specify those areas that contain equipment which require a manual/local action as specified in operating procedures used for normal plant operation, cooldown and shutdown. Areas in which actions of a contingent or emergency nature would be performed (e.g., an action to address an off-normal or emergency condition such as emergency repairs, corrective measures or emergency operations) are not included. In addition, the list specifies the plant mode(s) during which entry would be required for each room or area. See Chart 1 for the specific locations of areas listed in Table R-2. See Attachment 3 for more details of how the Table R-2 was developed (ref. 1).NEI 99-01 Basis: This IC addresses an event involving a release of a hazardous gas that precludes or impedes access to equipment necessary to maintain normal plant operation, or required for a normal plant cooldown and shutdown. This condition represents an actual or potential substantial degradation of the level of safety of the plant.An Alert declaration is warranted if entry into the affected room/area is, or may be, procedurally required during the plant operating mode in effect at the time of the gaseous release. The emergency classification is not contingent upon whether entry is actually necessary at the time of the release.Evaluation of the IC and EAL do not require atmospheric sampling; it only requires the Emergency Director's/Recover Manager's judgment that the gas concentration in the affected room/area is sufficient to preclude or significantly impede procedurally required access. This judgment may be based on a variety of factors including an existing job hazard analysis, report of ill effects on personnel, advice from a subject matter expert or operating experience with the same or similar hazards. Access should be considered as impeded if extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., requiring use of protective equipment, such as SCBAs, that is not routinely employed). An emergency declaration is not warranted if any of the following conditions apply:* The plant is in an operating mode different than the mode specified for the affected room/area (i.e., entry is not required during the operating mode in effect at the time of the gaseous release). For example, the plant is in Mode 1 when the gaseous release occurs, Page 129 of 195 Attachment I EP-RM-004 Revision [X]Page 156 of 295 and the procedures used for normal operation, cooldown and shutdown do not require entry into the affected room until Mode 4." The gas release is a planned activity that includes compensatory measures which address the temporary inaccessibility of a room or area (e.g., fire suppression system testing).* The action for which room/area entry is required is of an administrative or record keeping nature (e.g., normal rounds or routine inspections)." The access control measures are of a conservative or precautionary nature, and would not actually prevent or impede a required action." If the equipment in the listed room or area was already inoperable, or out-of-service, before the event occurred, then no emerqency should be declared since the event will have no adverse impact beyond that already allowed by Technical Specifications at the time of the event.An asphyxiant is a gas capable of reducing the level of oxygen in the body to dangerous levels.Most commonly, asphyxiants work by merely displacing air in an enclosed environment. This reduces the concentration of oxygen below the aer-mal-IDLH level of around 19.5%, which can lead to breathing difficulties, unconsciousness or even death.This EAL does not apply to firefighting activities that automatically or manually activate a fire suppression system in an area, or to intentional inerting of containment(£VAR-elp .Escalation of the emergency classification level would be via Recognition Category A R, C or F ICs.Susquehanna Basis Reference(s):

1. Attachment 3 Safe Operation

& Shutdown Areas Tables R-2 & H-2 Bases 2. NEI 99-01 HA5 Page 130 of 195 Attachment 1 EP-RM-004 Revision [X]Page 157 of 295 H -Hazards and Other Conditions Affecting Plant Safety 6 -Control Room Evacuation Inability to control a key safety function from outside the Control Room Category: Subcategory: Initiating Condition: EAL: HS6.1 Site Area Emergency An event has resulted in plant control being transferred from the Control Room to the Remote Shutdown Panels AND Control of any of the following key safety functions is not reestablished within 15 min. (Note 1): " Reactivity" RPV water level" RCS heat removal Note 1: The ED/RM should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Mode Applicability: All Definition(s): None Susquehanna Basis: The Shift Manager determines if the Control Room is uninhabitable and requires evacuation. Control Room inhabitability may be caused by fire, dense smoke, noxious fumes, bomb threat in or adjacent to the Control Room, or other life threatening conditions (Ref. 1, 2).NEI 99-01 Basis: This IC addresses an evacuation of the Control Room that results in transfer of plant control to alternate locations, and the control of a key safety function cannot be reestablished in a timely manner. The failure to gain control of a key safety function following a transfer of plant control to alternate locations is a precursor to a challenge to one or more fission product barriers within a relatively short period of time.The determination of whether or not "control" is established at the remote safe shutdown location(s) is based on Emergency Director/Recovery Manager judgment. The Emergency Director/Recovery Managqer is expected to make a reasonable, informed judgment within (the site speGcific timc for transfer)j5 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 Susquehanna Basis Reference(s): Page 131 of 195 Attachment 1 EP-RM-004 Revision [X]Page 158 of 295 1. ON-100(200)-009 Control Room Evacuation

2. NEI 99-01 HS6 Page 132 of 195 Attachment 1 EP-RM-004 Revision [X]Page 159 of 295 Category:

H -Hazards and Other Conditions Affecting Plant Safety Subcategory: 6 -Control Room Evacuation Initiating Condition: Control Room evacuation resulting in transfer of plant control to alternate locations EAL: HA6.1 Alert An event has resulted in plant control being transferred from the Control Room to the Remote Shutdown Panels Mode Applicability: All Definition(s): None Susquehanna Basis: The Shift Manager (SM) determines if the Control Room is uninhabitable and requires evacuation. Control Room inhabitability may be caused by fire, dense smoke, noxious fumes, bomb threat in or adjacent to the Control Room, or other life threatening conditions (Ref. 1, 2).Inability to establish plant control from outside the Control Room escalates this event to a Site Area Emergency per EAL HS6.1.NEI 99-01 Basis: This IC addresses an evacuation of the Control Room that results in transfer of plant control to alternate locations outside the Control Room. The loss of the ability to control the plant from the Control Room is considered to be a potential substantial degradation in the level of plant safety.Following a Control Room evacuation, control of the plant will be transferred to alternate shutdown locations. The necessity to control a plant shutdown from outside the Control Room, in addition to responding to the event that required the evacuation of the Control Room, will present challenges to plant operators and other on-shift personnel. Activation of the ERO and emergency response facilities will assist in responding to these challenges. Escalation of the emergency classification level would be via IC HS6.Susquehanna Basis Reference(s):

1. ON-100(200)-009 Control Room Evacuation

2. NEI 99-01 HA6 Page 133 of 195 Attachment 1 EP-RM-004 Revision [X]Page 160 of 295 Category:

H -Hazards and Other Conditions Affecting Plant Safety Subcategory: 7 -ED/RM Judgment Initiating Condition: Other conditions exist which in the judgment of the Emergency Director/Recovery Manager warrant declaration of a General Emergency EAL: HG7.1 General Emergency Other conditions exist which in the judgment of the Emergency Director/Recovery Manager 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 Susquehanna 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 Susquehanna. 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.As applied to Susquehanna, CONTAINMENT CLOSURE is established per NDAP-QA-0309 (ref. 2) for Primary Containment OR is established per NDAP-QA-0321 (ref. 3) for Secondary Containment. Susquehanna Basis: The Emergency Director/Recovery Manager is the designated onsite individual having the responsibility and authority for implementing the Emergency Response Plan. The Shift Manager(SM) initially acts in the capacity of the Emergency Director/Recovery Manager and takes actions as outlined in the Emergency Plan implementing procedures. If required by the emergency classification or if deemed appropriate by the Emergency Director/Recovery Manager, 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 Page 134 of 195 Attachment 1 EP-RM-004 Revision [X]Page 161 of 295 anticipation of the possible wide-ranging responsibilities associated with managing a major emergency (ref. 1).NEI 99-01 Basis: This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director/Recovery Manager to fall under the emergency classification level description for a General Emergency. Susquehanna Basis Reference(s):

1. Susquehanna LLC, Susquehanna Steam Electric Station Emergency Plan, Section 6.0 Organizational Control of Emergencies

2. NDAP-QA-0309 Primary Containment Access and Control 3. NDAP-QA-0321 Secondary Containment Integrity Control 4. NEI 99-01 HG7 Page 135 of 195 Attachment 1 EP-RM-004 Revision [X]Page 162 of 295 Category:

H -Hazards and Other Conditions Affecting Plant Safety Subcategory: 7 -ED/RM Judgment Initiating Condition: Other conditions existing that in the judgment of the Emergency Director/Recovery Manager warrant declaration of a Site Area Emergency EAL: HS7.1 Site Area Emergency Other conditions exist which in the judgment of the Emergency Director/Recovery Manager 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 EMERGENCY PLAN BOUNDARY Mode Applicability: All Definition(s): HOSTILE ACTION -An act toward Susquehanna 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 Susquehanna. 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).EMERGENCY PLAN BOUNDARY (EPB) -Same as the Exclusion Area Boundary, i.e., that area around SSES within a radius of 1800 feet determined in accordance with 10 CFR 100.11.Susquehanna Basis: The Emergency Director/Recovery Manager is the designated onsite individual having the responsibility and authority for implementing the Emergency Response Plan. The Shift Manager (SM) initially acts in the capacity of the Emergency Director/Recovery Manager and takes actions as outlined in the Emergency Plan implementing procedures. If required by the emergency classification or if deemed appropriate by the Emergency Director/Recovery Manager, emergency response personnel are notified and instructed to report to their emergency response locations. In this manner, the individual usually in charge of activities in the Control Room is responsible for initiating the necessary emergency response, but Plant Management is expected to manage the emergency response as soon as available to do so in anticipation of the possible wide-ranging responsibilities associated with managing a major emergency (ref. 1).Page 136 of 195 Attachment 1 EP-RM-004 Revision [X]Page 163 of 295 NEI 99-01 Basis: This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director/Recovery Manager to fall under the emergency classification level description for a Site Area Emergency. Susquehanna Basis Reference(s):

1. Susquehanna LLC, Susquehanna Steam Electric Station Emergency Plan, Section 6.0 Organizational Control of Emergencies

2. NEI 99-01 HS7 Page 137 of 195 Attachment 1 EP-RM-004 Revision [X]Page 164 of 295 Category:

H -Hazards and Other Conditions Affecting Plant Safety Subcategory: 7 -ED/RM Judgment Initiating Condition: Other conditions exist that in the judgment of the Emergency Director/Recovery Manager warrant declaration of an Alert EAL: HA7.1 Alert Other conditions exist which, in the judgment of the Emergency Director/Recovery Manager, 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 Susquehanna 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 Susquehanna. 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).Susquehanna Basis: The Emergency Director/Recovery Manager is the designated onsite individual having the responsibility and authority for implementing the Emergency Response Plan. The Shift Manager (SM) initially acts in the capacity of the Emergency Director/Recovery Manager and takes actions as outlined in the Emergency Plan implementing procedures. If required by the emergency classification or if deemed appropriate by the Emergency Director/Recovery Manager, emergency response personnel are notified and instructed to report to their emergency response locations. In this manner, the individual usually in charge of activities in the Control Room is responsible for initiating the necessary emergency response, but Plant Management is expected to manage the emergency response as soon as available to do so in anticipation of the possible wide-ranging responsibilities associated with managing a major emergency (ref. 1).NEI 99-01 Basis: This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director/Recovery Mananer to fall under the emergency classification level description for an Alert.Page 138 of 195 Attachment 1 EP-RM-004 Revision [X]Page 165 of 295 Susquehanna Basis Reference(s):

1. Susquehanna LLC, Susquehanna Steam Electric Station Emergency Plan, Section 6.0 Organizational Control of Emergencies

2. NEI 99-01 HA7 Page 139 of 195 Attachment 1 EP-RM-004 Revision [X]Page 166 of 295 Category:

H -Hazards and Other Conditions Affecting Plant Safety Subcategory: 7 -ED/RM Judgment Initiating Condition: Other conditions existing that in the judgment of the Emergency Director/Recovery Manager warrant declaration of a UE EAL: HU7.1 Unusual Event Other conditions exist which in the judgment of the Emergency Director/Recovery Manager indicate that events are in progress or have occurred which indicate a potential degradation of the level of safety of the plant or indicate a security threat to facility protection has been initiated. No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of SAFETY SYSTEMS occurs.Mode Applicability: All Definition(s): SAFETY SYSTEM- A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2): Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the reactor coolant pressure boundary;(2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures. Susquehanna Basis: The Emergency Director/Recovery Manager is the designated onsite individual having the responsibility and authority for implementing the Emergency Response Plan. The Shift Manager (SM) initially acts in the capacity of the Emergency Director/Recovery Manager and takes actions as outlined in the Emergency Plan implementing procedures. If required by the emergency classification or if deemed appropriate by the Emergency Director/Recovery Manager, emergency response personnel are notified and instructed to report to their emergency response locations. In this manner, the individual usually in charge of activities in the Control Room is responsible for initiating the necessary emergency response, but Plant Management is expected to manage the emergency response as soon as available to do so in anticipation of the possible wide-ranging responsibilities associated with managing a major emergency (ref. 1).NEI 99-01 Basis: This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Page 140 of 195 Attachment 1 EP-RM-004 Revision [X]Page 167 of 295 Director/Recovery Manaqer to fall under the emergency classification level description for an NGLJEUnusual Event.Susquehanna Basis Reference(s):

1. Susquehanna LLC, Susquehanna Steam Electric Station Emergency Plan, Section 6.0 Organizational Control of Emergencies, 2. NEI 99-01 HU7 Page 141 of 195 Attachment 1 EP-RM-004 Revision [XI Page 168 of 295 Category S -System Malfunction EAL Group: Hot Conditions (RCS temperature

> 200 0 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 4.16 kV ESS 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 power to or degraded voltage on the 125 VDC vital buses.3. Loss of Control Room Indications Certain events that degrade plant operator ability to effectively assess plant conditions within the plant warrant emergency classification. Losses of indicators are in this subcategory.

4. RCS Activity During normal operation, reactor coolant fission product activity is very low. Small concentrations of fission products in the coolant are primarily from the fission of tramp uranium in the fuel clad or minor perforations in the clad itself. Any significant 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 pressure vessel provides a volume for the coolant that covers the reactor core.The reactor pressure vessel and associated pressure piping (reactor coolant system)together provide a barrier to limit the release of radioactive material should the reactor fuel clad integrity fail. Excessive RCS leakage greater than Technical Specification limits indicates potential pipe cracks that may propagate to an extent threatening fuel clad, RCS and Primary Containment integrity.

6. RPS Failure This subcategory includes events related to failure of the Reactor Protection System (RPS)to initiate and complete reactor scrams. In the plant licensing basis, postulated failures of the RPS to complete a reactor scram comprise a specific set of analyzed events referred to Page 142 of 195 Attachment 1 EP-RM-004 Revision [X]Page 169 of 295 as Anticipated Transient Without Scram (ATWS) events. For EAL classification, however, ATWS is intended to mean any trip failure event that does not achieve reactor shutdown.

If RPS actuation fails to assure reactor shutdown, positive control of reactivity is at risk and could cause a threat to fuel clad, RCS and Primary Containment integrity.

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

8. Hazardous Event Affecting Safety Systems Certain hazardous natural and technological events may result in visible damage to or degraded performance of safety systems warranting classification.

Page 143 of 195 Attachment 1 EP-RM-004 Revision [X]Page 170 of 295 Category: S -System Malfunction Subcategory: 1 -Loss of Essential AC Power Initiating Condition: Prolonged loss of all offsite and all onsite AC power to essential buses EAL: SGI.1 General Emergency Loss of ALL offsite and ALL onsite AC power capability to ALL 4.16 kV ESS buses on EITHER unit AND EITHER: Restoration of at least one 4.16 kV ESS bus in < 4 hours is not likely (Note 1)OR RPV water level CANNOT BE RESTORED AND MAINTAINED > -179 in.Note 1: The ED/RM should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Mode Applicability: 1 -Power Operations, 2 -Startup, 3 -Hot Shutdown Definition(s): None Susquehanna Basis: The Class 1 E 4.16 kV system supplies all the Engineered Safety Feature (ESF) loads and other loads that are needed for a safe and orderly plant shutdown, and for keeping the plant in a safe shutdown condition. The eight Class 1 E 4.16 kV ESS Buses 1(2)A through 1(2)D receive power from either the four ESS 13.8/4.16 kV transformers or the diesel generators (A, B, C, D and additional diesel generator E). Buses 1A-1D supply Unit 1 and common loads and Buses 2A-2D supply Unit 2 loads. This configuration prevents a loss of all ESS Buses for one unit in the event one of the ESS Transformers is lost.During normal plant operation, ESS Transformer 101 supplies preferred power to ESS Bus 1A and 2A and is an alternate power supply to ESS bus 1D and 2D. ESS Transformer 111 supplies preferred power to ESS Bus 1C and 2C, and is an alternate power supply to ESS bus 1 B and 2B. ESS Transformer 201 supplies preferred power to ESS Bus 1D and 2D, and is an alternate power supply to ESS bus 1A and 2A. ESS Transformer 211 supplies preferred power to ESS Bus 1B and 2B, and is an alternate power supply to ESS bus 1C and 2C.On a loss of a preferred power source, the bus rapidly transfers to the alternate power source to maintain component power. If both the preferred and alternate power sources are lost, the associated standby diesel generator connects to the ESS bus. (ref. 2-6)Four hours is the station blackout coping time (ref. 7).Page 144 of 195 Attachment 1 EP-RM-004 Revision [X]Page 171 of 295 Indication of continuing core cooling degradation is manifested by the inability to restore and maintain RPV water level above the Minimum Steam Cooling Reactor Water Level (-179 in.)(ref. 8). Core submergence is the most desirable means of core cooling, however when RPV level is below TAF, the uncovered portion of the core can be cooled by less reliable means (i.e., steam cooling or spray cooling).NEI 99-01 Basis: 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.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.Susquehanna Basis Reference(s):

1. FSAR Section 8.2 Offsite Power System 2. FSAR Section 8.3 Onsite Power System 3. Technical Specifications 3.8.1 AC Sources -Operating 4. Technical Specifications 3.8.7 Distribution System -Operating 5. ON-104 (204)-001 Units 1(2) Response to Loss of All Offsite Power 6. EO-100 (200)-030 UNIT 1(2) Response to Station Blackout 7. FSAR Section 15.9 STATION BLACKOUT (SBO)8. EO-000-102 RPV Control 9. NEI 99-01 SGI Page 145 of 195 Attachment 1 EP-RM-004 Revision [X]Page 172 of 295 Category: Subcategory:

Initiating Condition: S -System Malfunction 1 -Loss of Essential AC Power Loss of all essential AC and vital DC power sources for 15 minutes or longer EAL: SG1.2 General Emergency Loss of ALL offsite and ALL onsite AC power capability to ALL 4.16 kV ESS buses on EITHER unit for -- 15 min.AND Indicated voltage is < 105 VDC on ALL of the following vital 125 VDC main distribution buses on the affected unit for -15 min. (Note 1): " 1D612 (2D612)* 1D622(2D622)

1D632 (2D632)* 1D642 (2D642)Note 1: The ED/RM should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Mode Applicability:

1 -Power Operations, 2 -Startup, 3 -Hot Shutdown Definition(s): None Susquehanna Basis: This EAL addresses operating experience from the March 2011 accident at Fukushima Daiichi.The Class 1E 4.16 kV system supplies all the Engineered Safety Feature (ESF) loads and other loads that are needed for a safe and orderly plant shutdown, and for keeping the plant in a safe shutdown condition. The eight Class 1E 4.16 kV ESS Buses 1(2)A through 1(2)D receive power from either the four ESS 13.8/4.16 kV transformers or the diesel generators (A, B, C, D and additional diesel generator E). Buses 1A-1D supply Unit 1 and common loads and Buses 2A-2D supply Unit 2 loads. This configuration prevents a loss of all ESS Buses for one unit in the event one of the ESS Transformers is lost.During normal plant operation, ESS Transformer 101 supplies preferred power to ESS Bus 1A and 2A and is an alternate power supply to ESS bus 1D and 2D. ESS Transformer 111 supplies preferred power to ESS Bus 1C and 2C, and is an alternate power supply to ESS bus I B and 2B. ESS Transformer 201 supplies preferred power to ESS Bus 1D and 2D, and is an alternate power supply to ESS bus 1A and 2A. ESS Transformer 211 supplies preferred power to ESS Bus 1B and 2B, and is an alternate power supply to ESS bus 1C and 2C.Page 146 of 195 Attachment 1 EP-RM-004 Revision [X]Page 173 of 295 On a loss of a preferred power source, the bus rapidly transfers to the alternate power source to maintain component power. If both the preferred and alternate power sources are lost, the associated standby diesel generator connects to the ESS bus. (ref. 1-6)The Class 1E Battery Banks are 1(2)D610 (Channel A), 1(2)D620 (Channel B), 1(2)D630 (Channel C), and 1(2)D640 (Channel D). Each bank consists of 60 cells connected in series.Each cell produces a nominal voltage of 2.06 VDC resulting in a total battery bank terminal voltage of 123.6 VDC. All battery banks are designed to supply power to its load center for four hours in the event of a loss of power from its battery charger (ref. 7-9).105 VDC is the minimum design voltage limit (ref. 10).Indicated voltage for the vital 125 VDC main distribution buses is local only. Local voltage indication is available for each bus based on dispatching a field operator in accordance with Control Room alarm response procedure AR-1(2)06-001 (A12,B12,C12,D12). Field observation of indicated voltage constitutes the point in time when availability of indications to plant operators that an emergency action level has been, or may be, exceeded.NEI 99-01 Basis: 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, 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 emerqency 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.Susquehanna Basis Reference(s):

1. FSAR Section 8.2 Offsite Power System 2. FSAR Section 8.3 Onsite Power System 3. Technical Specifications 3.8.1 AC Sources -Operating 4. Technical Specifications 3.8.7 Distribution System -Operating 5. ON-104 (204)-001 Units 1(2) Response to Loss of All Offsite Power 6. EO-100 (200)-030 UNIT 1(2) Response to Station Blackout 7. FSAR Section 8.3.2 DC Power Systems 8. Susquehanna Drawing No. E107159, Sheet 1, "Single Line Meter& Relay Diagram 125 VDC, 250 VDC & 120 VAC Systems" 9. Technical Specifications 3.8.5 DC Sources -Shutdown 10. ON-102(202)-610, -620, -630, -640 Loss of 125V DC 11. AR-1(2)06-001 Main Turbine/Generator, Computer HVAC, Instrument AC, 24V DC, 125V DC, 250V DC Panel 2C651 12. NEI 99-01 SG8 Page 147 of 195 Attachment 1 EP-RM-004 Revision [X]Page 174 of 295 Category:

S -System Malfunction Subcategory: 1 -Loss of Essential AC Power Initiating Condition: Loss of all offsite power and all onsite AC power to essential buses for 15 minutes or longer EAL: SS1.1 Site Area Emergency Loss of ALL offsite and ALL onsite AC power capability to ALL 4.16 kV ESS buses on EITHER unit for > 15 min. (Note 1)Note 1: The ED/RM should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Mode Applicability: 1 -Power Operations, 2 -Startup, 3 -Hot Shutdown Definition(s): None Susquehanna Basis: The Class IE 4.16 kV system supplies all the Engineered Safety Feature (ESF) loads and other loads that are needed for a safe and orderly plant shutdown, and for keeping the plant in a safe shutdown condition. See Figure S-1 (ref. 1, 2) The eight Class 1E 4.16 kV ESS Buses 1(2)A through 1(2)D receive power from either the four ESS 13.8/4.16 kV transformers or the diesel generators (A, B, C, D and additional diesel generator E). Buses 1A-1D supply Unit 1 and common loads and Buses 2A-2D supply Unit 2 loads. This configuration prevents a loss of all ESS Buses for one unit in the event one of the ESS Transformers is lost.During normal plant operation, ESS Transformer 101 supplies preferred power to ESS Bus 1A and 2A and is an alternate power supply to ESS bus 1 D and 2D. ESS Transformer 111 supplies preferred power to ESS Bus 1C and 2C, and is an alternate power supply to ESS bus 1B and 2B. ESS Transformer 201 supplies preferred power to ESS Bus 1 D and 2D, and is an alternate power supply to ESS bus 1A and 2A. ESS Transformer 211 supplies preferred power to ESS Bus 1 B and 2B, and is an alternate power supply to ESS bus 1C and 2C.On a loss of a preferred power source, the bus rapidly transfers to the alternate power source to maintain component power. If both the preferred and alternate power sources are lost, the associated standby diesel generator connects to the ESS bus. (ref. 2-6)The 15-minute interval was selected as a threshold to exclude transient or momentary power losses. The interval begins when both offsite and onsite AC power are lost.NEI 99-01 Basis: This IC addresses a total loss of AC power that compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink.Page 148 of 195 Attachment 1 EP-RM-004 Revision [X]Page 175 of 295 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 AG4RG1, FG1 or SG1.Susquehanna Basis Reference(s):

1. FSAR Section 8.2 Offsite Power System 2. FSAR Section 8.3 Onsite Power System 3. Technical Specifications 3.8.1 AC Sources -Operating 4. Technical Specifications 3.8.7 Distribution System -Operating 5. ON-104 (204)-001 Units 1(2) Response to Loss of All Offsite Power 6. EO-100 (200)-030 UNIT 1(2) Response to Station Blackout 7. NEI 99-01 SS1 Page 149 of 195 Attachment 1 EP-RM-004 Revision [X]Page 176 of 295 Figure S-1 ESS 13.8/4.16 kV Transformers and Distribution (ref. 1)Page 150 of 195 Attachment 1 EP-RM-004 Revision [X]Page 177 of 295 Category:

S -System Malfunction Subcategory: 1 -Loss of Essential AC Power Initiating Condition: Loss of all but one AC power source to essential buses for 15 minutes or longer EAL: SAI.1 Alert AC power capability to ALL 4.16 kV ESS buses on EITHER unit 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 ED/RM should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Mode Applicability: I -Power Operations, 2 -Startup, 3 -Hot Shutdown.Definition(s): SAFETY SYSTEM- A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2): Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the reactor coolant pressure boundary;(2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures. Basis: Susquehanna Basis: The Class 1E 4.16 kV system supplies all the Engineered Safety Feature (ESF) loads and other loads that are needed for a safe and orderly plant shutdown, and for keeping the plant in a safe shutdown condition. See Figure S-1 (ref. 1, 2) The eight Class 1E 4.16 kV ESS Buses 1(2)A through 1(2)D receive power from either the four ESS 13.8/4.16 kV transformers or the diesel generators (A, B, C, D and additional diesel generator E). Buses 1A-1D supply Unit 1 and common loads and Buses 2A-2D supply Unit 2 loads. This configuration prevents a loss of all ESS Buses for one unit in the event one of the ESS Transformers is lost.During normal plant operation, ESS Transformer 101 supplies preferred power to ESS Bus 1A and 2A and is an alternate power supply to ESS bus 1 D and 2D. ESS Transformer 111 supplies preferred power to ESS Bus 1 C and 2C, and is an alternate power supply to ESS bus Page 151 of 195 Attachment 1 EP-RM-004 Revision [X]Page 178 of 295 1 B and 2B. ESS Transformer 201 supplies preferred power to ESS Bus 1 D and 2D, and is an alternate power supply to ESS bus 1A and 2A. ESS Transformer 211 supplies preferred power to ESS Bus 1 B and 2B, and is an alternate power supply to ESS bus 1C and 2C.On a loss of a preferred power source, the bus rapidly transfers to the alternate power source to maintain component power. If both the preferred and alternate power sources are lost, the associated standby diesel generator connects to the ESS bus. (ref. 2-6)The 15-minute interval was selected as a threshold to exclude transient or momentary power losses. If the capability of a second source of emergency bus power is not restored within 15 minutes, an Alert is declared under this EAL.This hot condition EAL is equivalent to the cold condition EAL CU2.1.NEI 99-01 Basis: This IC describes a significant degradation of offsite and onsite AC power sources such that any additional single failure would result in a loss of all AC power to SAFETY SYSTEMS. In this condition, the sole AC power source may be powering one, or more than one, train of safety-related equipment. This IC provides an escalation path from IC SUI.An "AC power source" is a source recognized in AOPs and EOPs, and capable of supplying required power to an emergency bus. Some examples of this condition are presented below.* A loss of all offsite power with a concurrent failure of all but one emergency power source (e.g., an onsite diesel generator).

A loss of all offsite power and loss of all emergency power sources (e.g., onsite diesel generators) with a single train of emergency buses being back-fed from the unit main generator.
A loss of emergency power sources (e.g., onsite diesel generators) with a single train of emergency buses being back-fed from an offsite power source.Fifteen minutes was selected as a threshold to exclude transient or momentary losses of power.Escalation of the emergency classification level would be via IC SS1.Susquehanna Basis Reference(s):

1. FSAR Section 8.2 Offsite Power System 2. FSAR Section 8.3 Onsite Power System 3. Technical Specifications 3.8.1 AC Sources -Operating 4. Technical Specifications 3.8.7 Distribution System -Operating 5. ON-104 (204)-001 Units 1(2) Response to Loss of All Offsite Power 6. EO-100 (200)-030 UNIT 1(2) Response to Station Blackout 7. NEI 99-01 SA1 Page 152 of 195 Attachment 1 EP-RM-004 Revision [X]Page 179 of 295 Figure S-1 ESS 13.8/4.16 kV Transformers and Distribution (ref. 1)Page 153 of 195 Attachment 1 EP-RM-004 Revision [X]Page 180 of 295 Category:

S -System Malfunction Subcategory: 1 -Loss of Essential AC Power Initiating Condition: Loss of all offsite AC power capability to essential buses for 15 minutes or longer EAL: SUI.1 Unusual Event Loss of ALL offsite AC power capability to ALL 4.16 kV ESS buses on EITHER unit for-> 15 min. (Note 1)Note 1: The ED/RM should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Mode Applicability: 1 -Power Operations, 2 -Startup, 3 -Hot Shutdown Definition(s): None Basis: Susquehanna Basis: The Class 1E 4.16 kV system supplies all the Engineered Safety Feature (ESF) loads and other loads that are needed for a safe and orderly plant shutdown, and for keeping the plant in a safe shutdown condition. See Figure S-1 (ref. 1, 2) The eight Class 1E 4.16 kV ESS Buses 1(2)A through 1(2)D receive power from either the four ESS 13.8/4.16 kV transformers or the diesel generators (A, B, C, D and additional diesel generator E). Buses 1A-1D supply Unit 1 and common loads and Buses 2A-2D supply Unit 2 loads. This configuration prevents a loss of all ESS Buses for one unit in the event one of the ESS Transformers is lost.During normal plant operation, ESS Transformer 101 supplies preferred power to ESS Bus 1A and 2A and is an alternate power supply to ESS bus 1D and 2D. ESS Transformer 111 supplies preferred power to ESS Bus 1C and 2C, and is an alternate power supply to ESS bus 1B and 2B. ESS Transformer 201 supplies preferred power to ESS Bus 1D and 2D, and is an alternate power supply to ESS bus 1A and 2A. ESS Transformer 211 supplies preferred power to ESS Bus 1 B and 2B, and is an alternate power supply to ESS bus 1C and 2C.On a loss of a preferred power source, the bus rapidly transfers to the alternate power source to maintain component power. If both the preferred and alternate power sources are lost, the associated standby diesel generator connects to the ESS bus. (ref. 2-6)The 15-minute interval was selected as a threshold to exclude transient or momentary power losses.NEI 99-01 Basis: 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.Page 154 of 195 Attachment 1 EP-RM-004 Revision [X]Page 181 of 295 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.Susquehanna Basis Reference(s):

1. FSAR Section 8.2 Offsite Power System 2. FSAR Section 8.3 Onsite Power System 3. Technical Specifications 3.8.1 AC Sources -Operating 4. Technical Specifications 3.8.7 Distribution System -Operating 5. ON-104 (204)-001 Units 1(2) Response to Loss of All Offsite Power 6. EO-100 (200)-030 UNIT 1(2) Response to Station Blackout 7. NEI 99-01 SU1 Page 155 of 195 Attachment 1 EP-RM-004 Revision [X]Page 182 of 295 Figure S-1 ESS 13.814.16 kV Transformers and Distribution (ref. 1)Page 156 of 195 Attachment 1 EP-RM-004 Revision [X]Page 183 of 295 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: S82.1 Site Area Emergency Indicated voltage is < 105 VDC on ALL of the following vital 125 VDC main distribution buses on the affected unit for -15 min. (Note 1):* 1D612 (2D612)* 1D622 (2D622)* 1D632 (2D632)" 1D642 (2D642)Note 1: The ED/RM should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Mode Applicability: 1 -Power Operations, 2 -Startup, 3 -Hot Shutdown Definition(s): None Susquehanna Basis: The Class 1E Battery Banks are 1(2)D610 (Channel A), 1(2)D620 (Channel B), 1(2)D630 (Channel C), and 1(2)D640 (Channel D). Each bank consists of 60 cells connected in series.Each cell produces a nominal voltage of 2.06 VDC resulting in a total battery bank terminal voltage of 123.6 VDC. All battery banks are designed to supply power to its load center for four hours in the event of a loss of power from its battery charger (ref. 1-3).105 VDC is the minimum design voltage limit (ref. 4).Indicated voltage for the vital 125 VDC main distribution buses is local only. Local voltage indication is available for each bus based on dispatching a field operator in accordance with Control Room alarm response procedure AR-1(2)06-001 (A12,B12,C12,D12). Field observation of indicated voltage constitutes the point in time when availability of indications to plant operators that an emergency action level has been, or may be, exceeded.NEI 99-01 Basis: 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 AG4RG1, FG1 or SG8,SG1.Page 157 of 195 Attachment 1 EP-RM-004 Revision [XI Page 184 of 295 Susquehanna Basis Reference(s):

1. FSAR Section 8.3.2 DC Power Systems 2. Susquehanna Drawing No. E107159, Sheet 1, "Single Line Meter & Relay Diagram 125 VDC, 250 VDC & 120 VAC Systems" 3. Technical Specifications 3.8.5 DC Sources -Shutdown 4. ON-102(202)-610, -620, -630, -640 Loss of 125V DC 5. AR-1(2)06-001 Main Turbine/Generator, Computer HVAC, Instrument AC, 24V DC, 125V DC, 250V DC Panel 2C651 6. NEI 99-01 SS8 Page 158 of 195 Attachment I EP-RM-004 Revision [X]Page 185 of 295 Category: Subcategory:

Initiating Condition: S -System Malfunction 3 -Loss of Control Room Indications 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-1 parameters from within the Control Room for -- 15 min. (Note 1)AND Any significant transient is in progress, Table S-2 Note 1: The ED/RM should declare the event promptly upon determining that time limit has been exceeded, or Will likely be exceeded.Table S-1 Safety System Parameters" Reactor power* RPV water level* RPV pressure* Primary Containment pressure* Suppression Pool water level* Suppression Pool temperature Table S-2 Significant Transients

Reactor scram* Runback > 25% reactor power* RRC pump trip while > 25% reactor power* ECCS injection* Thermal power oscillations

> 10%Mode Applicability: 1 -Power Operations, 2 -Startup, 3 -Hot Shutdown Definition(s): UNPLANNED -A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.Page 159 of 195 Attachment 1 EP-RM-004 Revision [X]Page 186 of 295 Susquehanna 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 Process Computer (PPC) and SPDS are redundant compensatory indication 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 scrams, runbacks involving greater than 25% thermal power change, RRC pump trip while > 25% reactor power, ECCS injections, or thermal power oscillations of 10% or greater.NEI 99-01 Basis: This IC addresses the difficulty associated with monitoring rapidly changing plant conditions during a transient without the ability to obtain SAFETY SYSTEM parameters from within the Control Room. During this condition, the margin to a potential fission product barrier challenge is reduced. It thus represents a potential substantial degradation in the level of safety of the plant.As used in this EAL, an "inability to monitor" means that values for one or more of the listed parameters cannot be determined from within the Control Room. This situation would require a loss of all of the Control Room sources for the given parameter(s). For example, the reactor power level cannot be determined from any analog, digital 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-1 022) 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 / RPV level -BW4R-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 / RPV water level [B9WR-annot 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 AS4RS1 Susquehanna Basis Reference(s):

1. FSAR Section 18.1.17 Plant Safety Parameter Display System 2. OP-131(231)-002 Plant Computer Systems Page 160 of 195 Attachment 1 EP-RM-004 Revision [X]Page 187 of 295 3. EO-000-102 RPV Control 4. EO-000-103 Primary Containment Control 5. NEI 99-01 SA2 Page 161 of 195 Attachment 1 EP-RM-004 Revision [X]Page 188 of 295 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-1 parameters from within the Control Room for > 15 min. (Note 1)Note 1: The ED/RM should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Table S-1 Safety System Parameters

Reactor power" RPV water level* RPV pressure* Primary Containment pressure* Suppression Pool water level* Suppression Pool temperature Mode Applicability:

1 -Power Operations, 2 -Startup, 3 -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.Susquehanna 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 Process Computer (PPC) and SPDS are redundant compensatory indication which may be utilized in lieu of normal Control Room indicators (ref. 1, 2).NEI 99-01 Basis: This IC addresses the difficulty associated with monitoring normal plant conditions without the ability to obtain SAFETY SYSTEM parameters from within the Control Room. This condition is a precursor to a more significant event and represents a potential degradation in the level of safety of the plant.As used in this EAL, an "inability to monitor" means that values for one or more of the listed parameters cannot be determined from within the Control Room. This situation would require a loss of all of the Control Room sources for the given parameter(s). For example, the reactor Page 162 of 195 Attachment 1 EP-RM-004 Revision [X]Page 189 of 295 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 [PW.] ! RPV level [B9WR]-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 [PWR! -RPV water level [1/WRI 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 SA-2SA3.Susquehanna Basis Reference(s):

1. FSAR Section 18.1.17 Plant Safety Parameter Display System 2. OP-131(231)-002 Plant Computer Systems 3. EO-000-102 RPV Control 4. EO-000-103 Primary Containment Control 5. NEI 99-01 SU2 Page 163 of 195 Attachment 1 EP-RM-004 Revision [X]Page 190 of 295 Category:

S -System Malfunction Subcategory: 4 -RCS Activity Initiating Condition: Reactor coolant activity greater than Technical Specification allowable limits EAL: SU4.1 Unusual Event Offgas pretreatment monitor high-high radiation alarm Mode Applicability: 1 -Power Operations, 2 -Startup, 3 -Hot Shutdown Definition(s): None Susquehanna Basis: The Offgas Pretreatment RMS monitors radioactivity in the Offgas system downstream of the Motive Steam Jet Condenser. The monitor detects the radiation level that is attributable to the fission gases produced in the reactor and transported with steam through the turbine to the condenser (ref. 1). Log rad monitors and trip auxiliary units are located on Panel 1C604 in the Upper Relay Room. Instrument Channel 'A' is RITS-D12-1K601A and Instrument Channel 'B' is RITS-D12-1 K601B. Both channels output to Yokagowa Recorder RR-D12-1 R601 on Main Control Room Panel 1C600 (ref. 2, 3).OFFGAS HI-HI RADIATION (AR-1 06-F03) is located on Panel 1 C651. The setpoint is variable based on surveillance procedure (ref. 4).NEI 99-01 Basis: This IC addresses a reactor coolant activity value that exceeds an allowable limit specified in Technical Specifications. This condition is a precursor to a more significant event and represents a potential degradation of the level of safety of the plant.Escalation of the emergency classification level would be via ICs FA1 or the Recognition Category A-R ICs.Susquehanna Basis Reference(s):

1. FSAR Section 11.5 Process and Effluent Radiological Monitoring and Sampling Systems 2. Technical Specification 3.7.5 Main Condenser Offgas 3. AR-106(206)-001 F03 Offgas Hi Hi Radiation 4. SC-143(243)-101 Unit 1 (Unit 2) Main Condenser Air Ejector Monthly Noble Gas Activity 5. OP-179(279)-002 Process Radiation Monitoring System 6. NEI 99-01 SU3 Page 164 of 195 Attachment 1 EP-RM-004 Revision [X]Page 191 of 295 Category:

S -System Malfunction Subcategory: 4 -RCS Activity Initiating Condition: Reactor coolant activity greater than Technical Specification allowable limits EAL: SU4.2 Unusual Event Coolant activity > 0.2 pCi/gm dose equivalent 1-131 for > 48 hours OR Coolant activity > 4.0 pCi/gm dose equivalent 1-131 at any time Mode Applicability: 1 -Power Operations, 2 -Startup, 3 -Hot Shutdown Definition(s): None Susquehanna Basis: The specific iodine activity is limited to < 0.2 pCi/gm dose equivalent 1-131 (Condition.A) with a completion time of 48 hours. This limit ensures the source term assumed in the safety analysis for the Main Steam Line Break (MSLB) is not exceeded, so any release of radioactivity to the environment during an MSLB is less than a small fraction of the regulatory limits (ref. 1).The upper limit of 4.0 pCi/gm dose equivalent 1-131 (Condition B) ensures that the TEDE dose from an MSLB will not exceed the dose guidelines of 10 CFR 50.67 or Control Room operator dose limits specified in GDC 19 of 10 CFR 50, Appendix A (ref. 1).NEI 99-01 Basis: This IC addresses a reactor coolant activity value that exceeds an allowable limit specified in Technical Specifications. This condition is a precursor to a more significant event and represents a potential degradation of the level of safety of the plant.Escalation of the emergency classification level would be via ICs FA1 or the Recognition Category A-RICs.Susquehanna Basis Reference(s):

1. Technical Specifications section 3.4.7 RCS Specific Activity 2. NEI 99-01 SU3 Page 165 of 195 Attachment 1 EP-RM-004 Revision [X]Page 192 of 295 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 Primary Containment > 25 gpm for > 15 min.(Note 1)Note 1: The ED/RM should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Mode Applicability: 1 -Power Operations, 2 -Startup, 3 -Hot Shutdown Definition(s): None Susquehanna Basis: Leakage is monitored by utilizing the following techniques (ref. 1):* Monitoring changes in water level in the drywell floor drain sumps and drywell equipment drain tank" Sensing excess flow in piping systems* Sensing pressure and temperature changes in the primary containment

Monitoring for high flow and temperature through selected drains," Sampling airborne particulate and gaseous radioactivity.

Identified leakage is leakage into the drywell, such as that from pump seals or valve packing, that is captured and conducted to the drywell equipment drain tank; or leakage into the drywell atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be pressure boundary Unidentified leakage is all leakage into the drywell that is not identified leakage (ref. 2).Pressure boundary leakage is leakage through a nonisolable fault in a Reactor Coolant System (RCS) component body, pipe wall, or vessel wall (ref. 2).Two drywell floor drain sumps are located in the primary containment for collection of leakage from vent coolers, control rod drive flange leakage, chilled water drains, cooling water drains, Page 166 of 195 Attachment 1 EP-RM-004 Revision [X]Page 193 of 295 and overflow from the drywell equipment drain tank. The drywell floor drain sumps are located at the drywell diaphragm slab low point. Unidentified leakages will, by gravity, flow down the slab surface into the floor drain sumps. Water flow rate greater than 0.5 gpm can be detected by monitoring changes of level over a time period. The sump depth of 0-5 in. is displayed on a 0-100 percent recorder chart, which relates to the sump nominal capacity of 0-150 gal.The drywell equipment drain tank collects identified leakage within the primary containment from reactor head seal leak off, bulkhead drain, refueling bellows drain, RPV head vent, recirculation pump seals, reactor recirculation pump cooler drains, and RPV bottom drain (Unit 1 only). The measured tank depth of 36 in. is displayed on a 0-100 percent recorder chart. This relates directly to the measured tank capacity of 842 gal.RCS leakage outside of the Primary Containment that is not considered identified or unidentified leakage per Technical Specifications includes leakage via interfacing systems such as RCS to the Reactor Building Closed Cooling Water (RBCCW system), or systems that directly see RCS pressure outside primary containment such as Reactor Water Cleanup (RWCU), reactor water sampling system and Residual Heat Removal (RHR) system (when in the shutdown cooling mode) (ref. 1, 3).Indicated changes in drywell sump water level are used to calculate unidentified drywell leakage. Indicated changes in drywell equipment drain tank level are used to calculate identified drywell leakage. SO-100-006 and SO-200-006 are the procedures that specify how to complete these calculations. Drywell leakage calculations in SO-100(200)-006 take a finite period of time to complete.Leakage rates cannot be determined quickly by merely observing an indicator. For this reason, the 15 minutes clock starts after it is determined that leakage rates exceed the entry value.Upon determination that leakage has increased substantially, effort should be made to quantify this leakage in a timely manner.ON-1 (2)00-005, "Excessive Drywell Leakage Identification", contains methods of quickly estimating drywell leakage. These methods can be used in lieu of completing the calculations contained in SO-1(2)00-006. Means to directly quantify RCS leakage outside containment may not be available. For this reason, judgment must be used for assessment of the 25 gpm leak rate criterion. For example, a short steam plume that does not appreciably change room temperature or room radiation levels ca n be judged to be less than 25 gpm. A lea k that causes room temperature to rise rapidly above maximum safe temperatures could be judged to be greater than 25 gpm in the absence of measurable lea k rates, and thus judgment is an acceptable method to evaluate this criterion. Escalation of this EAL to the Alert level is via Category F, Fission Product Barrier Degradation, EAL FA1.1. The note has been added to remind the EAL-user to review Table F-1 for possible escalation to higher emergency classifications. NEI 99-01 Basis: 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.Page 167 of 195 Attachment 1 EP-RM-004 Revision [X]Page 194 of 295..... .1 andEAL..2The 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). E-AL-#4The third condition addresses an RCS mass loss caused by an UNISOLABLE leak through an interfacing system. These EALs-conditions thus apply to leakage into the ,,,ai ,e"t".. Prima,, Containment, a secondary side system (e.g., steam tube leakage in a PWR) or a location outside of entaikimentPrimary Containment. The leak rate values for each E-AL-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). E-AL-#1-The first condition uses a lower value that reflects the greater significance of unidentified or pressure boundary leakage.The release of m.ass f..ro the ROS due to the as designedexpected operatien of a relief valve does not warrant an emergency classification. For PWRs, an emergencY classification would be required if a mass loss is c0a-used by a relief valve that is noGt functioning as designod~expected (e.g., a relief valve sticks open and the line flow c-annot be isolated) F-of BWRs, a A stuck-open Safety Relief Valve (SRV) or SRV leakage is not considered either identified or unidentified leakage by Technical Specifications and, therefore, is not applicable to this EAL.The 15-minute threshold duration allows sufficient time for prompt operator actions to isolate the leakage, if possible.Escalation of the emergency classification level would be via ICs of Recognition Category A-R or F.Susquehanna Basis Reference(s):

1. FSAR Section 5.2.5 Detection of Leakage Through Reactor Coolant Pressure Boundary 2. Technical Specifications Definitions Section 1.1 3. ON-100(200)-005 Excess Drywell Leakage Identification

4. SO-100(200)-006

5. NEI 99-01 SU4 Page 168 of 195 Attachment 1 EP-RM-004 Revision [X]Page 195 of 295 Category: Subcategory:

Initiating Condition: S -System Malfunction 2 -RPS Failure Inability to shut down the reactor causing a challenge to RPV water level or RCS heat removal EAL: SS6.1 Site Area Emergency An automatic or manual scram fails to shut down the reactor AND ALL actions to shut down the reactor are not successful as indicated by reactor power > 5%AND EITHER:* RPV level CANNOT BE RESTORED AND MAINTAINED > -179 in. or CANNOT be determined OR" Suppression pool water temperature AND RPV pressure CANNOT BE MAINTAINED below the Heat Capacity Temperature Limit (Figure -HCTL)Figure -HCTL Heat Capacity Temperature Limit C 210 200 190 180 id 170 R 160 150 140" 130 Q) 120 12 13 14 15 16 17 18 19 20 21 22 23 24 25 26 27 28 29 30 31 32 33 34 35 36 37 38 SUPPRESSION POOL LEVEL (F")Page 169 of 195 Attachment 1 EP-RM-004 Revision [X]Page 196 of 295 Mode Applicability: 1 -Power Operations, 2 -Startup Definition(s): None Susquehanna Basis: This EAL addresses the following: " Any automatic reactor scram signal followed by a manual scram actions that fail to shut down the reactor to an extent the reactor is producing energy in excess of the heat load for which the safety systems were designed (EAL SA6.1), and" Indications that either core cooling is extremely challenged or heat removal is extremely challenged. Reactor shutdown achieved by use of EO-000-1 13, Control Rod Insertion, is also credited as a successful scram provided reactor power can be reduced below the APRM downscale trip setpoint before indications of an extreme challenge to either core cooling or heat removal exist.(ref. 1)The APRM downscale trip setpoint (5%) is a minimum reading on the power range scale that indicates 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 the APRM downscale trip setpoint, plant response will be similar to that observed during a normal shutdown. Nuclear instrumentation (APRM)indications or other reactor parameters (steam flow, RPV pressure, torus temperature trend)can be used to determine if reactor power is greater than 2% power (ref. 2, 3).The combination of failure of both front line and backup protection systems to function in response to a plant transient, along with the continued production of heat, poses a direct threat to the Fuel Clad and RCS barriers.Indication that core cooling is extremely challenged is manifested by inability to restore and maintain RPV water level above the Minimum Steam Cooling RPV Water Level (MSCRWL) (ref.3). The MSCRWL is the lowest RPV level at which the covered portion of the reactor core will generate sufficient steam to prevent any clad temperature in the uncovered part of the core from exceeding 1500 0 F. This water level is utilized in the EOPs to preclude fuel damage when RPV level is below the top of active fuel. RPV level below the MSCRWL for an extended period of time without satisfactory core spray cooling could be a precursor of a core melt sequence.When RPV level cannot be determined, EOPs require entry to EO-000-1 14, RPV Flooding. RPV water level indication provides the primary means of knowing if adequate core cooling is being maintained. When all means of determining RPV water level are unavailable, the fuel clad barrier is threatened and reliance on alternate means of assuring adequate core cooling must be attempted. The instructions in EO-000-1 14 specify these means, which include emergency depressurization of the RPV and injection into the RPV at a rate needed to flood to the elevation of the main steam lines or hold RPV pressure above the Minimum Steam Cooling Pressure (ref.4).Page 170 of 195 Attachment 1 EP-RM-004 Revision [X]Page 197 of 295 The Heat Capacity Temperature Limit (HCTL) is the highest suppression pool water temperature from which Emergency RPV Depressurization will not raise suppression chamber pressure above the Primary Containment Pressure Limit, while the rate of energy transfer from the RPV to the containment is greater than the capacity of the containment vent.The HCTL is a function of RPV pressure and suppression pool water level. It is utilized to preclude failure of the containment and equipment in the containment necessary for the safe shutdown of the plant. This threshold is met when the final step of section SP/T in EO-000-1 03, Primary Containment Control, is reached (ref. 5). This condition addresses loss of functions required for hot shutdown with the reactor at pressure and temperature. NEI 99-01 Basis: This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor (tF,- scram [B4')--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 shut down the reactor.A-The adequacy of reactor shutdown (<5%) is determined in accordance with applicable Emergency Operating Procedure criteria.Escalation of the emergency classification level would be via IC AG-1-RG1 or FGI.Page 171 of 195 Attachment 1 EP-RM-004 Revision [X]Page 198 of 295 Susquehanna Basis Reference(s):

1. EO-000-1 13 Control Rod Insertion 2. Technical Specifications Table 3.3.1.1-1 3. EO-000-102 RPV Control 4. EO-000-1 14 RPV Flooding 5. EO-000-103 Primary Containment Control 6. NEI 99-01 SS5 Page 172 of 195 Attachment 1 EP-RM-004 Revision [X]Page 199 of 295 Category:

S -System Malfunction Subcategory: 2 -RPS Failure Initiating Condition: Automatic or manual scram fails to shut down the reactor and subsequent manual actions taken at the reactor control consoles are not successful in shutting down the reactor EAL: SA6.1 Alert An automatic or manual scram fails to shut down the reactor AND Manual scram actions taken at the reactor control console (Manual PBs, Mode Switch, ARI) are not successful in shutting down the reactor as indicated by reactor power > 5% (Note 8)Note 8: A manual scram action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron injection strategies. Mode Applicability: 1 -Power Operations, 2 -Startup Definition(s): None Susquehanna Basis: This EAL addresses any automatic or manual reactor scram signal that fails to shut down the reactor followed by a subsequent manual scram that fails to shut down the reactor to an extent the reactor is producing energy in excess of the heat load for which the safety systems were designed.For the purposes of emergency classification, successful manual scram actions are those which can be quickly performed from the reactor control console (i.e., manual scram pushbuttons, mode switch, orARI initiation in accordance with EO-000-102 or EO-000-113). Reactor shutdown achieved by use of other control rod insertion methods (e.g. individual control rod insertion) directed by EO-000-113 does not constitute a successful manual scram (ref. 2, 3).For the purposes of this EAL, a successful automatic initiation of ARI that reduces reactor power below 5% is not considered a successful automatic scram. If automatic actuation of ARI has occurred and caused reactor shutdown, the automatic RPS scram must have failed. ARI is a backup means of inserting control rods in the unlikely event that an automatic RPS scram signal exists but the reactor continues to generate significant power. However, a successful automatic initiation of ARI is an acceptable means of establishing reactor shutdown conditions relative to the EAL threshold in the absence of any required subsequent manual scram actions.The APRM downscale trip setpoint (5%) is a minimum reading on the power range scale that indicates 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 Page 173 of 195 Attachment 1 EP-RM-004 Revision [X]Page 200 of 295 prevent subsequent core damage. Below the APRM downscale trip setpoint, plant response will be similar to that observed during a normal shutdown. Nuclear instrumentation (APRM)indications or other reactor parameters (steam flow, RPV pressure, torus temperature trend)can be used to determine if reactor power is greater than 5% power (ref. 1, 3).Escalation of this event to a General Emergency would be under EAL SG2.1 or Emergency Director/Recovery Manager judgment.NEI 99-01 Basis: This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor (tp-f rBWR]}-that results in a reactor shutdown, and subsequent operator manual actions taken at the reactor control consoles to shutdown the reactor are also unsuccessful. This condition represents an actual or potential substantial degradation of the level of safety of the plant. An emergency declaration is required even if the reactor is subsequently shutdown by an action taken away from the reactor control consoles since this event entails a significant failure of the RPS.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-(tri4p [P-WR] scram-f[-])). This action does not include manually driving in control rods or implementation of boron injection strategies. If this action(s) is unsuccessful, operators would immediately pursue additional manual actions at locations away from the reactor control consoles (e.g., locally opening breakers). Actions taken at back-panels or other locations within the Control Room, or any location outside the Control Room, are not considered to be "at the reactor control consoles". Taking the Reactor Mode Switch to SHUTDOWN is considered to be a manual scram action.[8w I The plant response to the failure of an automatic or manual reactor (trip [P-149. / scramrriL4rj. 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 ..r, cooling [PWR]! RPV water level-[3WR] or RCS heat removal safety functions, the emergency classification level will escalate to a Site Area Emergency via ICDepending upon plant responses and symptoms, escalation is also possible via IC FSI.Absent the plant conditions needed to meet either IC SS645 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-The adequacy of reactor shutdown (< 5%) is determined in accordance with applicable Emergency Operating Procedure criteria.Susquehanna Basis Reference(s):

1. Technical Specifications Table 3.3.1.1-1 2. EO-000-1 13 Control Rod Insertion 3. EO-000-102 RPV Control Page 174 of 195 Attachment 1 EP-RM-004 Revision [XI Page 201 of 295 4. NEI 99-01 SA5 Page 175 of 195 Attachment 1 EP-RM-004 Revision [X]Page 202 of 295 Category:

S -System Malfunction Subcategory: 6 -RPS Failure Initiating Condition: Automatic or manual scram fails to shut down the reactor EAL: SU6.1 Unusual Event An automatic scram did not shut down the reactor after any RPS setpoint is exceeded AND A subsequent automatic scram or manual scram action taken at the reactor control console (Manual PBs, Mode Switch, ARI) is successful in shutting down the reactor as indicated by reactor power < 5% (APRM downscale) (Note 8)Note 8: A manual scram action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron injection strategies. Mode Applicability: 1 -Power Operations, 2 -Startup Definition(s): None Susquehanna Basis: The first condition of this EAL identifies the need to cease critical reactor operations by actuation of the automatic Reactor Protection System (RPS) scram function. A reactor scram is automatically initiated by the Reactor Protection System (RPS) when certain continuously monitored parameters exceed predetermined setpoints (ref. 1).Following a successful reactor scram, 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 period. 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-scram response from an automatic reactor scram 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 scram has therefore occurred when there is sufficient rod insertion from the trip of RPS to bring the reactor power below the APRM downscale setpoint of 5%.For the purposes of emergency classification, successful manual scram actions are those which can be quickly performed from the reactor control console (i'e., manual scram pushbuttons, mode switch, or ARI initiation in accordance with EO-000-102 or EO-000-113). Reactor shutdown achieved by use of other control rod insertion methods (e.g. individual control rod insertion) directed by EO-000-1 13 does not constitute a successful manual scram (ref. 2, 3).Following any automatic RPS scram signal, operating procedures (e.g., EO-000-102) prescribe insertion of redundant manual scram signals to back up the automatic RPS scram function and ensure reactor shutdown is achieved. Even if the first subsequent manual scram signal inserts Page 176 of 195 Attachment I EP-RM-004 Revision [X]Page 203 of 295 all control rods to the full-in position immediately after the initial failure of the automatic scram, the lowest level of classification that must be declared is an Unusual Event. (ref. 3)Taking the mode switch to shutdown is a manual scram action. When the Mode Switch is taken out of the Run position, however, the nuclear instrumentation scram setpoint is lowered. If reactor power remains above the lowered setpoint, an automatic scram is initiated. For the purposes of this EAL, a successful automatic initiation of ARI that reduces reactor power below 5% is not considered a successful automatic scram. If automatic initiation of ARI has occurred and caused reactor shutdown, the automatic RPS scram must have failed. ARI is a backup means of inserting control rods in the unlikely event that an automatic RPS scram signal exists but the reactor continues to generate significant power. However, a successful automatic or manual initiation of ARI is an acceptable means of establishing reactor shutdown conditions relative to the EAL threshold in the absence of any required subsequent manual scram actions.In the event that the operator identifies a reactor scram is imminent and initiates a successful manual reactor scram before the automatic scram setpoint is reached, no declaration is required. The successful manual scram of the reactor before it reaches its automatic scram setpoint or reactor scram signals caused by instrumentation channel failures do not lead to a potential fission product barrier loss. If manual reactor scram 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 scram following receipt of an automatic scram signal and there are no clear indications that the automatic scram failed (such as a time delay following indications that a scram setpoint was exceeded), it may be difficult to determine if the reactor was shut down because of automatic scram or manual actions. If a subsequent review of the scram actuation indications reveals that the automatic scram did not cause the reactor to be shut down, then consideration should be given to evaluating the fuel for potential damage, and the reporting requirements of 10CFR 50.72 should be considered for the transient event.NEI 99-01 Basis: This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor (trip ! scram [B914q) that results in a reactor shutdown, and either a subsequent operator manual action taken at the reactor control consoles or an automatic (trip [P-Wq ! scram [B-VVI)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 (ti4p [PVR",, scram-fBWq), operators will promptly initiate manual actions at the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor (tFip.,P,14i scram-{B-gWR). 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 scram [944q) is unsuccessful, operators will promptly take manual action at another location(s) on the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor ,,,scram--B44iR). using a different switch).Depending upon several factors, the initial or subsequent effort to manually (tFip- ,[LW43- scram fB44R) the reactor, or a concurrent plant condition, may lead to the generation of an automatic reactor (t*ip-f1-,R] .scram [9144q) signal. If a subsequent manual or automatic (trip Page 177 of 195 Attachment 1 EP-RM-004 Revision [XJ Page 204 of 295 scram [&4q) 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 (tfi[PWR-] scramL{BWRD/). This action does not include manually driving in control rods or implementation of boron injection strategies. Actions taken at back-panels or other locations within the Control Room, or any location outside the Control Room, are not considered to be "at the reactor control consoles". Taking the Reactor Mode Switch to SHUTDOWN is considered to be a manual scram action.The plant response to the failure of an automatic or manual reactor (trip,[PWRI , scram [BWR])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 -A5SA6. Depending upon the plant response, escalation is also possible via IC FA1. Absent the plant conditions needed to meet either IC SA5-SA6_or FA1, an Unusual Event declaration is appropriate for this event.A-The adequacy of reactor shutdown (< 5%) is determined in accordance with applicable Emergency Operating Procedure criteria.Should a reactor ! ) signal be generated as a result of plant work (e.g., RPS setpoint testing), the following classification guidance should be applied." If the signal causes a plant transient that should have included an automatic reactor (t4p ,PR4.4-scram [BWRI) 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 (tFip-fPWR] 1-scramr-{BWR] 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. Susquehanna Basis Reference(s):

1. Technical Specifications Table 3.3.1.1-1 2. EO-000-1 13 Control Rod Insertion 3. EO-000-102 RPV Control 4. NEI 99-01 SU5 Page 178 of 195 Attachment 1 EP-RM-004 Revision [X]Page 205 of 295 Category:

S -System Malfunction Subcategory: 6 -RPS Failure Initiating Condition: Automatic or manual scram fails to shut down the reactor EAL: SU6.2 Unusual Event A manual scram did not shut down the reactor after any manual scram action was initiated AND A subsequent automatic scram or manual scram action taken at the reactor control console (Manual PBs, Mode Switch, ARI) is successful in shutting down the reactor as indicated by reactor power < 5% (APRM downscale) (Note 8)Note 8: A manual scram action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron injection strategies. Mode Applicability: 1 -Power Operations, 2 -Startup Definition(s): None Susquehanna Basis: This EAL addresses a failure of a manually initiated scram in the absence of having exceeded an automatic RPS trip setpoint and a subsequent automatic or manual scram is successful in shutting down the reactor (reactor power < 5%) (ref. 1).Following a successful reactor scram, 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 period. 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-scram response from a manual reactor scram 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 scram has therefore occurred when there is sufficient rod insertion from the trip of RPS to bring the reactor power below the APRM downscale setpoint of 5%.For the purposes of emergency classification, successful manual scram actions are those which can be quickly performed from the reactor control console (i.e., manual scram pushbuttons, mode switch, orARI initiation in accordance with EO-000-102 or EO-000-113). Reactor shutdown achieved by use of other control rod insertion methods (e.g. individual control rod insertion) directed by EO-000-1 13 does not constitute a successful manual scram (ref. 2, 3).Taking the mode switch to shutdown is a manual scram action. When the Mode Switch is taken out of the Run position, however, the nuclear instrumentation scram setpoint is lowered. If reactor power remains above the lowered setpoint, an automatic scram is initiated. Page 179 of 195 Attachment 1 EP-RM-004 Revision [X]Page 206 of 295 Successful automatic or manual initiation of ARI is an acceptable means of establishing reactor shutdown conditions relative to the EAL threshold in the absence of any required subsequent manual scram actions.If both subsequent automatic and subsequent manual reactor scram actions in the Control Room fail to reduce reactor power below the power associated with the safety system design (< 5%) following a failure of an initial manual scram, the event escalates to an Alert under EAL SA6.1 NEI 99-01 Basis: This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor (tF.p- [p-4R !scram [844M) that results in a reactor shutdown, and either a subsequent operator manual action taken at the reactor control consoles or an automatic (tuip [p-A4 ! scram {BW441R]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 (tuip,,['P.,,R scram-f8-WR]), operators will promptly initiate manual actions at the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor (tip-1,,rP]t. scram-[BW-R])). 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 (tFipP,,-,44R-i scram [&WR]) 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 [fP-44q i scram [9lo44)) using a different switch).Depending upon several factors, the initial or subsequent effort to manually (tRip PI4M 1 Iscram[84R]) the reactor, or a concurrent plant condition, may lead to the generation of an automatic reactor (trip [P-{PL44R scram signal. If a subsequent manual or automatic (t~ip-fPI4I-scram f9WRI) 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 #4p,scram--B.WR])). This action does not include manually driving in control rods or implementation of boron injection strategies. Actions taken at back-panels or other locations within the Control Room, or any location outside the Control Room, are not considered to be "at the reactor control consoles". Taking the Reactor Mode Switch to SHUTDOWN is considered to be a manual scram action.The plant response to the failure of an automatic or manual reactor (trip,.PWRI / scram [-WRI)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 SASA6. Depending upon the plant response, escalation is also possible via IC FAI. Absent the plant conditions needed to meet either IC SA5-SA6 or FA1, an Unusual Event declaration is appropriate for this event.Page 180 of 195 Attachment 1 EP-RM-004 Revision [X]Page 207 of 295 A-The adequacy of reactor shutdown (< 5%) is determined in accordance with applicable Emergency Operating Procedure criteria.Should a reactor (tiP, .PWR] , scram, ..R. signal be generated as a result of plant work (e.g., RPS setpoint testing), the following classification guidance should be applied." If the signal causes a plant transient that should have included an automatic reactor (tr4p[PWAR] ! scram [BWR]) 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 [PWR] ! scram-fBWR-) 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. Susquehanna Basis Reference(s):

1. Technical Specifications Table 3.3.1.1-1 2. EO-000-113 Control Rod Insertion 3. EO-000-102 RPV Control 4. NEI 99-01 SU5 Page 181 of 195 Attachment 1 EP-RM-004 Revision [X]Page 208 of 295 Category: Subcategory:

Initiating Condition: EAL: S -System Malfunction 7 -Loss of Communications Loss of all onsite or offsite communications capabilities SU7.1 Unusual Event Loss of ALL Table S-3 onsite communication methods OR Loss of ALL Table S-3 ORO communication methods OR Loss of ALL Table S-3 NRC communication methods Table S-3 Communication Methods System Onsite ORO NRC UHF Radio x Plant PA System x Dedicated Conference Lines x Commercial Telephone Systems x x x Cellular Telephone X x FTS-2001 (ENS) x x Mode Applicability: 1 -Power Operations, 2 -Startup, 3 -Hot Shutdown Definition(s): None Susquehanna Basis: Onsite/offsite communications include one or more of the systems listed in Table S-3 (ref. 1, 2, 3).UHF Radio Onsite portable radio communication systems are described in the Susquehanna SES Physical Security Plan and in the Susquehanna SES Emergency Plan. Four UHF channels, each Page 182 of 195 Attachment 1 EP-RM-004 Revision [X]Page 209 of 295 consisting of two frequencies for duplex operation through one of five in-plant repeaters, provide onsite portable radio communications. Operations is assigned two channels; one channel is assigned to Unit 1 and one to Unit 2. Operators in the plant on rounds and on specific assignments are equipped with handheld two-way radios.Plant PA System The plant PA system is an intra-plant public address providing the following functions: " A 5-channel page-talk handset intercom system for on-site communications between plant locations." Broadcast accountability and fire alarms designed to warn personnel of emergency conditions. The system consists of telephone handsets, amplifiers and loudspeakers located at various selected areas throughout the plant.Dedicated Conference Lines (Centrex Three (3) digjit dialing)The Dedicated Conference Lines are those normally used to communicate with several offsite agencies at one time (e.g., 191 conference line).Commercial Telephone Systems Two independent telecommunications networks exist to provide primary and backup telephone communications between ERFs and offsite agencies.Cellular Telephone Cell phones can be utilized to perform both ORO and NRC communications. FTS 2001 (ENS)This system is for NRC offsite communications but may also be used to perform ORO notifications. This EAL is the hot condition equivalent of the cold condition EAL CU5.1.NEI 99-01 Basis: This IC addresses a significant loss of on-site or offsite communications capabilities. While not a direct challenge to plant or personnel safety, this event warrants prompt notifications to 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.).EAL-#-IThe first EAL condition addresses a total loss of the cofimunications methods used in support of routine plant operations. EAk-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 Notes) the Commonwealth of Pennsylvania, Luzerne and Columbia County Page 183 of 195 Attachment 1 EP-RM-004 Revision [X]Page 210 of 295 EAL--#The third EAL addresses a total loss of the communications methods used to notify the NRC of an emergency declaration. Susquehanna Basis Reference(s):

1. EP-RM-007 Emergency Telephone Instructions and Directory 2. SSES Emergency Plan Section 8 3. FSAR Section 9.5.2 4. NEI 99-01 SU6 Page 184 of 195 Attachment 1 EP-RM-004 Revision [X]Page 211 of 295 Category: Subcategory:

Initiating Condition: EAL: S -System Malfunction 8 -Hazardous Event Affecting Safety Systems Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode SA8.1 Alert The occurrence of any Table S-4 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-4 Hazardous Events* Seismic event (earthquake)

Internal or external FLOODING event" High winds or tornado strike" FIRE" EXPLOSION" Other events with similar hazard characteristics as determined by the Shift Manager Mode Applicability:

1 -Power Operations, 2 -Startup, 3 -Hot Shutdown Definition(s): EXPLOSION -A rapid, violent and catastrophic failure of a piece of equipment due to combustion, chemical reaction or overpressurization. A release of steam (from high energy lines or components) or an electrical component failure (caused by short circuits, grounding, arcing, etc.) should not automatically be considered an explosion. Such events require a post-event inspection to determine if the attributes of an explosion are present.Page 185 of 195 Attachment 1 EP-RM-004 Revision [X]Page 212 of 295 FIRE -Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred but is NOT required if large quantities of smoke and heat are observed.FLOODING -A condition where water is entering a room or area faster than installed equipment is capable of removal, resulting in a rise of water level within the room or area.SAFETY SYSTEM -A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2): Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the reactor coolant pressure boundary;(2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures. VISIBLE DAMAGE -Damage to a 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. Susquehanna Basis:* The significance of a seismic event is discussed under EAL HU2.1 (ref. 1, 2).* Internal FLOODING may be caused by events such as component failures, equipment misalignment, or outage activity mishaps (ref. 3, 4, 5).* Seismic Category I structures are analyzed to withstand a sustained, design wind velocity of at least 80 mph. (ref. 6, 7).* Areas containing functions and systems required for safe shutdown of the plant are identified by Fire Zone in the fire response procedure (ref. 8, 9)." 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.NEI 99-01 Basis: 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.EAL 1-.b.4The 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.EAL.! b .2The 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 Page 186 of 195 Attachment 1 EP-RM-004 Revision [X]Page 213 of 295 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 A-4RS1.Susquehanna Basis Reference(s):

1. ON-000-002 Severe Weather / Natural Phenomena 2. FSAR Section 3.7 Seismic Design 3. ON-169(269)-001 Flooding in Turbine Building 4. ON-169(269)-002 Flooding in Reactor Building 5. FSAR Section 3.4 Water Level (Flood) Design 6. FSAR Section 3.3 Wind and Tornado Loadings 7. FSAR Section 3.5 Missile Projection

8. SSES-FPRR Section 6.2 Fire Area Description

9. ON-013-001 Response to Fire 10. NEI 99-01 SA9 Page 187 of 195 Attachment I EP-RM-004 Revision [X]Page 214 of 295 Category E -Independent Spent Fuel Storage Installation (ISFSI)EAL Group: ANY (EALs in this category are applicable to any plant condition, hot or cold)An independent spent fuel storage installation (ISFSI) is a complex that is designed and constructed for the interim storage of spent nuclear fuel and other radioactive materials associated with spent fuel storage. A significant amount of the radioactive material contained within a cask/canister must escape its packaging and enter the biosphere for there to be a significant environmental effect resulting from an accident involving the dry storage of spent nuclear fuel. Formal offsite planning is not required because the postulated worst-case accident involving an ISFSI has insignificant consequences to the public health and safety.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.A hostile security event that leads to a potential loss in the level of safety of the ISFSI is a classifiable event under Security category EAL HS1.1.Minor surface damage that does not affect storage cask/canister boundary is excluded from the scope of these EALs.Page 188 of 195 Attachment 1 EP-RM-004 Revision [X]Page 215 of 295 Category:

E -ISFSI Sub-category: None Initiating Condition: Damage to a loaded cask CONFINEMENT BOUNDARY EAL: EUI.1 Notification of Unusual Event Damage to a loaded canister CONFINEMENT BOUNDARY as indicated by a radiation reading on a loaded spent fuel cask > any of the following:

800 mrem/hr at 3 ft from the HSM surface* 200 mrem/hr on contact on the outside of the HSM door centerline
40 mrem/hr on contact on the end shield wall exterior Mode Applicability:

All Definition(s): CONFINEMENT BOUNDARY-The barrier(s) between spent fuel and the environment once the spent fuel is processed for dry storage. As related to the Susquehanna ISFSI, Confinement Boundary is defined as the Dry Shielded Canister (DSC).Susquehanna Basis: The SSES Independent Spent Fuel Storage Installation utilizes the standardized NUHMOS Horizontal Modular System. The standardized NUHMOS system is a horizontal canister system composed of a steel dry shielded canister (DSC) and a reinforced concrete horizontal storage module (HSM). The DSC provides confinement and criticality control for the storage and transfer of irradiated fuel. The HSM houses the DSC and provides for heat removal. An HSM is considered loaded when it houses a DSC containing spent fuel. (ref. 1, 2)The values shown represent 2 times the limits specified in the ISFSI Certificate of Compliance Technical Specification 1.2.7, HSM Dose Rates (ref. 1).NEI 99-01 Basis: This IC addresses an event that results in damage to the CONFINEMENT BOUNDARY of a storage cask containing spent fuel. It applies to irradiated fuel that is licensed for dry storage beginning at the point that the loaded storage cask is sealed. The issues of concern are the creation of a potential or actual release path to the environment, degradation of one or more fuel assemblies due to environmental factors, and configuration changes which could cause challenges in removing the cask or fuel from storage.The existence of "damage" is determined by radiological survey. The technical specification multiple of "2 times", is also used in Recognition Categoy A R IC RAW, 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 Page 189 of 195 Attachment 1 EP-RM-004 Revision [X]Page 216 of 295 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 lCs HUI and HA1.Susquehanna Basis Reference(s):

1. Certification of Compliance No. 1004 for the NUHOMS Storage System 2. TRM B3.10.3 Independent Spent Fuel Storage Installation (ISFSI)3. NEI 99-01 E-HU1 Page 190 of 195 Attachment 1 EP-RM-004 Revision [X]Page 217 of 295 Category F -Fission Product Barrier Degradation EAL Group: Hot Conditions (RCS temperature

> 200 0 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 is the reactor coolant system pressure boundary and includes the RPV and all reactor coolant system piping up to and including the isolation valves.C. Containment (PC): The drywell, the wetwell, their respective interconnecting paths, and other connections up to and including the outermost containment isolation valves comprise the PC barrier. Primary Containment Barrier thresholds are used as criteria for escalation of the ECL from Alert to either 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 Primary Containment Barrier.* Unusual Event ICs associated with RCS and Fuel Clad Barriers are addressed under System Malfunction ICs.* For accident conditions involving a radiological release, evaluation of the fission product barrier thresholds will need to be performed in conjunction with dose assessments to Page 191 of 195 Attachment 1 EP-RM-004 Revision [X]Page 218 of 295 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 SSES 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 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, there should be frequent assessments of containment radioactive inventory and integrity. Alternatively, if both the Fuel Clad and RCS fission product barriers were potentially lost, the Emergency Director/Recovery Manager would have more assurance that there was no immediate need to escalate to a General Emergency. Page 192 of 195 Attachment 1 EP-RM-004 Revision [X]Page 219 of 295 Category: Fission Product Barrier Degradation Subcategory: N/A Initiating Condition: Loss of any two barriers and loss or potential loss of the third barrier EAL: FGI.1 General Emergency Loss of any two barriers AND Loss or potential loss of the third barrier (Table F-I)Mode Applicability: 1 -Power Operations, 2 -Startup, 3 -Hot Shutdown Definition(s): None Susquehanna 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 NEI 99-01 Basis: None Susquehanna Basis Reference(s):

1. NEI 99-01 FG1 Page 193 of 195 Attachment 1 EP-RM-004 Revision [X]Page 220 of 295 Category:

Fission Product Barrier Degradation Subcategory: N/A Initiating Condition: Loss or potential loss of any two barriers EAL: FSI.1 Site Area Emergency Loss or potential loss of any two barriers (Table F-I)Mode Applicability: 1 -Power Operations, 2 -Startup, 3 -Hot Shutdown Definition(s): None Susquehanna 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 Director/Recovery Manager would have greater assurance that escalation to a General Emergency is less imminent.NEI 99-01 Basis: None Susquehanna Basis Reference(s):

1. NEI 99-01 FS1 Page 194 of 195 Attachment 1 EP-RM-004 Revision [X]Page 221 of 295 Category:

Fission Product Barrier Degradation Subcategory: N/A Initiating Condition: Any loss or any potential loss of either Fuel Clad or RCS EAL: FAI.1 Alert Any loss or any potential loss of EITHER Fuel Clad or RCS (Table F-I)Mode Applicability: 1 -Power Operations, 2 -Startup, 3 -Hot Shutdown Definition(s): None Susquehanna Basis: Fuel Clad, RCS and Primary 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 NEI 99-01 Basis: None Susquehanna Basis Reference(s):

1. NEI 99-01 FA1 Page 195 of 195 Attachment 2 EP-RM-004 Revision [X]Page 222 of 295 ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Introduction Table F-1 lists the threshold conditions that define the Loss and Potential Loss of the three fission product barriers (Fuel Clad, Reactor Coolant System, and Containment).

The table is structured so that each of the three barriers occupies adjacent columns. Each fission product barrier column is further divided into two columns; one for Loss thresholds and one for Potential Loss thresholds. The first column of the table (to the left of the Fuel Clad Loss column) lists the categories (types)of fission product barrier thresholds. The fission product barrier categories are: A. RPV Level B. RCS Leak Rate C. Primary Containment Conditions D. Primary Containment Radiation / RCS Activity E. Primary Containment Integrity or Bypass F. ED/RM Jugement Each category occupies a row in Table F-1 thus forming a matrix defined by the categories. The intersection of each row with each Loss/Potential Loss column forms a cell in which one or more fission product barrier thresholds appear. If NEI 99-01 does not define a threshold for a barrier Loss/Potential Loss, the word "None" is entered in the cell.Thresholds are assigned sequential numbers within each Loss and Potential Loss column beginning with number one. In this manner, a threshold can be identified by its category title and number. For example, the first Fuel Clad barrier Loss in Category A would be assigned "FC Loss A.1," the third Containment barrier Potential Loss would be assigned "PC 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-i, 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 Page 1 of 65 Attachment 2 EP-RM-004 Revision [X]Page 223 of 295 and RCS barriers and a Potential Loss of the Containment barrier can occur. Barrier Losses and Potential Losses are then applied to the algorithms given in EALs FG1.1, FS1.1, and FA1.1 to determine the appropriate emergency classification. In the remainder of this Attachment, the Fuel Clad barrier threshold bases appear first, followed by the RCS barrier and finally the Containment barrier threshold bases. In each barrier, the bases are given according category Loss followed by category Potential Loss beginning with Category A, then B,..., F.Table F-2 provides a human factors enhancement mechanism to track the threshold conditions that define the Loss and Potential Loss of the three fission product barriers (Fuel Clad, Reactor Coolant System, and Containment) to assist with quickly determining which initiating condition for EALs FGI.1, FS1.1, or FA1.1 is met.Page 2 of 65 Attachment 2 EP-RM-004 Revision [X]Page 224 of 295 Table F-1 Fission Product Barrier Threshold Matrix Fuel Clad Barrier Reactor Coolant System Barrier Primary Containment Barrier Category Loss Potential Loss Loss Potential Loss Loss Potential Loss 1. RPV level CANNOT BE 1. RPV level CANNOT BE A 1. SAGs entered RESTORED AND MAINTAINED > RESTORED AND MAINTAINED >1. SAvs enter-161 in. -161 in. None None 1. SAGs enetered RPV Level or CANNOT be determined or CANNOT be determined

1. UNISOLABLE break in any of the 1. UNISOLABLE primary system 1. UNISOLABLE primary system 1. oig U A r n o leakage that results in exceeding leakage that results in exceeding follOwing:

EITHER of the following: EITHER of the following:

Main Steam Une
One or more Max Normal
One or more Max Safe* HPCI Steam Line Reactor Building Radiation Reactor Building Radiation* RCIC Steam Line Limits (EO-000-104 Table 9) Limits (EO-000-104 Table 9)B RWCU that can be read in the Control that can be read in the None None Feedwater Room (Table F-3) Control Room (Table F-5) None RCS Leak Rate OR OR OR One or more Max Normal
One or more Max Safe 2. Emergency RPV Depressurization Reactor Building area Reactor Building area is required temperature Limits temperature Limits (EO-000-104 Table 8) that can (EO-000-104 Table 8) that be read in the Control Room can be read in the Control (Table F-4) Room (Table F-6)1. UNPLANNED rapid drop in 1. Primary Containment pressure Primary Containment pressure > 53 psig following Primary Containment OR None None 1. Primary Containment pressure pressure rise 2. Deflagration concentrations exist PC > 1.72 psig due to RCS leakage Noninside PC (e2 26% AND 02 2:5%)Conditions OR 2. Primary Containment pressure response not consistent with 3. Heat Capacity Temperature Limit LOCA conditions (HCTL) exceeded 1. CHRRM radiation

> 3.0E+3 R/hr DI 1 OR 1. CHRRM radiation > 7.0E+0 R/hr PC Red I None with indications of RCS leakage None None 1. CHRRM radiation > 4.DE+4 R/hr RCS 2. Primary coolant activity > 300 inside the drywell Activity pCi/gm 1-131 dose equivalent

1. UNISOLABLE direct downstream pathway to the environment exists after Primary Containment E isolation signal PC Integrity None None None None OR None or Bypass 2. Intentional Primary Containment venting per EP-DS-004 RPV and PC Venting 1. nyconditionin theopinionof

1. Any condition in the opinion of the 1. Any condition in the opinion of the 1. Any condition in the opinion of the 1. Any condition in the opinion of the 1. Any condition in the opinion of the Director/Recovery Manager that Emergency Director/Recovery Emergency Director/Recovery Emergency Director/Recovery Emergency Director/Recovery Emergency Director/Recovery ED/RM indicates toss of the fuel dad Manager that indicates potential Manager that indicates loss of the Manager that indicates potential loss Manager that indicates loss of the Manager that indicates potential loss Judgment bifarer loss of the fuel clad barrier RCS barrier of the RCS barrier Primary Containment barrier of the Primary Containment barrier Page 3 of 65 Attachment 2 EP-RM-004 Revision [X]Page 225 of 295 Table F-2 Fission Product Barrier Status Table Circle the X's in the FGI: General Emergency FS1: Site Area Emergency FA1: Alert table below for all Loss of any two barriers Loss or potential loss of any two barriers Any loss or any potential loss applicable and loss or potential loss of of EITHER Fuel Clad or RCS situations.

Declare third barrier, barrier the EAL based upon all circled X's in any column.Fuel Clad -Loss X X X X X X X X Fuel Clad -Potential X X X X X X Loss RCS -Loss X X X X X X X X RCS -Potential Loss X X X X X X Primary Containment X X X X X X X-Loss Primary Containment X X X X X-Potential Loss Page 4 of 65 Attachment 2 EP-RM-004 Revision [X]Page 226 of 295 Barrier: Fuel Clad Category: A. RPV Level Degradation Threat: Loss Threshold:

1. SAGs entered Definition(s):

N/A Susquehanna Basis: EOPs specify the requirement for entry to the SAGs when core cooling is severely challenged. These EOPs provide instructions to ensure adequate core cooling by maintaining RPV water level above prescribed limits or operating sufficient RPV injection sources when level cannot be determined. SAG entry is required when (ref. 1, 2): " RPV water level CANNOT BE RESTORED AND MAINTAINED above -179 in.(MSCRWL)AND* RPV water level above -210 in. (jet pump suction) with at least one core spray loop injecting into the RPV at > 6350 gpm OR* TSC confirmation that core damage is progressing due to inadequate core cooling The above EOP conditions represent a challenge to core cooling and are the minimum values to assure core cooling without further degradation of the clad.SAGs entry is also a Potential Loss of the Primary Containment barrier (PC P-Loss A.1) which constitutes a Site Area Emergency. The threshold for the RCS barrier (RCS Loss A.1) should also be evaluated for escalation to a General Emergency if SAGs entry results in meeting that threshold. NEI 99-01 Basis: The Loss threshold represents the EOP requirement for primary contAinment floodinGSAG entry. This is identified in the BWROG EPGs/SAGs when the phrase, "Primr. Containmont .,,e,,,SAG Entry Is Required," appears. Since a site-specific RPV water level is not specified here,. the Loss threshold phrase, "Primary containment floodingSAG entry required," also accommodates the EOP need to flood the primary containment when RPV water level cannot be determined and core damage due to inadequate core cooling is believed to be occurring. Susquehanna Basis Reference(s):

1. EO-000-102 RPV Control 2. EO-000-1 14 RPV Flooding Page 5 of 65 Attachment 2 EP-RM-004 Revision [X]Page 227 of 295 3. NEI 99-01 RPV Water Level Fuel Clad Loss 2.A Page 6 of 65 Attachment 2 EP-RM-004 Revision [X]Page 228 of 295 Barrier: Fuel Clad Category:

A. RPV Level Degradation Threat: Potential Loss Threshold:

1. RPV level CANNOT BE RESTORED AND MAINTAINED

> -161 in. or CANNOT be determined Definition(s): N/A Susquehanna Basis: An RPV water level instrument reading of -161 in. indicates RPV level is at the top of active fuel (TAF) (ref. 1). When RPV level is at or above the TAF, the core is completely submerged. Core submergence is the most desirable means of core cooling. When RPV level is below TAF, the uncovered portion of the core must be cooled by less reliable means (i.e., steam cooling or spray cooling). If core uncovery is threatened, the EOPs specify alternate, more extreme, RPV water level control measures in order to restore and maintain adequate core cooling. Since core uncovery begins if RPV level drops to TAF, the level is indicative of a challenge to core cooling and the Fuel Clad barrier.This Fuel Clad barrier Potential Loss is met only after either: 1) the RPV has been depressurized, or required emergency RPV depressurization has been attempted, giving the operator an opportunity to assess the capability of low-pressure injection sources to restore RPV water level or 2) no low pressure RPV injection systems are available, precluding RPV depressurization in an attempt to minimize loss of RPV inventory. When RPV water level cannot be determined, EOPs require entry to EO-000-1 14, RPV Flooding. RPV water level indication provides the primary means of knowing if adequate core cooling is being maintained (ref. 2). When all means of determining RPV water level are unavailable, the fuel clad barrier is threatened and reliance on alternate means of assuring adequate core cooling must be attempted. The instructions in EO-000-1 14 specify these means, which include emergency depressurization of the RPV and injection into the RPV at a rate needed to flood to the elevation of the main steam lines or hold RPV pressure above the Minimum Steam Cooling Pressure (in scram-failure events. If RPV water level cannot be determined with respect to the top of active fuel, a potential loss of the fuel clad barrier exists.This threshold is intended to be interpreted the same as in EO-000-1 14, that is, a loss of instrumentation is not, by itself, a loss of ability to determine level.Note that EO-000-1 13, Level/Power Control, may require intentionally lowering RPV water level to -161 in. and control level between the Minimum Steam Cooling RPV Water Level (MSCRWL)and the top of active fuel (ref. 3). Although such action is a challenge to core cooling and the Fuel Clad barrier, the immediate need to reduce reactor power is the higher priority. For such events, ICs SA6 or SS6 will dictate the need for emergency classification. Page 7 of 65 Attachment 2 EP-RM-004 Revision [X]Page 229 of 295 NEI 99-01 Basis: This water level corresponds to the top of the active fuel and is used in the EOPs to indicate a challenge to core cooling.The RPV water level threshold is the same as RCS barrier Loss threshold 2A.I. Thus, this threshold indicates a Potential Loss of the Fuel Clad barrier and a Loss of the RCS barrier that appropriately escalates the emergency classification level to a Site Area Emergency. This threshold is considered to be exceeded when, as specified in the site-specific EOPs, RPV water CANNOT BE RESTORED AND MAINTAINED above the specified level following depressurization of the RPV (either manually, automatically or by failure of the RCS barrier) or when procedural guidance or a lack of low pressure RPV injection sources preclude Emergency RPV depressurization. EOPs allow the operator a wide choice of RPV injection sources to consider when restoring RPV water level to within prescribed limits. EOPs also specify depressurization of the RPV in order to facilitate RPV water level control with low-pressure injection sources. In some events, elevated RPV pressure may prevent restoration of RPV water level until pressure drops below the shutoff heads of available injection sources.Therefore, this Fuel Clad barrier Potential Loss is met only after either: 1) the RPV has been depressurized, or required emergency RPV depressurization has been attempted, giving the operator an opportunity to assess the capability of low-pressure injection sources to restore RPV water level or 2) no low pressure RPV injection systems are available, precluding RPV depressurization in an attempt to minimize loss of RPV inventory. The term "CANNOT BE RESTORED AND MAINTAINED above" means the value of RPV water level is not able to be brought above the specified limit (top of active fuel). The determination requires an evaluation of system performance and availability in relation to the RPV water level value and trend. A threshold prescribing declaration when a threshold value CANNOT BE RESTORED AND MAINTAINED above a specified limit does not require immediate action simply because the current value is below the top of active fuel, but does not permit extended operation below the limit; the threshold must be considered reached as soon as it is apparent that the top of active fuel cannot be attained.In high-power ATWS/failure to scram events, EOPs may direct the operator to deliberately lower RPV water level to the top of active fuel in order to reduce reactor power. RPV water level is then controlled between the top of active fuel and the Minimum Steam Cooling RPV Water Level (MSCRWL). Although such action is a challenge to core cooling and the Fuel Clad barrier, the immediate need to reduce reactor power is the higher priority. For such events, ICs SA6_ or SS65 will dictate the need for emergency classification. Since the loss of ability to determine if adequate core cooling is being provided presents a significant challenge to the fuel clad barrier, a potential loss of the fuel clad barrier is specified. Susquehanna Basis Reference(s):

1. EO-000-102 RPV Control 2. EO-000-114 RPV Flooding 3. EO-000-1 13 Level/Power Control 4. NEI 99-01 RPV Water Level Fuel Clad Potential Loss 2.A Page 8 of 65 Attachment 2 EP-RM-004 Revision [X]Page 230 of 295 Barrier: Category: Degradation Threat: Threshold:

Fuel Clad B. RCS Leak Rate Loss None Page 9 of 65 Attachment 2 EP-RM-004 Revision [X]Page 231 of 295 Barrier: Category: Degradation Threat: Threshold: Fuel Clad B. RCS Leak Rate Potential Loss None j Page 10 of 65 Attachment 2 EP-RM-004 Revision [X]Page 232 of 295 Barrier: Category: Degradation Threat: Threshold: Fuel Clad C. PC Conditions Loss None Page 11 of 65 Attachment 2 EP-RM-004 Revision [X]Page 233 of 295 Barrier: Category: Degradation Threat: Threshold: Fuel Clad C. PC Conditions Potential Loss None Page 12 of 65 Attachment 2 EP-RM-004 Revision [X]Page 234 of 295 Barrier: Fuel Clad Category: D. PC Radiation I RCS Activity Degradation Threat: Loss Threshold:

1. CHRRM radiation

> 3.0E+3 R/hr Definition(s): None Susquehanna Basis: Two gamma radiation levels ion chambers are located inside the drywell to monitor post-accident radiation levels. The detectors are located in the drywell on elevation 719'. They are always in service and read out on the C601 panel. Range is 100 to 10 8 R/hr. Environmentally qualified since they are required for use after an accident. A U-234 bug source is installed in the detector to serve as a self-check for instrument operability. The source provides approximately a 1 R/hr dose rate with the reactor shutdown. When the plant is at 100% power, drywell rad indication is normally about 3-4 R/hr. A reading of 3,000 R/hr indicates the release of reactor coolant into the drywell with elevated activity indicative of fuel clad damage (ref. 2).For a fuel failure event equivalent to approximately 1% of cladding failure and an instantaneous and complete release of reactor coolant to the primary containment, the response of the Containment High Range Radiation Monitors (CHRRM) in the drywell will be approximately 3,450 R/hr immediately after shutdown (rounded to 3,000 R/hr, which approximates the dose rate 10 minutes after shutdown). This assumes that the release has occurred soon after reactor shutdown, and that the fuel cladding failures produce a coolant source term of 300 pCi/gm of I-131 dose equivalent just prior to the release into primary containment. (ref. 2)NEI 99-01 Basis: The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the primary containment, assuming that reactor coolant activity equals 300 [tCi/gm dose equivalent 1-131. Reactor coolant activity above this level is greater than that expected for iodine spikes and corresponds to an approximate range of 21% 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 D.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 emergency classification level to a Site Area Emergency. There is no Fuel Clad Barrier Potential Loss threshold associated with Primary Containment Radiation. Page 13 of 65 Attachment 2 EP-RM-004 Revision [X]Page 235 of 295 Susquehanna Basis Reference(s):

1. General Atomic High-Range Gamma Radiation Monitoring System Manual 2. Calculation EC RADN 0525 Rev 2, "Estimation of Containment High Range Radiation Monitor Response to a Loss of Coolant Accident for Emergency Planning Purposes," January 8, 2008 3. NEI 99-01 Primary Containment Radiation Fuel Clad Loss 4.A Page 14 of 65 Attachment 2 EP-RM-004 Revision [X]Page 236 of 295 Barrier: Fuel Clad Category:

D. PC Radiation / RCS Activity Degradation Threat: Loss Threshold:

2. Primary coolant activity > 300 pCi/gm 1-131 dose equivalent Susquehanna Basis: The Fuel Clad Barrier shall be declared "lost" if the Primary coolant activity is determined to be> 300 pCi/gm 1-131 dose equivalent.

Two separate methods can be used make this determination: " Sample collection and analysis of reactor coolant activity" Analysis of plant parameters to determine fuel clad damage > 1%Sample collection and analysis of reactor coolant activity are accomplished in accordance with CH-ON-007. It is recognized that sample collection and analysis of reactor coolant with highly elevated activity levels could require several hours to complete. The fuel clad damage analysis methodology in ref. 3 provides an alternative method to determine if the 1-131 dose equivalent activity is > 300 pCi/gm.Fuels Engineering determines if_> 1% fuel clad damage has occurred based on the analysis methodology in ref. 3. Fuel clad damage equal to 1% corresponds to at least 300 micro-Ci/gm 1-131 dose equivalent in the reactor coolant, and drywell radiation values of at least 3000 R/hr during LOCA events (breach inside primary containment) (ref. 1). However, drywell radiation levels can be significantly lower than 3000 R/hr with a similar amount of fuel damage without a LOCA (no breach inside primary containment), since the fission products remain in the reactor vessel/do not escape into the drywell space and the CHRM is shielded from the radiation source (ref. 2 and 3).NEI 99-01 Basis: This threshold indicates that RCS radioactivity concentration is greater than 300 .Ci/gm dose equivalent 1-131. Reactor coolant activity above this level is greater than that expected for iodine spikes and corresponds to an approximate range of 2-1% 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.Susquehanna Basis Reference(s):

1. Calculation EC RADN 0525 Rev 2, "Estimation of Containment High Range Radiation Monitor Response to a Loss of Coolant Accident for Emergency Planning Purposes," January 8, 2008 2. EWR 1642196 -Investigation of EPLAN 1% Clad Damage vs. Barrier Loss Page 15 of 65 Attachment 2 EP-RM-004 Revision [X]Page 237 of 295 3. EP-PS-324

-Fuels Lead Engineer/Core Thermal Hydraulics Engineer 4. CH-ON-007 -Emergency Sampling Procedures

5. NEI 99-01 RCS Activity Fuel Clad Loss 1.A Page 16 of 65 Attachment 2 EP-RM-004 Revision [X]Page 238 of 295 Barrier: Category: Degradation Threat: Threshold:

Fuel Clad D. PC Radiation / RCS Activity Potential Loss None Page 17 of 65 Attachment 2 EP-RM-004 Revision [X]Page 239 of 295 Barrier: Category: Degradation Threat: Threshold: Fuel Clad E. PC Integrity or Bypass Loss None Page 18 of 65 Attachment 2 EP-RM-004 Revision [X]Page 240 of 295 Barrier: Category: Degradation Threat: Threshold: Fuel Clad E. PC Integrity or Bypass Potential Loss None Page 19 of 65 Attachment 2 EP-RM-004 Revision [X]Page 241 of 295 Barrier: Fuel Clad Category: F. ED/RM Judgment Degradation Threat: Loss Threshold:

1. Any condition in the opinion of the Emergency Director/Recovery Manager that indicates loss of the Fuel Clad barrier Definition(s):

None Susquehanna Basis: The Emergency Director/Recovery Manager judgment threshold addresses any other factors relevant to determining if the Fuel Clad barrier is lost. Such a determination should include imminent barrier degradation, barrier monitoring capability and dominant accident sequences.

Imminent barrier degradation exists if the degradation will likely occur within two hours based on a projection of current safety system performance.

The term "imminent" refers to recognition of the inability to reach safety acceptance criteria before completion of all checks.* Barrier monitoring capability is decreased if there is a loss or lack of reliable indicators. This assessment should include instrumentation operability concerns, readings from portable instrumentation and consideration of offsite monitoring results." Dominant accident sequences lead to degradation of all fission product barriers and likely entry to the EOPs. The Emergency Director/Recovery Manager should be mindful of the Loss of AC power (Station Blackout) and ATWS EALs to assure timely emergency classification declarations. NEI 99-01 Basis: This threshold addresses any other factors that are to be used by the Emergency Director/Recovery Manager in determining whether the Fuel Clad barrier is lost Susquehanna Basis Reference(s):

1. NEI 99-01 Emergency Director Judgment Fuel Clad Loss 6.A Page 20 of 65 Attachment 2 EP-RM-004 Revision [XI Page 242 of 295 Barrier: Fuel Clad Category:

F. ED/RM Judgment Degradation Threat: Potential Loss Threshold:

1. Any condition in the opinion of the Emergency Director/Recovery Manager that indicates potential loss of the Fuel Clad barrier Susquehanna Basis: The Emergency Director/Recovery Manager judgment threshold addresses any other factors relevant to determining if the Fuel Clad barrier is potentially lost. Such a determination should include imminent barrier degradation, barrier monitoring capability and dominant accident sequences.

Imminent barrier degradation exists if the degradation will likely occur within two hours based on a projection of current safety system performance.

The term "imminent" refers to recognition of the inability to reach safety acceptance criteria before completion of all checks.* Barrier monitoring capability is decreased if there is a loss or lack of reliable indicators. This assessment should include instrumentation operability concerns, readings from portable instrumentation and consideration of offsite monitoring results.* Dominant accident sequences lead to degradation of all fission product barriers and likely entry to the EOPs. The Emergency Director/Recovery Manager should be mindful of the Loss of AC power (Station Blackout) and ATWS EALs to assure timely emergency classification declarations. NEI 99-01 Basis: This threshold addresses any other factors that are to be used by the Emergency Director/Recovery Manager in determining whether the Fuel Clad barrier is potentially lost. The Emergency Director/Recovery Manager should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored. Susquehanna Basis Reference(s):

1. NEI 99-01 Emergency Director Judgment Potential Fuel Clad Loss 6.A Page 21 of 65 Attachment 2 EP-RM-004 Revision [X]Page 243 of 295 Barrier: Reactor Coolant System Category:

A. RPV Level Degradation Threat: Loss Threshold:

1. RPV level CANNOT BE RESTORED AND MAINTAINED

> -161 in. or CANNOT be determined Definition(s): None Susquehanna Basis: An RPV level instrument reading of -161 in. indicates RPV level is at the top of active fuel (TAF)(ref. 1). TAF is significantly lower than the normal operating RPV level control band. To reach this level, RPV inventory loss would have previously required isolation of the RCS and Primary Containment barriers, and initiation of all ECCS. If RPV level cannot be maintained above TAF, ECCS and other sources of RPV injection have been ineffective or incapable of reversing the decreasing level trend. The cause of the loss of RPV inventory is therefore assumed to be a LOCA. By definition, a LOCA event is a Loss of the RCS barrier.This RCS barrier Loss is met only after either: 1) the RPV has been depressurized, or required emergency RPV depressurization has been attempted, giving the operator an opportunity to assess the capability of low-pressure injection sources to restore RPV water level or 2) no low pressure RPV injection systems are available, precluding RPV depressurization in an attempt to minimize loss of RPV inventory. When RPV water level cannot be determined, EOPs require entry to EO-000-1 14, RPV Flooding. RPV water level indication provides the primary means of knowing if adequate core cooling is being maintained (ref. 2). When all means of determining RPV water level are unavailable, the RCS barrier is threatened and reliance on alternate means of assuring adequate core cooling must be attempted. The instructions in EO-000-1 14 specify these means, which include emergency depressurization of the RPV and injection into the RPV at a rate needed to flood to the elevation of the main steam lines or hold RPV pressure above the Minimum Steam Cooling Pressure (in scram-failure events. If RPV water level cannot be determined with respect to the top of active fuel, a loss of the RCS barrier exists. This threshold is intended to be interpreted the same as in EO-000-1 14, that is, a loss of instrumentation is not, by itself, a loss of ability to determine level.The conditions of this threshold are also a Potential Loss of the Fuel Clad barrier (FC P-Loss A. 1). A Loss of the RCS barrier and Potential Loss of the Fuel Clad barrier requires a Site Area Emergency classification. Note that EO-000-1 13, Level/Power Control, may require intentionally lowering RPV water level to -161 in. and control level between the Minimum Steam Cooling RPV Water Level (MSCRWL)and the top of active fuel (ref. 3). Although such action is a challenge to core cooling and the Page 22 of 65 Attachment 2 EP-RM-004 Revision [X]Page 244 of 295 Fuel Clad barrier, the immediate need to reduce reactor power is the higher priority. For such events, ICs SA6 or SS6 will dictate the need for emergency classification. NEI 99-01 Basis: This water level corresponds to the top of active fuel and is used in the EOPs to indicate challenge to core cooling.The RPV water level threshold is the same as Fuel Clad barrier Potential Loss threshold 2 7 A.1.Thus, this threshold indicates a Loss of the RCS barrier and Potential Loss of the Fuel Clad barrier and that appropriately escalates the emergency classification level to a Site Area Emergency. This threshold is considered to be exceeded when, as specified in the site-specific EOPs, RPV water CANNOT BE RESTORED AND MAINTAINED above the specified level following depressurization of the RPV (either manually, automatically or by failure of the RCS barrier) or when procedural guidance or a lack of low pressure RPV injection sources preclude Emergency RPV depressurization. EOPs allow the operator a wide choice of RPV injection sources to consider when restoring RPV water level to within prescribed limits. EOPs also specify depressurization of the RPV in order to facilitate RPV water level control with low-pressure injection sources. In some events, elevated RPV pressure may prevent restoration of RPV water level until pressure drops below the shutoff heads of available injection sources.Therefore, this RCS barrier Loss is met only after either: 1) the RPV has been depressurized, or required emergency RPV depressurization has been attempted, giving the operator an opportunity to assess the capability of low-pressure injection sources to restore RPV water level or 2) no low pressure RPV injection systems are available, precluding RPV depressurization in an attempt to minimize loss of RPV inventory. Susquehanna Basis Reference(s):

1. EO-000-102 RPV Control 2. EO-000-1 14 RPV Flooding 3. EO-000-1 13 Level/Power Control 4. NEI 99-01 RPV Water Level RCS Loss 2.A Page 23 of 65 Attachment 2 EP-RM-004 Revision [X]Page 245 of 295 Barrier: Category: Degradation Threat: Threshold:

Reactor Coolant System A. RPV Level Potential Loss SNone Page 24 of 65 Attachment 2 EP-RM-004 Revision [X]Page 246 of 295 Barrier: Reactor Coolant System Category: B. RCS Leak Rate Degradation Threat: Loss Threshold:

1. UNISOLABLE break in ANY of the following:

Main Steam Line* HPCI Steam Line* RCIC Steam Line* RWCU o Feedwater Definition(s):

UNISOLABLE -An open or breached system line that cannot be isolated, remotely or locally.* The term UNISOLABLE also includes any decision by plant staff or procedure direction to not isolate a primary system.* Normal leakage past a closed isolation valve is not considered UNISOLABLE leakage.Susquehanna Basis: As used in this threshold, local isolation actions can only be credited if isolation can be completed promptly (i.e. within 15 min.).In the case of a failure of both isolation valves to close but in which no downstream flow path exists, emergency declaration under this threshold would not be required. Similarly, if the emergency response requires the normal process flow of a system outside primary containment (e.g., EOP requirement to bypass MSIV low RPV water level interlocks and maintain the main condenser as a heat sink using main turbine bypass valves), the threshold is not met. The combination of these threshold conditions represent the loss of both the RCS and Primary Containment (see PC Loss E.1) barriers and justifies declaration of a Site Area Emergency (i.e., Loss or Potential Loss of any two barriers). High steam flow and high steam tunnel temperature annunciators are an indication of a Main Steam Line break. Either parameter will cause an isolation of the MSIVs. Note that the high steam flow alarm may clear if any of the MSIVs close and flow is reduced below the setpoint. If the high steam flow alarm was received (even though it was subsequently cleared) or there is other indication of high flow along with the high temperature alarm, the entry condition for this threshold has been met.Even though RWCU and Feedwater systems do not contain steam, they are included in the list because an UNISOLABLE break could result in the high-pressure discharge of fluid that is flashed to steam from relatively large volume systems directly connected to the RCS. Note: Each of the two feedwater injection lines is isolated from the reactor vessel via a series of swing Page 25 of 65 Attachment 2 EP-RM-004 Revision [X]Page 247 of 295 check valves. The ability of these check valves to isolate cannot be determined until after feedwater is no longer injecting into the reactor vessel.NEI 99-01 Basis: Large high-energy lines that rupture outside primary containment can discharge significant amounts of inventory and jeopardize the pressure-retaining capability of the RCS until they are isolated. If it is determined that the ruptured line cannot be promptly isolated from the Go R-eAm, the RCS barrier Loss threshold is met.Susquehanna Basis Reference(s):

1. FSAR Section 5.4.5 2. FSAR Section 6.3 3. FSAR Section 5.4.6 4. FSAR Section 10.4.7 5. FSAR Section 5.4.8 6. P&ID M-141 Nuclear Boiler 7. NEI 99-01 RCS Leak Rate RCS Loss 3.A Page 26 of 65 Attachment 2 EP-RM-004 Revision [X]Page 248 of 295 Barrier: Reactor Coolant System Category:

B. RCS Leak Rate Degradation Threat: Loss Threshold:

2. Emergency RPV Depressurization is required Definition(s):

N/A Susquehanna Basis: Plant symptoms requiring Emergency RPV Depressurization per the EOPs are indicative of a loss of the RCS barrier. If emergency depressurization is required, the plant operators are directed to open safety relief valves (SRVs). Even though the RCS is being vented into the suppression pool, a loss of the RCS exists due to the diminished effectiveness of the RCS pressure barrier to a release of fission products beyond its boundary.NEI 99-01 Basis: Emergency RPV Depressurization in accordance with the EOPs is indicative of a loss of the RCS barrier. If Emergency RPV Depressurization is performed, the plant operators are directed to open safety relief valves (SRVs) and keep them open. Even though the RCS is being vented into the suppression pool, a Loss of the RCS barrier exists due to the diminished effectiveness of the RCS to retain fission products within its boundary.Susquehanna Basis Reference(s):

1. EO-000-102 RPV Control 2. EO-000-103 Primary Containment Control 3. EO-000-104 Secondary Containment Control 4. EO-000-105 Radioactivity Release Control 5. EO-000-1 12 Emergency RPV Depressurization

6. EO-000-1 13 Level Power Control 7. EO-000-1 14 RPV Flooding 8. NEI 99-01 RCS Leak Rate RCS Loss 3.B Page 27 of 65 Attachment 2 EP-RM-004 Revision [X]Page 249 of 295 Barrier: Category: Reactor Coolant System B. RCS Leak Rate Degradation Threat: Potential Loss Threshold:

1. UNISOLABLE primary system leakage that results in exceeding EITHER of the following:

One or more Max Normal Reactor Building Radiation Limits (EO-000-104 Table 9) that can be read in the control room (Table F-3)OR* One or more Max Normal Reactor Building Area Temperature Limits (EO-000-1 04 Table 8) that can be read in the control room (Table F-4)Table F-3 Max Normal Reactor Building Radiation Limits RB Area ARM Channel Max Norm ARM NumberDecito Elevation (ft) Description Rad Limit 818 35 Cask Stor Area Hi Alarm 14 Spent Fuel Crit Mon 15 Refuel Floor North (South U2)42 Refuel Floor West 47 (44 U2) Spent Fuel Crit Mon 749 8 RWCU Recirc PP Access Hi Alarm 10 Fuel Pool PP Area 11 Rx Bldg Sample St 719 5 CRD North Hi Alarm 6 CRD South 670 16 Remote Shutdown Room Hi Alarm 645 3 HPCI PP & Turbine Room Hi Alarm 2 RCIC PP & Turbine Room 25 RHRA C PP Room 1 RHR B D PP Room 4 RB/RW Sump Area Page 28 of 65 Attachment 2 EP-RM-004 Revision [X]Page 250 of 295 Table F-4 Max Normal Reactor Building Temperature Limits RB Area Max Normal Elevation Area Temperature Temp (ft) (OF)749 RWCU-Pump Room 120 RWCU-Heat Exch Room 120 RWCU-Penetration Room 120 719 Main Steam Line Tunnel 157 683 HPCI Pipe Routing Area 120 RCIC Pipe Routing Area 120 645 HPCI-Equip Area 120 HPCI-Emerg Area Cooler 120 645 RCIC-Emerg Area Cooler 120 RCIC-Equip Area 120 645 RHR Equip Area 1 110 645 RHR Equip Area 2 110 Definition(s):

UNISOLABLE -An open or breached system line that cannot be isolated, remotely or locally." The term UNISOLABLE also includes any decision by plant staff or procedure direction to not isolate a primary system (e.g. EP-O00-104 has direction not to isolate systems if they are needed for EOP actions or damage control)* Normal leakage past a closed isolation valve is not considered UNISOLABLE leakage.Susquehanna Basis: This RCS Potential Loss threshold is limited to primary system leakage that results in exceeding one or more Max Normal Reactor Building Radiation or Temperature Limits that can be remotely determined from within the control room (ref. 1).NRC regulations require the SSES 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.Max Normal conditions shall be assumed to be from RCS leakage until proven otherwise.

The 15 minute classification requirement (EAL "trigger")

for this threshold begins when one or more of the above Max Normal Reactor Building Radiation or Temperature Limits are exceeded.Page 29 of 65 Attachment 2 EP-RM-004 Revision [X]Page 251 of 295" If subsequent actions taken to isolate the leak are successful within the 15 minute classificatibn period, this EAL should not be declared. (Note: EAL SU5.1 should be evaluated). o If subsequent investigation, within the 15 minute classification period, reveals that the Max Norm conditions are not due to RCS leakage, this EAL should not be declared.Note that a RCS leak could cause the fire suppression systems to actuate. If this occurs, the potential exists that the fire suppression systems could cause the area temperatures to be lower than the values specified in the EO-000-104 even though there is a RCS leak.* Once it is known that the cause of exceeding Max Norm temperatures is due to a FIRE or loss of ventilation then this threshold is not met. The applicable FIRE EAL should be evaluated.

If there is certainty that the fire suppression system actuation was caused by a RCS leak and NOT a fire then it is appropriate to judge that the MAX NORM temperature limits have been met even if the actual area temperatures are lower than those listed in the EO-000-104.

The presence of elevated general area temperatures or radiation levels in the secondary containment may be indicative of UNISOLABLE primary system leakage outside the primary containment. The Max Normal Reactor Building Limit values define this RCS threshold because they are the maximum normal operating values and signify the onset of abnormal system operation. When parameters reach this level, equipment failure or misoperation may be occurring. Elevated parameters may also adversely affect the ability to gain access to or operate equipment within the affected area. The locations into which the primary system discharge is of concern correspond to the areas addressed in EO-000-1 04, Secondary Containment Control, Table 8 that can be read in the control room (ref. 2).Cycling of SRVs to reduce primary system overpressure is not considered reactor coolant leakage. Inventory loss events, such as a stuck open SRV, venting and draining the RCS during cold shutdown or refueling, should not be considered when referring to "RCS leakage" because they are not indications of a break which could propagate. For these events entry into this threshold is not warranted however consideration should be given to RCS Loss -RCS Leak Rate threshold B2.In general, multiple indications should be used to validate that a primary system is discharging outside Primary Containment. For example, a high area radiation condition may be caused by radiation shine from nearby steam lines or the movement of radioactive materials. Conversely, a high area radiation condition in conjunction with other indications (e.g. room flooding, high area temperatures, reports of steam in the secondary containment, an unexpected rise in feedwater flowrate, or unexpected main turbine control valve closure) may indicate that a primary system is discharging into the secondary containment. Page 30 of 65 Attachment 2 EP-RM-004 Revision [X]Page 252 of 295 NEI 99-01 Basis: Potential loss of RCS based on primary system leakage outside the primary containment is determined from EOP temperature or radiation Max Normal Operating values in areas such as main steam line tunnel, RCIC, HPCI, etc., which indicate a direct path from the RCS to areas outside primary containment. A Max Normal Operating value is the highest value of the identified parameter expected to occur during normal plant operating conditions with all directly associated support and control systems functioning properly.The indicators reaching the threshold barriers and confirmed to be caused by RCS leakage from a primary system warrant an Alert classification. A primary system is defined to be the pipes, valves, and other equipment which connect directly to the RPV such that a reduction in RPV pressure will effect a decrease in the steam or water being discharged through an unisolated break in the system.An UNISOLABLE leak which ic indicated-by Max Normal Operating Valuesas described above escalates to a Site Area Emergency when combined with Containment Barrier Loss threshold ,3B.I1 (after a containment isolation) and a General Emergency when the Fuel Clad Barrier criteria is also exceeded.Susquehanna Basis Reference(s):

1. NCV 05000387; 388/2013005-04, Inadequate Instrumentation to Implement EALs for Fission Product Barrier Degradation

2. EO-000-104 Secondary Containment Control 3. NEI 99-01 RCS Leak Rate RCS Potential Loss 3.A Page 31 of 65 Attachment 2 EP-RM-004 Revision [X]Page 253 of 295 Barrier: Reactor Coolant System Category:

C. PC Conditions Degradation Threat: Loss Threshold:

1. Primary Containment pressure > 1.72 psig due to RCS leakage Definition(s):

None Susquehanna Basis: The drywell high pressure scram setpoint is an entry condition to EO-000-102, RPV Control, and EO-000-103, Primary Containment Control (ref. 1, 2). Normal primary containment pressure control functions (e.g., operation of drywell coolers, vent through SGT, etc.) are specified in EO-000-103 in advance of less desirable but more effective functions (e.g., operation of drywell or suppression pool sprays, etc.).In the design basis, primary containment pressures above the drywell high pressure scram setpoint are assumed to be the result of a high-energy release into the containment for which normal pressure control systems are inadequate or incapable of reversing the increasing pressure trend. Pressures of this magnitude, however, can be caused by non-LOCA events such as a loss of drywell cooling or inability to control primary containment vent/purge (ref. 3).The threshold phrase "...due to RCS leakage" focuses the barrier failure on the RCS instead of the non-LOCA malfunctions that may adversely affect primary containment pressure. PC pressure greater than 1.72 psig with corollary indications (e.g., drywell temperature, indications of loss of RCS inventory) should therefore be considered a Loss of the RCS barrier. Loss of drywell cooling that results in pressure greater than 1.72 psig should not be considered an RCS barrier Loss.NEI 99-01 Basis: The (site specific value) PFiFmay pressur1 .72 psiq is the drywell high pressure setpoint which indicates a LOCA by automatically initiating the ECCS or equivalent makeup system.There is no Potential Loss threshold associated with Primary Containment Pressure.Susquehanna Basis Reference(s):

1. EO-000-102 RPV Control 2. EO-000-103 Primary Containment Control 3. FSAR Section 6.2.1 Primary Containment Functional Design 4. NEI 99-01 Primary Containment Pressure RCS Loss 1.A Page 32 of 65 Attachment 2 EP-RM-004 Revision [X]Page 254 of 295 Barrier: Category: Degradation Threat: Threshold:

Reactor Coolant System C. PC Conditions Potential Loss None Page 33 of 65 Attachment 2 EP-RM-004 Revision [X]Page 255 of 295 Barrier: Reactor Coolant System Category: D. PC Radiation / RCS Activity Degradation Threat: Loss Threshold:

1. CHRRM radiation

> 7.OE+0 R/hr with indication of a RCS leak inside the drywell Definition(s): N/A Susquehanna Basis: Two gamma radiation levels ion chambers are located inside the drywell to monitor post-accident rad levels. The detectors are located in the drywell on elevation 719'. They are always in service and read out on the C601 panel. Range is 100 to 108 R/hr. Environmentally qualified since they are required for use after an accident. A U-234 bug source is installed in the detector to serve as a self-check for instrument operability. The source provides approximately a 1 R/hr dose rate with the reactor shutdown (ref. 1). When the plant is at 100% power, drywell rad indication is normally about 3-4 R/hr. Containment High Range Radiation Monitor (CHRRM)readings of approximately 3 R/hr indicate an instantaneous release of reactor coolant at normal operating concentrations of 1-131 to the drywell atmosphere. Adding this value to the normal CHRRM background readings of 3-4 R/hr (100% power normal operation) provides the value of 7 R/hr. (ref. 2)Indication of a RCS leak into the drywell is added to qualify the radiation monitor indication to avoid declaring the loss of the RCS barrier for situations where the radiation increase is not due to a primary system leak. For situations that involve failure of the Fuel Clad Barrier alone, containment Radiation levels would increase to greater than 30 R/hr potentially giving a false indication of a loss of the RCS barrier. Therefore the EAL contains a qualifier to preclude over classification of the event if only the fuel clad barrier has failed. Indication of a leak should be determined by observing other containment indications such as sump level, drywell pressure and ambient temperature. NEI 99-01 Basis: The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the primary containment, assuming that reactor coolant activity equals Technical Specification allowable limits. This value is lower than that specified for Fuel Clad Barrier Loss threshold D.1 since it indicates a loss of the RCS Barrier only.There is no RCS Barrier Potential Loss threshold associated with Primary Containment Radiation. Susquehanna Basis Reference(s):

1. General Atomic High-Range Gamma Radiation Monitoring System Manual Page 34 of 65 Attachment 2 EP-RM-004 Revision [X]Page 256 of 295 2. Calculation EC RADN 0525 Rev 2, "Estimation of Containment High Range Radiation Monitor Response to a Loss of Coolant Accident for Emergency Planning Purposes," January 8, 2008 3. NEI 99-01 Primary Containment Radiation RCS Loss 4.A Page 35 of 65 Attachment 2 EP-RM-004 Revision [X]Page 257 of 295 Barrier: Category: Degradation Threat: Threshold:

Reactor Coolant System D. PC Radiation / RCS Activity Potential Loss None Page 36 of 65 Attachment 2 EP-RM-004 Revision [X]Page 258 of 295 Barrier: Category: Degradation Threat: Threshold: Reactor Coolant System E. PC Integrity or Bypass Loss None Page 37 of 65 Attachment 2 EP-RM-004 Revision [X]Page 259 of 295 Barrier: Category: Degradation Threat: Threshold: Reactor Coolant System E. PC Integrity or Bypass Potential Loss None Page 38 of 65 Attachment 2 EP-RM-004 Revision [X]Page 260 of 295 Barrier: Reactor Coolant System Category: F. ED/RM Judgment Degradation Threat: Loss Threshold:

1. Any condition in the opinion of the Emergency Director/Recovery Manager that indicates loss of the RCS barrier Definition(s):

None Susquehanna Basis: The Emergency Director/Recovery Manager 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 two hours based on a projection of current safety system performance.

The term "imminent" refers to the recognition of the inability to reach safety acceptance criteria before completion of all checks." Barrier monitoring capability is decreased if there is a loss or lack of reliable indicators. This assessment should include instrumentation operability concerns, readings from portable instrumentation and consideration of offsite monitoring results." Dominant accident sequences lead to degradation of all fission product barriers and likely entry to the EOPs. The Emergency Director/Recovery Manager should be mindful of the Loss of AC power (Station Blackout) and ATWS EALs to assure timely emergency classification declarations. NEI 99-01 Basis: This threshold addresses any other factors that may be used by the Emergency Director/Recovery Manager in determining whether the RCS Barrier is lost.Susquehanna Basis Reference(s):

1. NEI 99-01 Emergency Director Judgment RCS Loss 6.A Page 39 of 65 Attachment 2 EP-RM-004 Revision [X]Page 261 of 295 Barrier: Reactor Coolant System Category:

F. ED/RM Judgment Degradation Threat: Potential Loss Threshold:

1. Any condition in the opinion of the Emergency Director/Recovery Manager that indicates potential loss of the RCS barrier Definition(s):

None Susquehanna Basis: The Emergency Director/Recovery Manager 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 two hours based on a projection of current safety system performance. The term "imminent" refers to the inability to reach final safety acceptance criteria before completing all checks.* Barrier monitoring capability is decreased if there is a loss or lack of reliable indicators. This assessment should include instrumentation operability concerns, readings from portable instrumentation and consideration of offsite monitoring results." Dominant accident sequences lead to degradation of all fission product barriers and likely entry to the EOPs. The Emergency Director/Recovery Manager should be mindful of the Loss of AC power (Station Blackout) and ATWS EALs to assure timely emergency classification declarations. NEI 99-01 Basis: This threshold addresses any other factors that may be used by the Emergency Director/Recovery Manaqer in determining whether the RCS Barrier is potentially lost. The Emergency Director/Recovery Manager should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored. Susquehanna Basis Reference(s):

1. NEI 99-01 Emergency Director Judgment RCS Potential Loss 6.A Page 40 of 65 Attachment 2 EP-RM-004 Revision [X]Page 262 of 295 Barrier: Category: Degradation Threat: Threshold:

Primary Containment A. RPV Level Loss None Page 41 of 65 Attachment 2 EP-RM-004 Revision [X]Page 263 of 295 Barrier: Primary Containment Category: A. RPV Level Degradation Threat: Potential Loss Threshold:

1. SAGs entered Definition(s):

None Susquehanna Basis: EOPs specify the requirement for entry to the SAGs when core cooling is severely challenged. These EOPs provide instructions to ensure adequate core cooling by maintaining RPV water level above prescribed limits or operating sufficient RPV injection sources when level cannot be determined. SAG entry is required when (ref. 1, 2):* RPV water level CANNOT BE RESTORED AND MAINTAINED above -179 in.(MSCRWL)AND" RPV water level CANNOT BE RESTORED AND MAINTAINED above -210 in. (jet pump suction) with at least one core spray loop injecting into the RPV at > 6350 gpm OR* TSC confirmation that core damage is progressing due to inadequate core cooling The above EOP conditions represent a challenge to core cooling and are the minimum values to assure core cooling without further degradation of the clad.SAGs entry is also a Potential Loss of the Fuel Clad barrier (FC P-Loss A. 1) which constitutes a Site Area Emergency. The threshold for the RCS barrier (RCS Loss A.1) should also be evaluated for escalation to a General Emergency if SAGs entry results in meeting that threshold. NEI 99-01 Basis: The Potential Loss threshold is identical to the Fuel Clad Loss RPV Water Level threshold 2-.AA.1. The Potential Loss requirement for Pimar.,' Containment FloodiegSAG entry is required indicates adequate core cooling CANNOT BE RESTORED AND MAINTAINED and that core damage is possible. BWR EPGs/SAGs specify the conditions that require primary containmet floe entry. When p.rma. y containment f.. ,dingS A entry is required, the EPGs are exited and SAGs are entered. Entry into SAGs is a logical escalation in response to the inability to restore and maintain adequate core cooling.PRA studies indicate that the condition of this Potential Loss threshold could be a core melt sequence which, if not corrected, could lead to RPV failure and increased potential for primary Page 42 of 65 Attachment 2 EP-RM-004 Revision (X]Page 264 of 295 containment failure. In conjunction with the RPV water level Loss thresholds in the Fuel Clad and RCS barrier columns, this threshold results in the declaration of a General Emergency. Susquehanna Basis Reference(s):

1. EO-000-1 02 RPV Control 2. EO-000-1 14 RPV Flooding 3. NEI 99-01 RPV Water Level PC Potential Loss 2.A Page 43 of 65 Attachment 2 EP-RM-004 Revision [X]Page 265 of 295 Barrier: Category: Degradation Threat: Primary Containment B. RCS Leak Rate Loss Threshold:

1. UNISOLABLE primary system leakage that results in exceeding EITHER of the following:

One or more Max Safe Reactor Building Radiation Limits (EO-000-104 Table 9) that can be read in the control room (Table F-5)OR* One or more Max Safe Reactor Building area temperature Limits (EO-000-104 Table 8) that can be read in the control room (Table F-6)Table F-5 Max Safe Reactor Building Radiation Limits Max Safe Rad RB Area ARM Number ARM Channel Limit Elevation (ft) Description (RIHR)818 49 Refuel Floor Area 10 749 52 RWCU Recirc PP Access 10 54 Fuel Pool PP Area 719 50 CRD North 10 51 CRD South 670 53 Remote Shutdown Room 10 645 48 HPCI PP & Turbine Room 10 57 RCIC PP & Turbine Room 55 RHRA C PP Room 56 RHR B D PP Room Page 44 of 65 Attachment 2 EP-RM-004 Revision [X]Page 266 of 295 Table F-6 Max Safe Reactor Building Temperature Limits RB Area Max Safe Elevation Area Temperature Temp (ft) (OF)749 RWCU-Pump Room 147 RWCU-Heat Exch Room 147 RWCU-Penetration Room 131 719 Main Steam Line Tunnel 177 683 HPCI Pipe Routing Area 167 RCIC Pipe Routing Area 167 645 HPCI-Equip Area 167 HPCI-Emerg Area Cooler 167 645 RCIC-Emerg Area Cooler 167 RCIC-Equip Area 167 645 RHR Equip Area 1 142 645 RHR Equip Area 2 142 Definition(s):

UNISOLABLE -An open or breached system line that cannot be isolated, remotely or locally.* The term UNISOLABLE also includes any decision by plant staff or procedure direction to not isolate a primary system (e.g. EP-O00-104 has direction not to isolate systems if they are needed for EOP actions or damage control)* Normal leakage past a closed isolation valve is not considered UNISOLABLE leakage.Susquehanna Basis: This Primary Containment Loss threshold is limited to primary system leakage that results in exceeding one or more Max Safe Reactor Building Radiation or Temperature Limits that can be remotely determined from within the control room (ref. 1).NRC regulations require the SSES 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.The 15 minute classification requirement (EAL "trigger") for this threshold begins when one or more of the above Max Safe Reactor Building Radiation or Temperature Limits are exceeded.Max Safe conditions shall be assumed to be from RCS leakage until proven otherwise. Page 45 of 65 Attachment 2 EP-RM-004 Revision [XI Page 267 of 295* If subsequent actions taken to isolate the leak are successful, the threshold for this EAL is still met and must be declared. Once leakage is isolated, downgrading the emergency may be appropriate.

If subsequent investigation, within the 15 minute classification period, reveals that the Max Safe conditions are not due to RCS leakage, this EAL should not be declared.Note that a RCS leak could cause the fire suppression systems to actuate. If this occurs the potential exists that the fire suppression systems could cause the area temperatures to be lower than the values specified in the EO-000-104 even though there is a RCS.* Once it is known that the cause of exceeding MAX SAFE temperatures is due to a FIRE or loss of ventilation then this threshold is not met. The applicable FIRE EAL should be evaluated.
If there is certainty that the fire suppression system actuation was caused by a RCS leak and NOT a fire then it is appropriate to judge that the MAX NORM temperature limits have been met even if the actual area temperatures are lower than those listed in the EO-000-104 The presence of elevated general area temperatures or radiation levels in the secondary containment may be indicative of UNISOLABLE primary system leakage outside the primary containment.

The Maximum Safe Reactor Building Limit values define this Containment barrier threshold because they are indicative of problems in the secondary containment that are spreading and pose a threat to achieving a safe plant shutdown. This threshold addresses problematic discharges outside primary containment that may not originate from a high-energy line break. The locations into which the primary system discharge is of concern correspond to the areas addressed in EO-000-104, Secondary Containment Control, Table 8 that can be read in the control room (ref. 2).In general, multiple indications should be used to validate that a primary system is discharging outside Primary Containment. For example, a high area radiation condition may be caused by radiation shine from nearby steam lines or the movement of radioactive materials. Conversely, a high area radiation condition in conjunction with other indications (e.g. room flooding, high area temperatures, reports of steam in the secondary containment, an unexpected rise in feedwater flowrate, or unexpected main turbine control valve closure) may indicate that a primary system is discharging into the secondary containment. NEI 99-01 Basis: The Max Safe Operating Temperature and the Max Safe Operating Radiation Level are each the highest value of these parameters at which neither: (1) equipment necessary for the safe shutdown of the plant will fail, nor (2) personnel access necessary for the safe shutdown of the plant will be precluded. EOPs utilize these temperatures and radiation levels to establish conditions under which RPV depressurization is required.The temperatures and radiation levels should be confirmed to be caused by RCS leakage from a primary system. A primary system is defined to be the pipes, valves, and other equipment which connect directly to the RPV such that a reduction in RPV pressure will effect a decrease in the steam or water being discharged through an unisolated break in the system.Page 46 of 65 Attachment 2 EP-RM-004 Revision [X]Page 268 of 295 In combination with RCS Potential Loss &AB.1 this threshold would result in a Site Area Emergency. There is no Potential Loss threshold associated with Primary Containment Isolation Failure.Susquehanna Basis Reference(s):

1. NCV 05000387; 388/2013005-04, Inadequate Instrumentation to Implement EALs for Fission Product Barrier Degradation

2. EO-000-104 Secondary Containment Control 3. NEI 99-01 RCS Leak Rate PC Loss 3.C Page 47 of 65 Attachment 2 EP-RM-004 Revision [X]Page 269 of 295 Barrier: Category: Degradation Threat: Threshold:

Primary Containment B. RCS Leak Rate Potential Loss None I Page 48 of 65 Attachment 2 EP-RM-004 Revision [X]Page 270 of 295 Barrier: Primary Containment Category: C. PC Conditions Degradation Threat: Loss Threshold:

1. UNPLANNED rapid drop in Primary Containment pressure following Primary Containment pressure rise Definition(s):

UNPLANNED -A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.Susquehanna Basis: None NEI 99-01 Basis: Rapid UNPLANNED loss of primary containment pressure (i.e., not attributable to drywell spray or condensation effects) following an initial pressure increase indicates a loss of primary containment integrity. Primar,' containment pressure sho-uld i n.rease as a result of mass and energy release into the primary containment from. a L-OCA. Thus, primary containment pressure not inrGeasing undcr there cGondtons indicates a loss of primary containment itgiy Tiese-This thresholds rely-relies on operator recognition of an unexpected response for the condition and therefore a specific value is not assigned. The unexpected (UNPLANNED) response is important because it is the indicator for a containment bypass condition. Susquehanna Basis Reference(s):

1. NEI 99-01 Primary Containment Conditions PC Loss 1.A Page 49 of 65 Attachment 2 EP-RM-004 Revision [X]Page 271 of 295 Barrier: Primary Containment Category:

C. PC Conditions Degradation Threat: Loss Threshold:

2. Primary Containment pressure response not consistent with LOCA conditions Definition(s):

None Susquehanna Basis: The calculated pressure response of the containment is shown in Figure 6.2-2 (ref. 1)(reproduced on next page). The maximum calculated drywell pressure (63.3 psia or 48.6 psig)is well below the design allowable pressure of 53 psig (ref. 2).Due to conservatisms in LOCA analyses, actual pressure response is expected to be less than the analyzed response. For example, blowdown mass flowrate may be only 60-80% of the analyzed rate.NEI 99-01 Basis: Rapid UNPLANNED loss Of Primnar' containment pressuro (i.e., not attributablo to dr,'well spray Or condensation effocts) following an initial proeGGure inrGease-indicates a loss of primnary integrit. Primary containment pressure should increase as a result of mass and energy release into the primary containment from a LOCA. Thus, primary containment pressure not increasing under these conditions indicates a loss of primary containment integrity. T-heee-This threshold& fety-relies on operator recognition of an unexpected response for the condition and therefore a specific value is not assigned. The unexpected (UNPLANNED) response is important because it is the indicator for a containment bypass condition. Susquehanna Basis Reference(s):

1. FSAR Figure 6.2-2 2. FSAR Section 6.2.1.1.3.1

3. NEI 99-01 Primary Containment Conditions PC Loss 1.B Page 50 of 65 Attachment 2 EP-RM-004 Revision [X]Page 272 of 295 Short-Term RSLB Pressure Response 70 60 50.940 30 V 20 10 0 ifW Pres6-VVW Press 20 0 10 15 Time (se-)25 30 FSAR REV..65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 AND 2 FINAL SAFETY ANALYSIS REPORT PRESSURE RESPONSE FOR RECIRCULATION LINE BREAK'FRGUqE 6.2-2, Rev. 56 Auto Cad: Flgurm Fear 6._2.d2dg Page 51 of 65 Attachment 2 EP-RM-004 Revision [X]Page 273 of 295 Barrier: Containment Category:

C. PC Conditions Degradation Threat: Potential Loss Threshold:

1. Primary Containment pressure > 53 psig Definition(s):

None Susquehanna Basis: When the primary containment exceeds the maximum allowable value (53 psig) (ref. 1), primary containment venting may be required even if offsite radioactivity release rate limits will be exceeded (ref. 2). The drywell and suppression chamber maximum allowable value of 53 psig is based on the primary containment design pressure as identified in the SSES accident analysis.If this threshold is exceeded, a challenge to the containment structure has occurred because assumptions used in the accident analysis are no longer valid and an unanalyzed condition exists. This constitutes a Potential Loss of the Containment barrier even if a containment breach has not occurred.NEI 99-01 Basis: The threshold pressure is the primary containment internal design pressure. Structural acceptance testing demonstrates the capability of the primary containment to resist pressures greater than the internal design pressure. A pressure of this magnitude is greater than those expected to result from any design basis accident and, thus, represent a Potential Loss of the Containment barrier.Susquehanna Basis Reference(s):

1. FSAR Section 6.2.1.1.3.1

2. EO-000-103 Primary Containment Control 3. NEI 99-01 Primary Containment Conditions PC Potential Loss I.A Page 52 of 65 Attachment 2 EP-RM-004 Revision [X]Page 274 of 295 Barrier: Containment Category:

C. PC Conditions Degradation Threat: Potential Loss Threshold:

2. Deflagration concentrations exist inside PC (H 2 6% AND 02 5%)Definition(s):

None Susquehanna Basis: Deflagration (explosive) mixtures in the primary containment are assumed to be elevated concentrations of hydrogen and oxygen. BWR industry evaluation of hydrogen generation for development of EPGs/SAGs indicates that any hydrogen concentration above minimum detectable is not to be expected within the short term. Post-LOCA hydrogen generation primarily caused by radiolysis is a slowly evolving, long-term condition. Hydrogen concentrations that rapidly develop are most likely caused by metal-water reaction. A metal-water reaction is indicative of an accident more severe than accidents considered in the plant design basis and would be indicative, therefore, of a potential threat to primary containment integrity. Hydrogen concentration of approximately 6% is considered the global deflagration concentration limit (ref. 1).Except for brief periods during plant startup and shutdown, oxygen concentration in the primary containment is maintained at insignificant levels by nitrogen inertion. The specified values for this Potential Loss threshold are the minimum global deflagration concentration limits (6%hydrogen and 5% oxygen) and readily recognizable because 6% hydrogen is well above the EO-000-103, Primary Containment Control, entry condition (ref. 2, 3). Values above the hydrogen/oxygen concentrations (6% and 5%, respectively) require intentional primary containment venting, which is defined to be a Loss of Containment (PC Loss E.2).Hydrogen and oxygen concentration in the drywell during normal operation. These monitors are isolated by accident isolation signals. However, monitors CAC-AT-4409 and 4410 will be realigned to the primary containment for post-accident monitoring via an operator actuated isolation signal override circuit when directed by the EOPs.NEI 99-01 Basis: If hydrogen concentration reaches or exceeds the lower flammability limit, as defined in plant EOPs, in an oxygen rich environment, a potentially explosive mixture exists. If the combustible mixture ignites inside the primary containment, loss of the Containment barrier could occur.Susquehanna Basis Reference(s):

1. BWROG EPG/SAG Revision 3, Subsection PC/G 2. EO-000-103 Primary Containment Control 3. EP-DS-001 Containment Combustible Gas Control Page 53 of 65 Attachment 2 EP-RM-004 Revision [X]Page 275 of 295 4. NEI 99-01 Primary Containment Conditions PC Potential Loss 1..B Page 54 of 65 Attachment 2 EP-RM-004 Revision [X]Page 276 of 295 Barrier: Category: Containment C. PC Conditions Potential Loss Degradation Threat: Threshold:

3. Heat Capacity Temperature Limit (Figure -HCTL) exceeded Figure -HCTL Heat Capacity Temperature Limit 12 13 14 15 16 17 18 19 20 21 22 23 24 25 26 27 28 29 30 31 32 33 34 35 36 37 38 SUPPRESSION POOL LEVEL (FP)Definition(s):

None Susquehanna Basis: This threshold is met when the final step of section SP/T in EO-000-1 03, Primary Containment Control, is reached (ref. 1).Page 55 of 65 Attachment 2 EP-RM-004 Revision [X]Page 277 of 295 NEI 99-01 Basis: The Heat Capacity Temperature Limit (HCTL) is the highest suppression pool temperature from which Emergency RPV Depressurization will not raise: Suppression chamber temperature above the maximum temperature capability of the suppression chamber and equipment within the suppression chamber which may be required to operate when the RPV is pressurized, OR Suppression chamber pressure above Primary Containment Pressure Limit A, while the rate of energy transfer from the RPV to the containment is greater than the capacity of the containment vent.The HCTL is a function of RPV pressure, suppression pool temperature and suppression pool water level. It is utilized to preclude failure of the containment and equipment in the containment necessary for the safe shutdown of the plant and therefore, the inability to maintain plant parameters below the limit constitutes a potential loss of containment. Susquehanna Basis Reference(s):

1. EO-000-103 Primary Containment Control 2. NEI 99-01 Primary Containment Conditions PC Potential Loss 1.C Page 56 of 65 Attachment 2 EP-RM-004 Revision [X]Page 278 of 295 Barrier: Primary Containment Category:

D. PC Radiation / RCS Activity Degradation Threat: Loss Threshold: None Page 57 of 65 Attachment 2 EP-RM-004 Revision [XI Page 279 of 295 Barrier: Primary Containment Category: D. PC Radiation / RCS Activity Degradation Threat: Potential Loss Threshold:

1. CHRRM radiation

> 4.OE+4R/hr Definition(s): None Susquehanna Basis: Two gamma radiation levels ion chambers are located inside the drywell to monitor post-accident radiation levels. The detectors are located in the drywell on elevation 719'. They are always in service and read out on the C601 panel. Range is 100 to 108 R/hr. Environmentally qualified since they are required for use after an accident. A U-234 bug source is installed in the detector to serve as a self-check for instrument operability. The source provides approximately a 1 R/hr dose rate with the reactor shutdown (ref. 1). When the plant is at 100% power, Containment High Range Radiation Monitor (CHRRM) indication is normally about 3-4 R/hr. A reading of 40,000 R/hr indicates the release of reactor coolant into the drywell with elevated activity indicative of 20% fuel clad damage (ref. 2).An analysis of the Containment High Range Radiation Monitor (CHRRM) response to a Loss-Of-Coolant Accident (LOCA) is given in reference

2. Results are summarized in Reference 2, Section 2 for the case of 1% clad damage. For this threshold a reactor shutdown time of 1 hour is conservatively assumed. Although CHRRM data is always available for emergency planning, a one hour shutdown time is conservative for times less than one hour and is applicable to the timing of the sequence of events for a LOCA that would result in both the release of reactor coolant activity to containment and damage to 20% of the fuel clad. From reference 2, Section 2, the calculated CHRRM dose rate at one hour after shutdown for a complete loss of reactor coolant activity to containment with 1% clad damage is 2160 R/hr. Multiplying this value by a factor of 20 to account for 20% clad damage results in a CHRRM dose rate of 43,200 R/hr. This value is rounded to 40,000 R/hr for human factors considerations.

A Containment High Range Rad Monitor reading > 40,000 R/hr is a value which indicates significant fuel damage well in excess of that required for loss of RCS and Fuel Clad. A major failure of fuel cladding which allows radioactive material to be released from the core into the reactor coolant could result in a major release of radioactivity requiring offsite protective actions.Regardless of whether containment is challenged, this amount of activity in containment corresponding to a CHHRM reading > 40,000 R/hr, if released, could have such severe consequences that it is prudent to treat this as a potential loss of containment, such that a General Emergency declaration is warranted. In order to reach this Containment barrier Potential Loss threshold, a loss of the RCS barrier (RCS Loss D.1) and a loss of the Fuel Clad barrier (FC Loss D.1) have already occurred. This threshold, therefore, represents at a General Emergency classification. Page 58 of 65 Attachment 2 EP-RM-004 Revision [X]Page 280 of 295 NEI 99-01 Basis: The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the primary 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 emergency classification level to a General Emergency. Susquehanna Basis Reference(s):

1. General Atomic High-Range Gamma Radiation Monitoring System Manual 2. Calculation EC RADN 0525 Rev 2, "Estimation of Containment High Range Radiation Monitor Response to a Loss of Coolant Accident for Emergency Planning Purposes," January 8, 2008 3. NEI 99-01 Primary Containment Radiation Fuel Clad Potential Loss 1.D Page 59 of 65 Attachment 2 EP-RM-004 Revision [X]Page 281 of 295 Barrier: Primary Containment Category:

E. PC Integrity or Bypass Degradation Threat: Loss Threshold:

1. UNISOLABLE direct downstream pathway to the environment exists after Primary Containment isolation signal Definition(s):

UNISOLABLE-An open or breached system line that cannot be isolated, remotely or locally.* The term UNISOLABLE also includes any decision by plant staff or procedure direction to not isolate a primary system (e.g. EP-O00-104 has direction not to isolate systems if they are needed for EOP actions or damage control)* Normal leakage past a closed isolation valve is not considered UNISOLABLE leakage.Susquehanna Basis: This threshold addresses failure of open isolation devices which should close upon receipt of a manual or automatic containment isolation signal resulting in a significant radiological release pathway directly to the environment. The concern is the unisolable open pathway to the environment. A failure of the ability to isolate any one line with a pathway directly to the environment indicates a breach of primary containment integrity. Examples include:* This EAL is applicable if plant operators attempt to close isolation valves before any automatic setpoints are reached, and both valves fail to close AND a downstream pathway to the environment exists. If subsequent actions are successful, downgrading the emergency may be appropriate.

Main Steam Line breaks with two MSIVs in one line failing to close.* HPCI, RCIC or RWCU steam line breaks with all isolation valves in one line failing to close.* UNISOLABLE containment atmosphere vent paths." Automatic closure of both isolation valves in one line are disabled AND an isolation signal occurs AND a downstream pathway exits.The adjective "direct" modifies "release pathway" to discriminate against release paths through interfacing liquid systems. The following examples do not meet the EAL threshold:
If the main condenser is available with an UNISOLABLE main intact steam line, there may be releases through the steam jet air ejectors and gland seal exhausters.

These pathways are monitored, however, and do not meet the intent of an UNISOLABLE direct release path to the environment. These minor releases are assessed using the Category R, Abnormal Rad Release / Rad Effluent, EALs.* Leakage into a closed system (e.g. RHR, Core Spray) is to be considered only if the closed system is breached and thereby creates a significant pathway to the environment.

Normal leakage past a closed isolation valve is not considered UNISOLABLE leakage.Page 60 of 65 Attachment 2 EP-RM-004 Revision [X]Page 282 of 295" SCRAM Discharge Volume valves fail to close. The SDV drains to the Reactor Building Sump, which creates the pathway to the environment.
The existence of an in-line charcoal filter (SGTS) does not make a release path indirect since the filter is not effective at removing fission noble gases. Typical filters have an efficiency of 95-99% removal of iodine. Given the magnitude of the core inventory of iodine, significant releases could still occur. In addition, since the fission product release would be driven by boiling in the reactor vessel, the high humidity in the release stream can be expected to render the filters ineffective in a short period." EO-000-103, Primary Containment Control, Section PC/P may specify primary containment venting and intentional bypassing of the containment isolation valve logic, even if offsite radioactivity release rate limits are exceeded (ref. 1). Under these conditions with a valid containment isolation signal, the Containment barrier should be considered lost.Declaration of this EAL threshold constitutes a radiological release in progress.NEI 99-01 Basis: The use of the modifier "direct" in defining the release path discriminates against release paths through interfacing liquid systems or minor release pathways, such as instrument lines, not protected by the Primary Containment Isolation System (PCIS).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.Following the leakage of RCS mass into primary containment and a rise in primary containment pressure, there may be minor radiological releases associated with allowable primary containment leakage through various penetrations or system components. Minor releases may also occur if a primary containment isolation valve(s) fails to close but the primary containment atmosphere escapes to an enclosed system. These releases do not constitute a loss or potential loss of primary containment but should be evaluated using the Recognition Category A R ICs.Susquehanna Basis Reference(s):

1. EO-000-103 Primary Containment Control 2. NEI 99-01 Primary Containment Isolation Failure PC Loss 3.A Page 61 of 65 Attachment 2 EP-RM-004 Revision [X]Page 283 of 295 Barrier: Primary Containment Category:

E. PC Integrity or Bypass Degradation Threat: Loss Threshold:

2. Intentional Primary Containment venting per EP-DS-004 RPV and PC Venting Definition(s):

None Susquehanna Basis: EO-000-103, Primary Containment Control, may specify primary containment venting and intentional bypassing of the containment isolation valve logic, even if offsite radioactivity release rate limits are exceeded (ref. 1). The threshold is met when the operator begins venting the primary containment in accordance with EP-DS-004, RPV and PC Venting, not when actions are taken to bypass interlocks prior to opening the vent valves (ref. 2).Because the containment vent valves are not qualified for opening/reclosure in a post-accident environment, there is no guarantee that venting, once initiated, can be terminated. Thus it is assumed that once the vents are opened with a source term, they are not re-closed (ref. 3).NEI 99-01 Basis: EOPs may direct primary containment isolation valve logic(s) to be intentionally bypassed, even if offsite radioactivity release rate limits will be exceeded. Under these conditions with a valid primary containment isolation signal, the containment should also be considered lost if primary containment venting is actually performed. Intentional venting of primary containment for primary containment pressure or combustible gas control to the secondary containment and/or the environment is a Loss of the Containment. Venting for primary containment pressure control when not in an accident situation (e.g., to control pressure below the drywell high pressure scram setpoint) does not meet the threshold condition. Susquehanna Basis Reference(s):

1. EO-000-103 Primary Containment Control 2. EP-DS-004 RPV and PC Venting 3. NL-98-036, SSES Safety Evaluation for EP-DS-004, Primary Containment and RPV Venting 4. NEI 99-01 Primary Containment Isolation Failure PC Loss 3.B Page 62 of 65 Attachment 2 EP-RM-004 Revision [X]Page 284 of 295 Barrier: Category: Degradation Threat: Threshold:

Primary Containment E. PC Integrity or Bypass Potential Loss None Page 63 of 65 Attachment 2 EP-RM-004 Revision [X]Page 285 of 295 Barrier: Primary Containment Category: F. ED/RM Judgment Degradation Threat: Loss Threshold:

1. Any condition in the opinion of the Emergency Director/Recovery Manager that indicates loss of the Primary Containment barrier Definition(s):

None Susquehanna Basis: The Emergency Director/Recovery Manager judgment threshold addresses any other factors relevant to determining if the Primary Containment barrier is lost. Such a determination should include imminent barrier degradation, barrier monitoring capability and dominant accident sequences.

Imminent barrier degradation exists if the degradation will likely occur within two hours based on a projection of current safety system performance.

The term "imminent" refers to recognition of the inability to reach safety acceptance criteria before completion of all checks.* Barrier monitoring capability is decreased if there is a loss or lack of reliable indicators. This assessment should include instrumentation operability concerns, readings from portable instrumentation and consideration of offsite monitoring results.* Dominant accident sequences lead to degradation of all fission product barriers and likely entry to the EOPs. The Emergency Director/Recovery Manager should be mindful of the Loss of AC power (Station Blackout) and ATWS EALs to assure timely emergency classification declarations. NEI 99-01 Basis: This threshold addresses any other factors that may be used by the Emergency Director/Recovery Manager in determining whether the Primary Containment Barrier is lost.Susquehanna Basis Reference(s):

1. NEI 99-01 Emergency Director Judgment PC Loss 6.A Page 64 of 65 Attachment 2 EP-RM-004 Revision [X]Page 286 of 295 Barrier: Primary Containment Category:

F. ED/RM Judgment Degradation Threat: Potential Loss Threshold:

1. Any condition in the opinion of the Emergency Director/Recovery Manager that indicates potential loss of the Primary Containment barrier Definition(s):

None Susquehanna Basis: The Emergency Director/Recovery Manager judgment threshold addresses any other factors relevant to determining if the Primary Containment barrier is potentially lost. Such a determination should include imminent barrier degradation, barrier monitoring capability and dominant accident sequences." Imminent barrier degradation exists if the degradation will likely occur within two hours based on a projection of current safety system performance. The term "imminent" refers to recognition of the inability to reach safety acceptance criteria before completion of all checks.* Barrier monitoring capability is decreased if there is a loss or lack of reliable indicators. This assessment should include instrumentation operability concerns, readings from portable instrumentation and consideration of offsite monitoring results.* Dominant accident sequences lead to degradation of all fission product barriers and likely entry to the EOPs. The Emergency Director/Recovery Manager should be mindful of the Loss of AC power (Station Blackout) and ATWS EALs to assure timely emergency classification declarations. NEI 99-01 Basis: This threshold addresses any other factors that may be used by the Emergency Director/Recovery Manager in determining whether the Primary Containment Barrier is lost.Susquehanna Basis Reference(s):

1. NEI 99-01 Emergency Director Judgment PC Loss 6.A Page 65 of 65 Attachment 3 EP-RM-004 Revision [X]Page 287 of 295 ATTACHMENT 3 Safe Operation

& Shutdown Areas Tables R-2 & H-2 Bases Background NEI 99-01 Revision 6 ICs AA3 (SSES RA3.2) and HA5 (SSES HA5.1) prescribe declaration of an Alert based on impeded access to rooms or areas (due to either area radiation levels or hazardous gas concentrations) where equipment necessary for normal plant operations, cooldown or shutdown is located. These areas are intended to be plant operating mode dependent. Specifically the Developers Notes For AA3 and HA5 states: The "site-specific list of plant rooms or areas with entry-related mode applicability identified" should specify those rooms or areas that contain equipment which require a manual/local action as specified in operating procedures used for normal plant operation, cooldown and shutdown. Do not include rooms or areas in which actions of a contingent or emergency nature would be performed (e.g., an action to address an off-normal or emergency condition such as emergency repairs, corrective measures or emergency operations). In addition, the list should specify the plant mode(s) during which entry would be required for each room or area.The list should not include rooms or areas for which entry is required solely to perform actions of an administrative or record keeping nature (e.g., normal rounds or routine inspections). Further, as specified in IC HA5: The list need not include the Control Room if adequate engineered safety/design features are in place to preclude a Control Room evacuation due to the release of a hazardous gas.Such features may include, but are not limited to, capability to draw air from multiple air intakes at different and separate locations, inner and outer atmospheric boundaries, or the capability to acquire and maintain positive pressure within the Control Room envelope.Site-Specific Analysis The Control Room is excluded from Table R-2 since it is included EAL RA3.1 for all modes.NEI 99-01 Rev. 6 developer notes state that the control room does not need to be included Table R-2 for this reason.The Control Room is included in Table H-2 for all modes since adequate engineered safety/design features are not in place to preclude a Control Room evacuation due to the release of a hazardous gas at SSES.The remaining list of mode dependent rooms in Tables R-2 and H-2 is a list of rooms where actions are absolutely required to be performed to move the plant from normal operations through cool down and to achieve and maintain cold shutdown. These areas are not areas requiring entry to solely meet surveillance requirements. This does not include actions called out for in the procedures that are not absolutely needed to move the plant into and maintain cold shutdown (e.g. shutting down turbine lube oils systems, opening of system drains). In order to transition from normal operations (Modes 1 and 2) to hot shutdown (Mode 3) the only location required is the Control Room since the plant is able to be placed into hot shutdown from the control room without the need for any other room entry.Page 1 of 3 Attachment 3 EP-RM-004 Revision [X]Page 288 of 295 The remaining rooms/areas shown in in Tables R-2 and H-2 were determined based on the discussion section above and Modes 1 and 2 were excluded. Therefore, the analysis starting point was a review the steps of GO-100/200-005 "PLANT SHUTDOWN TO HOT/COLD SHUTDOWN" after the Unit is has reached Mode 3. The analysis then branched to other procedure steps described in the GO's if those procedures steps were not excluded from consideration as noted in the discussion section. The analysis concluded that Reactor Building areas and elevations listed in in Tables R-2 and H-2 contain equipment that must be manipulated in Mode 3 to achieve and maintain cold shutdown. Modes 4 and 5 are also included for these areas to account for swapping loops for long term operation of shutdown cooling.Tables R-2 and H-2 Bases A review of station operating procedures identified the following mode dependent in-plant actions and associated areas that are required for normal plant operation, cooldown or shutdown Unit I RA3.2 / HA5.1 Elevation Area(s)670 27 683 27, 28 & 29 703 28 & 29 719 25 & 29 749 25 & 29 729* 12,21 Mode(s)3,4,5 3,4,5 3,4,5 3,4,5 3,4,5 1, 2,3,4,5,D Mode(s)3,4,5 3,4,5 3,4,5 3,4,5 3,4,5 1, 2,3,4,5,D Unit 2 RA3.2 Elevation 670 683 703 719 749 729 */ HA5.1 Area(s)32 32, 33 & 34 33 & 34 30 & 34 32 & 33 12, 21* Control Room -only applicable to HA5 Page 2 of 3 Attachment 3 EP-RM-004 Revision [X]Page 289 of 295 Table R-2 & H-2 Results Table R-2 & H-2 Safe Operation & Shutdown Areas Elevation Unit I Area(s) ** Unit 2 Area(s) ** Mode(s)670' RB 27 RB 32 3,4,5 683' RB 27, 28, 29 RB 32, 33, 34 3,4,5 703' RB 28, 29 RB 33, 34 3,4,5 719' RB 25, 29 RB 30, 34 3,4,5 749' RB 25, 29 RB 32, 33 3,4,5 729' CS 12, 21* CS 12, 21* 1, 2, 3,4, 5, D* Control Room -only applicable to HA5** See Chart 1 for the specific locations of areas listed in Table R-2 and H-2.Chart 1- Plant Area Key Plan r__ LOW LEVEL WATER TREATMENT SR RADWASTE H SPRAY POND ACIDAND VALVE VAULT CHLORINE 152 47 BUILDING PUMPHOUSE ESSW 46 (2q8 ?3 9? 12 PUMPHOUSEH UNIT#2 EI' NIT#1 ?TURBINE BLDG. TURBINE BLDG. RADWASTE STURB. 16 115 14 113 4 3 2 1 3.[38 37 t 20 24 19 1 18 117 8 17 6 5 40 39 42 41 23 22 1 21 12 1 11 10 9 (36 44 35 43 COND & REF STORAGE=.,DIESEL GENERATOR I R IVER INTAKE 57 STRUCTURE E#2 E#1 REACTOR REACTOR+E DIESEL GENERATOR Page 3 of 3 Attachment 4 EP-RM-004 Revision [X]Page 290 of 295 Form EP-RM-004-R Attachment 5 EP-RM-004 Revision [X]Page 291 of 295~C-4CIflCICC(WI1 v- T..-b ,-E --.10 -C.CCC-o,0w Ca1WC 1 501.1(C C.C 9CCO~d NoteC~1 sC¶I dI.ICCC01Cl CCCC*CIfOW-~ D*invel equipment drain tank Drywall sumps Reactor Building sump LRW coltection taink Main condenser hotwet Suppression pool visual observation CONTAINMENT CLOSURE not established (Note 6)* PC hydrogen concentration > 6%* UNPLANNED rise in PC pressure* Exceeding one or more Secondary Containment Control Mac Safe Radiation Level. (EO-E000 04 Table 9) that Can be read in the Control Room (Table C-6)ha' n CST 0 tCrnC ICCC CIeIIC CICIOCCO CCC I ICC 0ot Cal --.T C cCaC"aa-E 'System OnCslts ORO NRC UHF Radio X Plant PA System X Dedicated Conference Unes X Comnurerclst Telephone Systems X X X Celluar Telephone X X FTS-2001 (ENS) X X SeisnocC event (earthquake) Internal or external FLOODING event High wInds or tornado strike FIRE EXPLOSION Other events wIth similar hazard characteristicC as determined by the Shift Manager 1E~E Wa 05Cm -aM Cot C.CCICCC 119 CC ~CCS CI C CC~CsO5..S IC-CI C.9O10CC.CI 500109 ¶01W IC Modes: r-L-T [I2Z [iZ-Z [I4Z r-- [I I Susquehanna-w-. m,, ýýn -w.. p- I Nuclear, LLC COLD CONDrnONS Form EP-RM-004-C Attachment 6 EP-RM-004 Revision [X]Page 292 of 295 GENERL .J EMREC IEAE MRGNY AETUUULVN I*7a$tEAcfl~~.~m*06 OtMO 62 WA7. 07 0. A0.,A, 620 i i ,f t sA AcTAO...mbnso.ty.6a&mu, I m Uan...0.a A -1 1 346.OOW60nA~~O.0 A4.m H A 5 -AHimn0.00A7n.0nACP ? 27 0 772 H bA* A AfI0 O fA A 027 Modes:= F2 ] [ ] W-4--- 1 S I DI 00.AO0),A20 5227 -2062. -22.22. CA72A 20-Susquehanna Nuclear, LLC ALL CONDITIONS Form EP-RM-004-H Attachment 7 EP-RM-004 Revision [XI Page 293 of 295 t1 b, b . ssul.4513.bf 0543,[.0435.3".- -I --2-4-.34 S 3 ,-me-4 5-r a CT) N M= - -:W .W -_.------, 003 563 4l3 7o.5 4-~~40~~~e--5 2 Notes eTA- U"0 U.6s.6 TeA"Wb 154 F.-ft.-- -- --- -Modes: Form F z3z ii4zi Fz5z [i-z Susquehanna Pý op su" -SNI-050 o." Nuclear, LLCI HOT CONDMONS (RCS > 2001F)Form EP-RM-004-S Attachment 8 EP-RM-004 Revision [X]Page 294 of 295 H1 Modes: F-2-1 [III] F-4] F[Y 1 Fý [ Susquehanna M ds Hý ý -" -I -Nuclear, LLC ALL CONDITIONS Form EP-RM-004-E Attachment 9 EP-RM-004 Revision [X]Page 295 of 295-- I 1.... .... 57201 22-12520072 120-- ....... T- --275025201 SOOCO21 0i i .s~2.272 52, 7252574--u--n-lx0m --ta. *520520200, 5201102 2072 ____ sm.100 70 22 70 p2.,l20,0l0~ 11051 CR202172 10 070 52 022070020077220 2-20 51 52c25052~#25p 141*10 7% 0-m 2 9 I M Susquehanna W.Modes: i-I F F-4[- 1Nuclear, LLC HOTCONDITlONS Nuler ... (RCS > 2001F)Form EP-RM-004-F}}

NEI 99-01, Susquehanna, Units 1 and 2 - Evaluation of Proposed Change Regarding Proposed License Amendment to Adopt Emergency Action Level Scheme

Text

5.4 Conclusions

1.0 DESCRIPTION

5.0 REGULATORY EVALUATION

5.3 Conclusions

7.0 REFERENCES

4.0 REFERENCES

2.0 DISCUSSION

2.1 Background

4.0 REFERENCES

4.1 Developmental

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