ML18260A366

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Response to Request for Additional Information Related to the Application for Adoption of Emergency Action Level Schemes Pursuant to NEI 99-01, Revision 6
ML18260A366
Person / Time
Site: Arkansas Nuclear Entergy icon.png
Issue date: 09/17/2018
From: Anderson R L
Entergy Operations
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
0CAN091801
Download: ML18260A366 (730)


Contents

Text

10 CFR 50.90

0CAN091801

September 17, 2018

ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555

SUBJECT:

Response to Request for Additional Information Related to the Application for Adoption of Emergency Action Level Schemes Pursuant to NEI 99-01, Revision 6 Arkansas Nuclear One, Units 1 and 2 Docket Nos. 50-313 and 50-368 License Nos. DPR-51 and NPF-6

Dear Sir or Madam:

By letter dated March 29, 2018 (Reference 1), Entergy Operations, Inc. (Entergy), requested NRC approval of a proposed change to the Arkansas Nuclear One, Units 1 and 2 (collectively referred to as ANO) licenses. The proposed change involved revising the Emergency Plan for ANO to adopt the Nuclear Energy Institute's (NEI's) revised Emergency Action Level (EAL) scheme described in NEI 99-01, Revision 6, "Development of Emergency Action Levels for Non-Passive Reactors."

By email dated August 2, 2018 (Reference 2), the NRC informed Entergy that additional information is needed to support the Staff's continued review of the application. A clarification call between the NRC and the licensee was previously held on August 1, 2018. Reference 2 requires a response no later than September 17, 2018. Enclosure 1 of this letter includes a summary of the request for additional information (RAI) and Entergy's response to each question.

The Entergy response requires changes to certain enclosures included in the original Reference 1 application. For completeness and ease of NRC review, the following Reference 1 enclosures are being resubmitted in full as part of Entergy's response to the NRC's RAI:

2. Proposed EAL Technical Basis Document (Markup) 3. Proposed EAL Technical Basis Document (Clean)
4. NEI 99-01, Rev. 6, Deviations and Differences, ANO Units 1 and 2
5. Proposed EAL Matrix Chart and Review Table (for information only) 6. Supporting Referenced Document Pages (EP-CALC-ANO-1701 only) Entergy Operations, Inc. 1448 S.R. 333 Russellville, AR 72802 Tel 479-858-3110 Richard L. Anderson ANO Site Vice President

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No new regulatory commitments are included in this letter.

In accordance with 10 CFR 50.91, Entergy is notifying the State of Arkansas of this amendment request by transmitting a copy of this letter and enclosure to the designated State Official.

If there are any questions or if additional information is needed, please contact Stephenie Pyle

at 479-858-4704.

I declare under penalty of perjury that the foregoing is true and correct. Executed on September 17, 2018.

Sincerely, ORIGINAL SIGNED BY RICHARD L. ANDERSON

RLA/dbb

Enclosures:

1. Response to Request for Additional Information - ANO EAL Revisions 2. Proposed EAL Technical Basis Document (Markup)
3. Proposed EAL Technical Basis Document (Clean)
4. NEI 99-01, Rev. 6, Deviations and Differences, ANO Units 1 and 2 5. Proposed EAL Matrix Chart and Review Table (for information only) 6. Supporting Referenced Document Pages

REFERENCES:

1. Entergy letter dated March 29, 2018, License Amendment Request - Adoption of Emergency Action Level Schemes Pursuant to NEI 99-01, Revision 6, Arkansas Nuclear One, Units 1 and 2, 0CAN031801 (ML18088B412)
2. NRC email dated August 2, 2018, ANO-1 and 2 Final RAI RE: License Amendment Request to Adopt EAL Scheme Change Per NEI 99-01 Revision 6, (EPID L-2018-LLA-0082) (0CNA081801) (ML18218A221)

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cc: Mr. Kriss Kennedy Regional Administrator U. S. Nuclear Regulatory Commission

Region IV 1600 East Lamar Boulevard Arlington, TX 76011-4511 NRC Senior Resident Inspector

Arkansas Nuclear One P. O. Box 310 London, AR 72847 U. S. Nuclear Regulatory Commission Attn: Mr. Thomas Wengert MS O-08B1A One White Flint North 11555 Rockville Pike

Rockville, MD 20852 Mr. Bernard R. Bevill

Arkansas Department of Health Radiation Control Section 4815 West Markham Street

Slot #30 Little Rock, AR 72205

Enclosure 1 to 0CAN091801 Response to Request for Additional Information ANO EAL Revisions to 0CAN091801

Page 1 of 19 Response to Request for Additional Information ANO EAL Revisions By letter dated March 29, 2018 (Reference 1), Entergy Operations, Inc. (Entergy), requested NRC approval of a proposed change to the Arkansas Nuclear One, Units 1 and 2 (collectively referred to as ANO) licenses. The proposed change involved revising the Emergency Plan for ANO to adopt the Nuclear Energy Institute's (NEI's) revised Emergency Action Level (EAL)

scheme described in NEI 99-01, Revision 6, "Development of Emergency Action Levels for Non-Passive Reactors."

By email dated August 2, 2018 (Reference 2), the NRC informed Entergy that additional information is needed to support the Staff's continued review of the application. A clarification call between the NRC and the licensee was previously held on August 1, 2018. Reference 2 requires a response no later than September 17, 2018. The following includes a summary of the request for additional information (RAI) and Entergy's response to each question.

The Entergy response to several questions contained herein requires changes to certain enclosures included in the original Reference 1 application. For completeness and ease of NRC review, the following Reference 1 enclosures are being resubmitted in full as part of

Entergy's response to the NRC's RAI:

2. Proposed EAL Technical Basis Document (Markup) 3. Proposed EAL Technical Basis Document (Clean)
4. NEI 99-01, Rev. 6, Deviations and Differences, ANO Units 1 and 2
5. Proposed EAL Matrix Chart and Review Table (for information only) 6. Supporting Referenced Document Pages

ANO RAI 1 Concerning plant or procedure changes that could impact the ANO EAL scheme, please address the following:

a. Section 4.7, "EAL/Threshold References to AOP [Abnormal Operating Procedure]

and EOP [Emergency Operating Procedure] Setpoints/Criteria," of NEI 99-01, Revision 6, states:

As reflected in the generic guidance, the criteria/values used in several EALs and fission product barrier thresholds may be drawn from a plant's AOPs and EOPs. This approach is intended to maintain good alignment between operational diagnoses and emergency classification assessments. Developers should verify that appropriate administrative controls are in place to ensure that a subsequent change to an AOP or EOP is screened to determine if an evaluation pursuant to 10 CFR 50.54(q) is required.

Please explain what controls are in place at ANO to ensure that a subsequent change to an AOP or EOP is screened to determine if an evaluation pursuant to 10 CFR 50.54(q) is

required.

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Page 2 of 19 Entergy Response:

Entergy fleet-wide procedure EN-LI-100, "Process Applicability Determination," provides a method to determine impacts to licensing basis documents (LBDs) and processes when changes are proposed to activities/procedures, including changes to AOPs and EOPs. The Process Applicability Determination (PAD) form specifically questions the effect of a proposed change on the Emergency Plan and the associated EALs, along with all other LBDs. When impacts are identified, the procedure requires review and approval by Emergency Planning department personnel via a 10 CFR 50.54(q) review. Personnel can then determine whether a change to the applicable EAL is required.

b. Proposed EALs AA3.2 and HA5.1 are applicable only during Modes 3 and 4, based on current site operational requirements. Please explain what process is in place to ensure that plant equipment or procedural changes would be adequately screened to ensure that the EAL would be appropriately modified if needed.

Entergy Response:

In general, changes to procedures and plant modifications fall under the PAD process as described in response to Part 'a' above. Plant modifications are also governed by the Entergy Engineering Change (EC) process. This process not only assesses impacts to LBDs, regulation, commitments, etc., but a PAD is required for all commercial and nuclear plant modifications. Engineering equivalencies may not require a PAD, but reviews similar to those required by the PAD form are perform ed and documented for equivalencies within the EC package. Since a PAD (and if necessary, a 10 CFR 50.59 evaluation) must be performed for nearly all modifications, any impact to the Emergency Plan and EALs would be readily identified, as discussed in Part 'a' above.

ANO RAI 2 Section 4.4, "Presentation of Scheme Information to Users," of NEI 99-01, Revision 6, provides that an alternative method for presenting EAL scheme information may be developed for use, provided that it contains all the information needed to make a correct emergency classification. This information includes the Initiating Conditions (ICs), Operating Mode Applicability criteria, EALs, and Notes. The licensee provided an EAL Matrix Chart and Review Table (EAL Matrix) as an alternative presentation method. However, the EAL Matrix is not consistent with the proposed EAL Technical Basis document. This could lead to inaccurate or delayed emergency classifications. A partial list of examples of inconsistencies follows:

a. The proposed EAL basis document Table 1A-1, "Unit 1 Effluent Monitor Classification Thresholds," provides that radiation monitor "RX-9830" is the fuel handling area release point monitor, while the EAL Matrix provides that radiation monitor "RX-9820" is the fuel handling area release point monitor.
b. The proposed EAL basis document Table 2A-1, "Unit 2 Effluent Monitor Classification Thresholds," provides that radiation monitor "RX-9830" is the fuel handling area release point monitor, while the EAL Matrix provides that radiation monitor "RX-9820" is the fuel handling area release point monitor. Additionally, monitors 2RE-2330 (BMS Liquid Discharge), 2RE-4423 (Regenerative Waste Discharge), and 2RX-9820 (Containment Purge) are not consistent with the ANO EAL Basis document values. to 0CAN091801

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c. The proposed EAL basis document Table 2A-2, "Unit 1[2] Fuel Damage Radiation Monitors," provides radiation monitor 2RE-8915 (Spent Fuel Area). However, there is no

corresponding monitor on the EAL Matrix. Additionally, the EAL Matrix shows 2RE-9825 as the instrument numbers for both the radwaste area monitors, as well as the containment high range monitor.

d. The EAL Matrix Fission Product Barriers threshold value for an Alert has FS1.1 as an EAL identifier vice FA1.1.
e. The proposed EAL Matrix shows that HS6.1 is only applicable when defueled vice in all modes. The ANO EAL Basis document shows all modes.

Please explain how the apparent differences between the EAL Matrix and the EAL basis document would not present human factors issues that could impact timely and accurate EAL assessments, or revise accordingly to address. (Note: the above items are intended to highlight NRC staff concerns and should not be considered as a complete list of potential issues.)

Entergy Response:

Entergy concurs that any differences between the EAL Matrix and the EAL basis document can present human factors issues which could impact timely and accurate EAL assessments.

Entergy has revised the associated Reference 1 attachments to address all of the items identified in ANO RAI 2 above and in addition, has performed two independent verifications of the attachments to ensure consistency throughout.

ANO RAI 3 The proposed EAL AU1.1 threshold values for an Unusual Event classification have substantially changed from the currently-approved EAL threshold values for ANO based on NEI 99-01, Revision 5 guidance. Considering that the guidance for AU1.1 is similar between Revision 5 and Revision 6 to NEI 99-01, the proposed changes in values do not appear to be reasonable. NRC staff could not determine a valid reason for the setpoint changes based on the information provided in the proposed EAL scheme change. Although it appeared that ANO used a similar methodology to determine RU1.1 threshold values as the Offsite Dose Calculation Manual (ODCM), the proposed values for several monitors are approximately 2 orders of magnitude lower than expected. (Note: this assumes that the General Emergency, Site Area Emergency, and Alert threshold values for ANO were properly calculated.)

The threshold values for AU1.1 are intended to address a potential reduction, as indicated by a low-level radiological release that exceeds regulatory commitments for an extended period of time. Section A.4 to Appendix A of NEI 99-01, Revision 5, discusses the usage of ODCM values as threshold values for AU1. This attachment is still applicable to NEI 99-01, Revision 6.

Please explain why the ANO ODCM calculated values for effluent flow paths were not used as a basis for the ANO AU1.1 threshold values. This explanation should include a justification for using a shutdown source term, not apportioning the threshold values to account for multiple release stacks, and Notice of Unusual Event threshold values that differ from the proposed Alert values by a factor of approximately 100 to approximately 10,000. to 0CAN091801

Page 4 of 19 Entergy Response:

ANO did not use similar methods in determining the Unusual Event (UE) effluent thresholds between the NEI 99-01, Revision 5, and Revision 6 EALs. For example, the Revision 5 thresholds were based on methods that considered allocation fractions used to maintain effluent monitor setpoints below the annual station limit. The Revision 6 thresholds do not use setpoint safety factors or allocation fractions in order to provide a value that approximates the 2x ODCM limit versus the 2x monitor set point limit. The Revision 6 method is more consistent with and supports the process of multi-unit/multi-release point dose assessment (MUDA) guidance, which was a consideration for the change in methods.

Additional differences exist between the Revision 5 and Revision 6 methods. For example, the Revision 5 methods originated from 1981 calculations updated through EAL scheme changes. The Revision 6 method uses current industry and current ODCM methods and inputs for the UE threshold. As such, several other inputs are different between the Revision 5 and Revision 6 thresholds, which further contribute to the difference in values.

The Revision 6 UE gaseous source term is based upon the NUREG-1940, Table 1-6, noble gas activity fractions available at shutdown, which would represent the maximum possible equilibrium activity fractions. This is not a "shutdown source term" where activity is lost due to decay.

There is no direct association between the UE and the Alert thresholds in the Revision 6 EALs, as there was in the Revision 5 EALs. The Revision 6 Alert thresholds are 1/100 of the General Emergency (GE) thresholds and are based on a fuel clad accident source term, accident X/Qs, and EPA-400 dose conversion factors and protective action guideline (PAG) limits developed using the URI/RASCAL dose model. The Revision 5 Alert thresholds are 200x the ODCM limit and are based on annual release noble gas source term fractions, annual average X/Q, Regulatory Guide (RG) 1.109 dose conversion factors, and 10 CFR 20 limits. Revision 5 UE and Alert thresholds were factors of each other. Revision 6 UE and Alert thresholds are not factors of each other.

ANO RAI 4 ANO removed 2RX-9840 (Post Accident Sampling Building), 2RX-9845 (Auxiliary Building Extension), and 2RX-9850 (Low Level Radwaste Storage Building) from the proposed AU1.1. These are monitored effluent paths that are not accounted for in the proposed ANO scheme, but are in use in the current approved scheme.

Please provide a justification to explain why all continuous radioactivity releases from monitored gaseous effluent pathways were not included in the proposed ANO EAL scheme change, or revise accordingly. (Note: considering that ALL effluent flow path radiation monitor readings are typically significantly lower than two times t he ODCM limit, simply providing that an effluent path had low values for some amount of time is not adequate justification for eliminating one or more monitored effluent flow paths.)

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Page 5 of 19 Entergy Response:

The Post Accident Sampling System (PASS) Building is an annex to the ANO-2 Auxiliary Building and was designed to accommodate the high activity associated with sampling the Reactor Coolant System (RCS) following an accident. The PASS is no longer used and has been abandoned in place (piping separated/capped from other plant systems). The building does not contain any other radiological systems or piping and is physically separated from Auxiliary Building ventilation systems. Therefore, the associated Super Particulate Iodine and Noble Gas (SPING) monitor (2RX-9840) associated with this annex has been abandoned. Should the PASS ever be restored, the plant modification process would identify any support systems necessary, including offsite radiological monitoring requirements and EAL impacts, for the PASS Building (see response to ANO RAI 1).

The Auxiliary Building Extension (ABE) is a clean area that was originally designed to process radioactive waste following a major Steam Generator (SG) tube rupture event assumed to grossly contaminate the secondary system. The original design, which included evaporator units, was abandoned not long after original plant startup. The lower level of the building is an open area with large compartments where low level radioactive equipment/material is sometimes stored. Other than a fire event, there is no potential of radioactive release from the ABE. Should a fire occur, it is unlikely that any release of the fixed contamination would pass through the SPING monitor. While the ventilation system is normally in service, it is not a "continuous release" from a radiological perspective. The NEI 99-01 basis for gaseous effluent thresholds specifies the release point be a normally occurring continuous radioactivity release or a planned batch release from non-continuous release pathways, neither of which applies to 2RX-9845 (SPING 10) for the ABE. Should the ABE ever be designated as a radiological area that would require consideration for offsite release potential, the plant modification process would identify any impacts to systems, operations, and regulatory requirements, including EAL impacts (see response to ANO RAI 1).

Per the ANO-2 Safety Analysis Report (SAR), Section 11.5.6.1, "Low Level Radioactive Waste Storage Building (LLRWSB)," the LLRWSB is designed to provide a controlled environment for receiving and shipping, inspection, equipment sorting, compaction, and decontamination activities associated with on-site storage and off-site shipment of LLRW.

The only design release of radioactivity would occur during compacting operations; however, this process is not used at ANO. The LLRWSB is separate from the Reactor and Auxiliary Buildings; therefore, radioactivity cannot pass through those buildings into the LLRWSB to the environment. The radioactivity contained in the LLRWSB, primarily particulate isotopes, is not sufficient to exceed the UE threshold of 2x O DCM limit for 60 minutes for any credible event.

The potential for an accident level release via the LLRWSB pathway is extremely remote.

The normal radio-isotopic makeup of the building contents mainly consists of particulate isotopes which would not be released to the atmosphere in sufficient quantities that would require a declaration of an emergency class. The most probable event for a significant release would be fire in the building which would most likely bypass the monitoring system.

Note that the LLRWSB contains a fire detection and suppression system. The NEI 99-01 basis for gaseous effluent thresholds specifies the release point be a normally occurring continuous radioactivity release or a planned batch release from non-continuous release pathways, neither of which applies to 2RX-9850 (SPING 11) for the LLRWSB.

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Page 6 of 19 ANO RAI 5 ANO calculation EP-CALC-ANO 1701 states that AU1.3 will still be valid. However, AU1.3 was not provided in the ANO EAL scheme change.

If monitors 2RX-9840, 2RX-9845 and 2RX-9850 were intended to be assessed by AU1.2 vice AU 1.3, please explain how a timely and accurate classification can be made for effluent flow paths that may not have an active discharge permit.

Entergy Response:

Entergy has revised EP-CALC-ANO-1701 to address RAI #4. The calculation no longer refers to AU1.3 for monitors 2RX-9845 and 2RX-9850. As discussed in response to RAI #4, 2RX-9840 has been abandoned. The revised calculation is provided with the revised Reference 1 enclosures, as applicable.

ANO RAI 6 The proposed EAL CU3.1, contains the condition, "-due to the loss of RCS cooling," which is not consistent with NEI 99-01, Revision 6. This could result in potential misclassification for an event that causes RCS temperature to rise above 200 degrees Fahrenheit (°F) when decay heat removal capability has not been lost.

Please provide justification, in greater detail, for adding the condition, "-due to the loss of RCS cooling," to the EAL CU3.1 threshold value, or revise accordingly.

Entergy Response:

Entergy agrees with the noted concern and, subsequently, the subject phrase has been removed from EAL CU3.1. The EAL will now state:

UNPLANNED rise in RCS temperature to > 200 °F Affected pages of the applicable Reference 1 enclosures have been updated accordingly and included in this letter.

ANO RAI 7 The proposed EAL CA1.1 threshold values equate to levels that are approximately at the bottom of the hot leg. The guidance provided by NEI 99-01, Revision 6, states "- [t]he minimum level that supports operation of normally used heat removal systems (e.g., Residual Heat Removal or Shutdown Cooling)."

Please explain what unique ANO design features support a CA1.1 threshold value that is substantially lower than the value at which the heat removal systems can operate, or revise accordingly.

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Page 7 of 19 Entergy Response:

For ANO-1, the minimum level for Decay Heat Removal (DHR) operation is elevation 370' 3", in accordance with OP-1104.004, "Decay Heat Removal Operating Procedure," Attachment B, "Minimum Height of Water to Avoid Vortex Formation vs. Decay Heat Flow." The value selected was based on 1000 gpm DHR flow which is the flow rate at which the low flow alarm is received. For consistency with NEI 99-01, EAL CA1.1 has been revised to include a value of 370.2 ft. (indication in Control Room reads only to one decimal place). While this value is slightly below the 370' 3" noted above, it ensures the EAL is not declared when flow remains above 1000 gpm. In addition, a flow rate slightly less than 1000 gpm does present an immediate challenge to the core. The RVLMS approximate equivalent value remains unchanged at Levels 1 through 8 (dry). The associated EAL Basis is also revised to support the updated RCS level value. Affected pages of the applicable Reference 1 enclosures have been updated accordingly and included in this letter.

For ANO-2, the minimum level for Shutdown Cooling operation is 24" above the bottom of the RCS hot leg or elevation 371' 11/2", in accordance with OP-2203.029, "Loss of Shutdown Cooling," Attachment A, "RCS Level." For consistency with NEI 99-01, EAL CA1.1 has been revised to include the 24" value. The RVLMS approximate equivalent value remains unchanged at Levels 1 through 5 (dry). The associated EAL Basis is also revised to support the updated RCS level values. Affected pages of the applicable Reference 1 enclosures have been updated accordingly and included in this letter.

ANO RAI 8 The proposed EAL CA3.1 contains the condition "-due to the loss of RCS [reactor coolant system] cooling," which is not consistent with NEI 99-01, Revision 6. This deviation could result in potential misclassification for an event other than a loss of RCS cooling that leads to an unplanned RCS pressure increase.

Please provide justification in greater detail for this deviation, or revise accordingly.

Entergy Response:

Entergy agrees with the noted concern and, subsequently, the subject phrase has been removed from EAL CA3.1. The EAL will now state:

UNPLANNED RCS pressure rise > 10 psig (this EAL does not apply during water-solid plant conditions)

Affected pages of the applicable Reference 1 enclosures have been updated accordingly and included in this letter.

ANO RAI 9 For proposed EALs CU5.1 and SU7.1, State and local agency communications methods include the INFORM notification system (INFORM). Based on the information provided, the NRC staff could not determine whether INFORM was independent of the telephone systems provided on Table 1[2]C-5 or if INFORM supported two-way communications. Additionally, usage of the INFORM system was not identified in the emergency plan. to 0CAN091801

Page 8 of 19 Please provide a justification for including the INFORM notification system as a State and local agency communication method. This justification should explain whether or not INFORM is independent of the provided telephone systems and if INFORM supports two-way communications.

Entergy Response:

Although INFORM is independent of the other telephone systems provided in the associated tables for these EALs, it does not support two-way communication and is, therefore, removed from Table 1[2]C-5 in EALs CU5.1 and SU7.1. Affected pages of the applicable Reference 1 enclosures have been updated accordingly and included in this letter.

ANO RAI 10 Proposed EALs FA1.1, FS1.1 and FG1.1 are assessed using threshold values that are provided by Table 1[2]F-1, "Fission Product Barrier Threshold Matrix."

a. The proposed Reactor Coolant System Barrier (RCB) 2 and Containment Barrier (CNB) 1 threshold values do not appear to be directly tied to having an RCS leak that is greater than the capacity of a charging pump, as indicated by either direct indications from control room panels or with the methodology provided by the excessive RCS leakage AOP. Using either a direct reading of RCS leakage or the AOP criteria to determine RCS leakage for both the RCS leakage AOP and the RCS barrier potential loss would facilitate timely and accurate assessment. As proposed, it appears that a mass balance must be performed to assess RCS leak rate.

Please explain how ANO can assess RCB2 and CNB1 in a timely and accurate manner, given the proposed RCB2 and CNB1 threshold value wording, or revise accordingly.

Entergy Response:

NEI 99-01 Revision 6, Pressurized Water Reactor (PWR) RCS Potential Loss 1.A (RCB2) is based on unisolable RCS or SG tube leakage requiring the start of a standby charging pump.

This threshold is based on the assumption that such a leak is in excess of normal charging pump capacity. However, not all plants have low capacity makeup pumps and placing additional pumps in service is not always the desired response to leakage. In addition, a plant may be running more than one pump to support current operations. The PWR RCS Potential Loss 1.A Developer's Notes provide for the use of alternative threshold wording, specifying an RCS leak rate (excluding normal reductions such as letdown and Reactor Coolant Pump (RCP) seal leakoff) in lieu of the starting of a standby charging pump. For ANO-1, a leak rate of 50 gpm is specified (ANO-1 does not have low capacity charging pumps). For ANO-2 a leak rate of 44 gpm is specified (capacity of a charging pump).

Per the generic bases for Containment Loss 1.A (CNB1), the condition of the faulted SG is determined consistent with either RCS Potential Loss 1.A (RCB2) or RCS Loss 1.A (RCB1).

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Page 9 of 19 Both SG leakage and RCS leakage AOPs (listed below) provide guidance for the timely and prompt estimation of RCS leak rates relative to RCB2 and CNB1.

1203.023 - Small Steam Generator Tube Leaks 1203.039 - Excess RCS Leakage 2203.016 - Excess RCS Leakage 2203.038 - Primary to Secondary Leakage

b. The proposed Unit 1 RCB3 threshold appears to be more aligned with a "typical" pressurized water reactor (PWR) vice site-specific values for ANO Unit 1.

ANO Unit 1 provides a threshold value that requires both the pressurized thermal shock limits of RT14 being applicable and an RCS pressure versus temperature that is to the left of the Nil Ductility Transition Temperature/

Low Temperature Overpressure (NDTT/LTOP) limit lines provided by EOP Figure 3. However, it appears that either of these conditions could indicate that an extreme challenge to the RCS pressure barrier exists.

For the Unit 1 RCB3 threshold value, please provide a justification for including an "AND" logic to the PTS limits of RT14 and the NDTT/ LTOP limit lines of EOP Figure 3. (Note: this justification should explain differences between RT14 and EOP Figure 3 and explain why both conditions are required as a threshold value when either condition appears to be a severe challenge to the integrity of the RCS pressure boundary.)

Entergy Response:

Generically, pressurized thermal shock (PTS) concerns occur when a combination of two conditions exists:

1. Excessive RCS cooldown AND 2. The RCS is at a pressure and temperature on the wrong side of the Pressure-Temperature (P-T or PTS) curve (NDTT limit).

The combination of these two conditions can challenge the integrity of the RCS boundary and thus warrant a Potential Loss of the RCS due to potentially high thermal stresses induced by the excessive RCS cooldown combined with excessive RCS pressure. In accordance with the Technical Specification (TS) Basis associated with P-T curves, the limits define allowable operating regions and permit a large number of operating cycles while providing a wide margin to non-ductile failure. Therefore, operation beyond the acceptable regions defined on the curves is unlikely to pose an immediate threat to the integrity of the reactor coolant pressure boundary (RCPB). In fact, the TSs require an engineering evaluation of the transient impact on the RCPB, and may or may not require placing the unit in Cold Shutdown. One of the major concerns is the temperature gradient that may exist across the vessel or piping walls. An excessive RCS cooldown is required to present a temperature gradient that would be of significant concern. However, without coincident excessive pressure, the temperature gradient is not likely to pose an immediate threat to the RCPB. to 0CAN091801

Page 10 of 19 If the "AND" condition were changed to an either/or, the potential for unnecessary Alert or higher classifications is increased. Because both a rapid cooldown and high pressure is required to pose a significant challenge to the RCPB, and because the P-T curves are established with significant margin, an Alert or higher classification is likely unwarranted if only one of the two conditions exists. Therefore, Entergy believes the "AND" condition is appropriate for this EAL. This is consistent with plant EOP bases that PTS concerns only apply when both the specified cooldown rate limits and NDTT/LTOP curve limits are

exceeded.

For ANO-1, the first condition is met when, per Routine Task 14 (RT14), PTS limits apply.

This results from any one of the following three conditions existing:

RCS cooldown rate > 100 °F with Tcold < 355 °F RCS cooldown rate > 50 °F with Tcold < 300 °F High Pressure Injection (HPI) in service with all RCPs off For ANO-1, the second condition is met when RCS pressure and temperature are left of the NDTT/LTOP limit lines on EOP Figure 3.

Therefore, the appropriate logic wording for ANO-1 RCS Potential Loss threshold RCB3 is "AND." c. The proposed Unit 2 RCB3 threshold appears to be more aligned with a "typical" PWR vice site-specific values for ANO Unit 2. ANO Unit 2 provides a threshold value that requires both an uncontrolled RCS cooldown and an RCS pressure/temperature that is to the left of the pressure-temperature (P-T) limit lines provided by Standard Attachment 1. It appears that operating the unit in the region to the left of the P-T limit lines provided by Standard , by itself, could indicate t hat an extreme challenge to the RCS pressure barrier exists.

For the unit 2 RCB3 threshold value, please provide a justification for including "AND" logic to an uncontrolled RCS cooldown and the P-T limits of Standard Attachment 1.

Entergy Response:

Generically, PTS concerns exist when a combination of two conditions exists:

1. Excessive RCS cooldown.

AND 2. The RCS is at a pressure and temperature on the wrong side of the P-T (PTS) curve (NDTT limit).

The combination of these two conditions can challenge the integrity of the RCS boundary and thus warrant a Potential Loss of the RCS due to potentially high thermal stresses induced by the excessive RCS cooldown combined with excessive RCS pressure (see response to

Part b above). to 0CAN091801

Page 11 of 19 For ANO-2, the first condition is met if there is an uncontrolled RCS cooldown (currently stated as a 50 °F step change below 500 °F from normal operating temperature (NOT).

For ANO-2, the second condition is met when RCS pressure and temperature are left of Line B (200 °F margin-to-saturation) on Standard Attachment 1, P-T limits.

Therefore, the appropriate logic wording for ANO-2 RCS Potential Loss threshold RCB3 is "AND." In addition to the above, the 1 st bullet of the current ANO-2 EAL contains the following parenthetical:

1. Uncontrolled RCS cooldown (50°F step change which is below 500°F from NOT)

Operations has determined that the wording of the parenthetical statement is unclear and, therefore, proposes to revise the wording consistent with the RCS cooldown TS limits:

1. Uncontrolled RCS cooldown (> 50 °F step change or > 100 °F change in less than a one-hour period)

In accordance with TS 3.4.9.1, "Pressure/Temperature Limits," the cooldown limits are applicable when RCS temperature is between 50 °F and 560 °F; therefore, the current reference to 500 °F is irrelevant and removed. The aforementioned revision provides clarity and meets the intent of the EAL bullet. Affected pages of the applicable Reference 1 enclosures have been updated accordingly and included in this letter.

ANO RAI 11 The Fuel Clad Barrier (FCB) 4, RCB4, and EAL SS6.1 threshold values include the condition "HPI [High Pressure Injection] [Once Through] cooling initiated." For some PWRs, implementation of procedural guidance would provide cooling by injecting water into the RCS and removing that water, such that core cooling is established. The proposed wording implies that not only are the steam generators ineffective for heat removal, but that an alternate heat removal path has been established. This is not consistent with the guidance in NEI 99-01, Revision 6, which provides a threshold value of "[i]nadequate heat removal capability via steam generators as indicated by (site-specific indications)."

Please explain how a timely and accurate assessment can be performed for FCB4, RCB4, and SS6.1 with the proposed condition requiring HPI [Once Through] cooling initiation, rather than the HPI [Once Through] cooling procedure implementation, or revise accordingly.

Entergy Response:

To resolve the subject concern, Entergy proposes to revise the subject condition to state the following:

An on-shift SRO has determined that the procedure conditions are met to commence initiation of HPI[Once Through] cooling. to 0CAN091801

Page 12 of 19 The revised wording will ensure prompt declaration of the appropriate EALwhen conditions for HPI have been met, rather than waiting until HPI is actually placed in service. Entergy believes the revised wording meets the intent of NEI 99-01, Revision 6, as described in the RAI. Affected pages of the applicable Reference 1 enclosures have been updated accordingly and included in this letter.

ANO RAI 12 The threshold value for FCB5 is based on 300 microcuries/gram dose equivalent I-131 which typically corresponds to 2 percent to 5 percent fuel cladding damage. For ANO, this corresponds to fuel cladding damage of 1.49 percent for Unit 1 and 1.13 percent for Unit 2.

Please explain why the proposed FCB5 EAL radiation monitor threshold values do not correspond to 2 percent to 5 percent of cladding damage.

Entergy Response:

The NEI 99-01, Revision 6, developer notes direct that "the reading should be determined assuming the instantaneous release and dispersal of the reactor coolant noble gas and iodine inventory, with RCS radioactivity concentration equal to 300 Ci/gm dose equivalent I-131, into the containment atmosphere." The associated NEI basis states in part "reactor coolant activity above this level is greater than that expected for iodine spikes and corresponds to an approximate range of 2% to 5% fuel clad damage."

The 2% - 5% rule of thumb used in the NEI basis for PWRs and BWRs originated in NUMARC/NESP-007 Rev 2 and could be confirm ed using typical enrichment values and NUREG-1228 related source term/partitioning assumptions. Currently, most units operate with higher enrichment than what was common in the late 1980s and have been approved for power uprates. Additionally, newer source term and partitioning guidance such as NUREG-1940 are a factor. These combine such that typical reactor coolant concentrations to percent clad damage are approximately half of what was calculated using historical inputs and guidance. In addition, the use of a 300 uCi/gm dose equivalent I-131 (DEI) source term for FCB5 (NEI Fuel Clad Barrier Loss threshold 3.A) provides agreement within the EAL scheme with FCB6 (NEI Fuel Clad Barrier Loss threshold 3.B) which directly refers to a 300 uCi/gm DEI value.

ANO RAI 13 Please justify using a containment hydrogen concentration of greater than 3 percent for the proposed CNB7 threshold value, as this is not consistent with the explosive mixture provided by NEI 99-01, Revision 6. (Note: the proposed threshold value of containment hydrogen concentration of greater than 3 percent could result in an early or unwarranted General Emergency declaration.)

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Page 13 of 19 Entergy Response:

The containment hydrogen concentration in CNB7 is changed to "> 4%" for consistency with

NEI 99-01, Revision 6. The statement "the 4% hydrogen concentration is generally considered the lower limit for hydrogen deflagrations" has been added to the basis information. Affected pages of the applicable Reference 1 enclosures have been updated accordingly and included in this letter.

ANO RAI 14 The proposed HA1.1 and HS1.1 definitions of the owner controlled area (OCA) and the protected area (PA) appear to be the same. The definition of the OCA indicates that it is

demarcated by a vehicle barrier system and a security fence with access controlled by an access control point. The PA is "[a]n area encompassed by physical barriers (i.e., the security fence) and to which access is controlled." As such, it appears that HA1.1 and HS1.1 have

similar threshold value criteria that could cause a delayed or inaccurate EAL classification.

Please explain how ANO can perform a timely and accurate assessment of HA1.1 and HS1.1 with the proposed definitions, or revise accordingly.

Entergy Response:

The definition of OCA provided in the subject EALs states that this area is considered the

Security Owner Controlled Area (SOCA) at ANO. The SOCA is basically designated as the outside fence around the facility, while the PA is the inner fence. To enhance clarification of what these areas represent, the current OCA and Protected Area definitions contained in EALs HA1.1 and HS1.1 are revised as follows (deletions are struck through, additions are underlined):

SECURITY OWNER CONTROLLED AREA (SOCA) - For the purposes of classification this is the Security Owner Controlled Area (SOCA). The SOCA is the area demarcated as a Vehicle Barrier System (VBS) consisting of passive elements including a series of large concrete blocks on the inside ofand a detection fence on the outside and a delay fence with early warning capabilitieson the inside of the passive and active barriers. The SOCA is the area betweeninside the SOCA Fence andVBS up to the PROTECTED AREA Boundaryfence line. Access to this area is controlled by the SOCA Personnel Access Control Point.

PROTECTED AREA - An area clearly demarcated by a fence or building wall with an entrance portal that is regulated by Secu rity Personnelencompassed by physical barriers (i.e., the security fence) and to which access is controlled access.

The above definitions are consistent with those contained in security procedure EN-NS-232, "General Employee Security Responsibilities." The revision of the above terms will prevent confusion or misinterpretation by maintaining consistent definitions of these areas. Affected pages of the applicable Reference 1 enclosures have been updated accordingly and included in this letter.

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Page 14 of 19 ANO RAI 15 The proposed EALs HU4.1 and HU4.2 appear to cover a wider range of areas than that provided by NEI 99-01, Revision 6 (EAL HU4). Please provide justification that all areas identified for this EAL contain equipment needed for safe operation, safe shutdown or safe cool-down, or revise as necessary to support accurate and timely assessment.

Entergy Response:

For ANO-1, there are approximately 40 rooms/areas within the Auxiliary Building associated with fire safe shutdown. For ANO-2, there are approximately 50 rooms/areas within the Auxiliary Building associated with fire safe shutdown. The large numbers of rooms/areas is

largely due to cable runs traversing many different rooms/areas throughout the Auxiliary Building. Even if areas associated only with cable runs are neglected, there would remain 25 ANO-1 and 35 ANO-2 areas in the Auxiliary Building that would need to be listed in the associated EALs. Based on this large number of areas that can potentially affect fire safe shutdown equipment, Entergy proposes to maintain the generic listing in Tables 1H-1 and 2H-1 of "Auxiliary Building" for both units, including the exceptions listed in Table 1H-1.

However, the reference to the Auxiliary Building Extension in Table 2H-2 is being established as a separate table entry and the specific area associated with fire safe shutdown in this building is added. This eliminates a very large area of the Auxiliary Building Extension from being of concern.

The generic "Reactor Building" entry in both tables is also being maintained, again largely

due to the various areas where safety related equipment and cables may traverse. Further discussion of the Reactor Building in relation to station fires is included in response to RAI 16 below. With respect to the Turbine Buildings, both tables are being modified to significantly reduce the areas of concern. For both units, this includes the path in which electrical cabling from offsite power and the Alternate AC Diesel Generator (also referred to as the station blackout diesel) enter the plant, connect with non-vital switchgear, and continue to the vital switchgear located in the Auxiliary Buildings.

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Page 15 of 19 The following changes are proposed:

Table 1H-1 Unit 1 Fire Areas Reactor Building All elevations Auxiliary Building All elevations including: Penthouse/MSIV Room Exceptions: Boric Acid Mix Tank Room (Chem Add Area), 404' (157-B), EDG Exhaust Fan area on 386' (1-E and 2-E)

Turbine Building All elevations on the west side of Turbine Building and including: Pipechase under ICW Coolers, CRD Pump Pit/T-28 Room/Area under ICW Pumps 372' Elevation from non-vital switchgear area to Auxiliary Building wall at DR 56 Outside Areas Manholes adjacent to Startup #2 XFMR (MH-03/MH-04) Manholes adjacent to Intake Structure (MH-05/MH-06) Intake Structure (354' and 366')

Diesel Fuel Vault Diesel Fuel Vault Pump Manholes (MH-09 and MH-10)

Table 2H-1 Unit 2 Fire Areas Reactor Building All elevations Auxiliary Building All elevations including: MG Set Room, UNEPR, LNEPR, 2B-53 Room Auxiliary Building Extension MSIV Room Turbine Building All elevations on the west side of Turbine Building and 372' Elevation from non-vital switchgear area to Auxiliary Building wall at DR 340 Outside Areas Intake Structure (354' and 366') Concrete Manhole East, NE of intake (2MH-01)

Concrete Manhole East of Turbine Building next to train bay (2MH-03) Diesel Fuel Vault Diesel Fuel Vault Pump Manholes (MH-09 and MH-10) to 0CAN091801

Page 16 of 19 The proposed changes to Tables 1H-1 and 2H-1 result in a significant reduction in the number of areas of concern. This subsequently significantly reduces the potential of unnecessary EAL declarations associated with HU4.1 and HU4.2. Affected pages of the applicable Reference 1 enclosures have been updated accordingly and included in this letter.

ANO RAI 16 The proposed EAL HU4.2 - Tables 1H-1, "Unit 1 Fire Areas," and 2H-1, "Unit 2 Fire Areas," include all elevations of the Reactor Building. This could result in an event declaration due to the spurious actuation of a single fire alarm. Based on the information provided in the license amendment request, the NRC staff could not determine if the containment fire detection system at ANO, in combination with the ANO containment ventilation system, supported the inclusion of the Reactor Building as a fire area for EAL HU4.2.

Please provide sufficient justification that demonstrates why, or why not, including the Reactor Building Table 1[2]H-1 could not result in unnecessary event declarations. If this justification demonstrates that including the Reactor Building is not appropriate as a fire area for HU4.2, please modify accordingly.

Entergy Response:

ANO EAL HU4.1 addresses the condition where a fire is reported and verified in a listed Fire Area. This verification could be from a report in the field or because multiple fire detection device alarms are received. This EAL includes a table that lists fire areas of concern, including the Reactor Building.

ANO EAL HU4.2 addresses receipt of a single fire detector without a corresponding verification. Entergy makes an exception in EAL HU4.2 to exclude the Reactor Building in Modes 1 and 2. Personnel safety concerns preclude entry into certain areas of the Reactor Building during these modes. In addition, there are areas within the Reactor Building where fire detectors are located that would be inaccessible during these modes due to elevated radiation levels. Industry experience has demonstrated that including the Reactor Building in

Modes 1 and 2 in EAL HU4.2 can lead to unus ual event emergency classifications based on a single spurious fire alarm, requiring subsequent emergency declaration retractions.

With respect to Reactor Building fire alarms, it can reasonably be expected that a fire that burns for 15 minutes would produce sufficient products of combustion to cause multiple fire detection devices to alarm. This is due to the products of combustion being transported to other areas inside the Reactor Building due to the forced flow ventilation system in operation.

Likewise, receipt of a single fire alarm would likely be due to a spurious detector actuation.

There are four safety related Reactor Building Fan Cooling units located in each ANO Reactor Building. At least 3 of the 4 fans in each unit's Reactor Building are normally in operation in Modes 1 and 2. For ANO-1, each fan unit delivers an air flow of approximately 30,000 cfm. For ANO-2, each fan unit delivers an air flow of approximately 27,000 cfm.

The fan units draw return air from the Reactor Building atmosphere and discharge into a common header which delivers cooled ventilating air to multiple areas inside the Reactor Building. This constant flow of air would draw any smoke towards the cooling units past the to 0CAN091801

Page 17 of 19 installed detectors, thus initiating multiple smoke detector alarms. Actuation of more than one smoke detector is the most reliable indication of an actual fire because of the high volumetric air flow throughout the Reactor Building. Due to construction of the intermediate floors and multiple openings in the floors, it can be expected that smoke would migrate throughout the Reactor Building in a very short period and that 2 or more smoke detectors would alarm. Entergy considers basing emergency classifications on receiving more than one smoke detector actuation as the most reliable indication of a valid alarm and accurately meets the Initiating Condition of HU4, "FIRE potentially degrading the level of safety of the

plant." With consideration to the above discussion, Note 14 is added to EAL HU4.2 as follows:

During Modes 1 and 2, HU4.2 is not applicable to a single fire alarm in the Reactor Building.

The following information is added to the basis for HU4.2:

This EAL is not applicable for the Reactor Building in Modes 1 and 2. The Reactor Building air flow design and TS requirements for operation of Reactor Building Fan

Coolers are such that multiple smoke detectors would be expected to alarm for a fire in the Reactor Building. A fire in the Reactor Building in these modes would therefore be classified under EAL HU4.1.

Verification of a single Reactor Building fire alarm that is likely to be spurious does not warrant the potential elevated exposure risks and industrial safety risks associated with an emergency entry of the Reactor Building in Modes 1 and 2. Therefore, Entergy proposes the aforementioned revision to EAL HU4.2 such that the EAL is only applicable to a single fire alarm in the Reactor Building in Modes 3, 4, 5 and 6.

The structure of the HU4 IC/EAL is modelled after Seabrook Station's adoption of NEI 99-01, Revision 6, EALs containing a similar exception, which was approved by the NRC in Amendment 152 to the Seabrook Station Facility, Operating License No. NPF-86, on February 10, 2017 (ML16358A411).

Based on the information above, Entergy considers the proposed revision to be an acceptable deviation from the generic NEI 99-01, Revision 6, guidance. This deviation is consistent with proposed Emergency Plan (EP) Frequently Asked Question (FAQ) 2018-03 (ML18081A309). Affected pages of the applicable Reference 1 enclosures have been updated accordingly and included in this letter.

Entergy believes the preceding information is in accordance with discussions held between the NRC and the licensee on August 1, 2018.

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Page 18 of 19 In addition to the RAI responses above, Entergy proposes two additional changes which were identified during the response preparation associated with ANO RAI 16 above.

1. To avoid unnecessary declarations of EAL HU4.1, the following Note 13 is proposed to be added to HU4.1:

Note 13: Bullet 2 of this EAL (multiple fire alarm indications) is not applicable for LOCAs or MSL breaks in containment.

Steam releases are known to trigger fire detectors in the vicinity of the leak. Because other EALs address LOCAs and MSL breaks, Entergy believes declaring HU4.1 during such an event would suggest to non-Entergy agencies (local/state government, NRC, etc.) that a fire is also occurring at the site. While it may be possible to experience a fire inside containment coincident with a LOCA or MSL break, such is unlikely, especially given the steam/water atmosphere created by the LOCA / MSL break. Therefore, Entergy proposes to add the aforementioned note and the following language to the HU4.1 basis that would eliminate this potential communication with offsite agencies when multiple fire detectors actuate inside containment coincident with a LOCA or MSL break inside containment.

Because steam release due to a LOCA or MSL break inside containment can result in invalid fire detector actuations in containment, Note 13 eliminates the potential for EAL HU4.1 declaration due to such invalid alarms. This is reasonable based on the low probability of a fire occurring inside containment coincident with a LOCA or MSL break inside containment, and due to the low probability of a significant fire existing in a steam/water atmospheric environment.

2. With respect to EAL HU4.2, one of the listed criteria is associated with a fire that is NOT verified to exist within 30 minutes. The wording implies that a fire must be verified to exist within 30 minutes or the EAL is applicable. However, the intent of this criterion is to address those scenarios where a fire has not been proved or disproved within 30 minutes. While this intent is discussed in the associated basis, Entergy requests a clarification be made to this respective criterion in the EAL proper in order to avoid delays in determining if the EAL is applicable to a given event:

The existence of a FIRE is not verified (i.e., proved or disproved) within 30 min. of alarm receipt (Note 1)

The Entergy response requires changes to certain enclosures included in the original Reference 1 application. For completeness and ease of NRC review, the following Reference 1 enclosures are being resubmitted in full as part of Entergy's response to the NRC's RAI:

2. Proposed EAL Technical Basis Document (Markup) 3. Proposed EAL Technical Basis Document (Clean)
4. NEI 99-01, Rev. 6, Deviations and Differences, ANO Units 1 and 2
5. Proposed EAL Matrix Chart and Review Table (for information only) 6. Supporting Referenced Document Pages to 0CAN091801

Page 19 of 19

REFERENCES:

1. Entergy letter dated March 29, 2018, License Amendment Request - Adoption of Emergency Action Level Schemes Pursuant to NEI 99-01, Revision 6 , Arkansas Nuclear One, Units 1 and 2, 0CAN031801 (ML18088B412)
2. NRC email dated August 2, 2018, ANO-1 and 2 Final RAI RE: License Amendment Request to Adopt EAL Scheme Change Per NEI 99-01 Revision 6, (EPID L-2018-LLA-0082) (0CNA081801) (ML18218A221)

Enclosure 2 to 0CAN091801 Proposed EAL Technical Basis Document (Markup) to 0CAN091801

Page 1 of 257 Table of Contents Section Page

1.0 INTRODUCTION

.................................................................................................................2

2.0 DISCUSSION

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

......2 2.1 Background ................................................................................................................

2 2.2 Fission Product Barriers .............................................................................................3 2.3 Fission Product Barrier Classification Criteria ............................................................3 2.4 EAL Organization .......................................................................................................4

2.5 Technical

Basis Information .......................................................................................5

2.6 Operations

Mode Applicability

....................................................................................7

3.0 GUIDANCE

ON MAKING EMERGENCY CLASSIFICATIONS ..........................................8 3.1 General Considerations .............................................................................................8

3.2 Classification

Methodology ......................................................................................10

4.0 REFERENCES

..................................................................................................................13 4.1 Developmental .........................................................................................................13 4.2 Implementing............................................................................................................13

5.0 DEFINITIONS, ACRONYMS & ABBREVIATIONS ...........................................................13 5.1 Definitions (ref. 4.1.1 except as noted) ....................................................................13 5.2 Abbreviations/Acronyms ..........................................................................................18 6.0 ANO-TO-NEI 99-01 Rev. 6 EAL CROSS-REFERENCE ..................................................22

7.0 ATTACHMENTS

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

24 7.1 Attachment 1, Emergency Action Level Technical Bases ........................................24 Category A - Abnormal Rad Levels / Rad Effluents ................................................25 Category C - Cold Shutdown / Refueling System Malfunction ................................70 Category E - Independent Spent Fuel Storage Installation (ISFSI) .......................109 Category F - Fission Product Barrier Degradation ................................................112 Table 1[2]F-1, Fission Product Barrier Threshold Matrix & Bases ...119 Category H - Hazards and Other Conditions Affecting Plant Safety .....................164 Category S - System Malfunction ..........................................................................212 7.2 Attachment 2, Safe Operation & Shutdown Areas Tables 1[2]A-3 & 1[2]H-2 Bases .....................................................255 to 0CAN091801

Page 2 of 257

1.0 INTRODUCTION

This document provides an explanation and rationale for each Emergency Action Level (EAL) included in the EAL Upgrade Project for Arkansas Nuclear One (ANO). It should be used to facilitate review of the ANO EALs and provide historical documentation for future reference. Decision-makers responsible for implementation of 1903.010, Emergency Action Level Classification, may use this document as a technical reference in support of EAL interpretation.

This information may assist the Emergency Director in making classifications, particularly those involving judgment or multiple events. The basis information may also be useful in training and for explaining event classifications to off-site officials.

The expectation is that emergency classifications are to be made as soon as conditions are present and recognizable for the classification, but within 15 minutes or less in all cases when conditions are present and have been recognized. Use of this document for assistance is not intended to delay the emergency classification.

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

10 CFR 50.54(q).

2.0 DISCUSSION

2.1 Background

EALs are the plant-specific indications, conditions or instrument readings that are utilized to classify emergency conditions defined in the ANO Emergency Plan.

In 1992, the NRC endorsed NUMARC/NESP-007 "Methodology for Development of Emergency Action Levels" as an alternative to NUREG-0654 EAL guidance.

NEI 99-01 (NUMARC/NESP-007) Revisions 4 and 5 were subsequently issued for industry implementation. Enhancements over earlier revisions included:

Consolidating the system malfunction initiating conditions and example emergency action levels which address conditions that may be postulated to occur during plant shutdown conditions. Initiating conditions and example emergency action levels that fully address conditions that may be postulated to occur at permanently Defueled Stations and Independent Spent Fuel Storage Installations (ISFSIs). Simplifying the fission product barrier EAL threshold for a Site Area Emergency.

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 t he Development of Emergency Action Levels for Non-Passive Reactors," November 2012 (ref. 4.1.1), ANO conducted an EAL implementation upgrade project that produced the EALs discussed herein.

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Page 3 of 257

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 a greater probability of barrier loss and reduced certainty of maintaining the barrier.

The primary fission product barriers are:

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

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

C. Containment Barrier (CNB): The Containment Barrier includes the Reactor Building and connections up to and including the outermost containment isolation valves. This barrier also includes the main steam, feedwater, and blowdown line extensions outside the Reactor Building up to and including the outermost secondary side isolation valve.

Containment Barrier thresholds are used as criteria for escalation of the Emergency Classification Level (ECL) from an 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

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Page 4 of 257 2.4 EAL Organization

The ANO EAL scheme includes the following features:

Division of the EAL set into three broad groups:

o EALs applicable under any plant operating modes - This group would be reviewed by the EAL-user any time emergency classification is considered.

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

o EALs applicable only under cold operating modes - This group would only be reviewed by the EAL-user when the plant is in Cold Shutdown, Refueling or Defueled mode. The purpose of the groups is to avoid review of hot condition EALs when the plant is in a cold condition and avoid review of cold condition EALs when the plant is in a hot condition. This approach significantly minimizes the total number of EALs that must be reviewed by the EAL-user for a given plant condition, reduces EAL-user reading burden and, thereby, speeds identification of the EAL that applies to the emergency.

Within each group, assignment of EALs to categories and subcategories:

Category and subcategory titles are selected to represent conditions that are operationally significant to the EAL-user. The ANO EAL categories are aligned to and represent the NEI 99-01, "Recognition Categories." Subcategories are used in the ANO scheme as necessary to further divide the EALs of a category into logical sets of possible emergency classification thresholds. The ANO EAL categories and subcategories are listed below.

The primary tool for determining the emergency classification level is the EAL Classification Matrix. The user of the EAL Classification Matrix may (but is not required to) consult the EAL technical bases in order to obtain additional information concerning the EALs under classification consideration. The user should consult Section 3.0 and Attachment 1 of this document for such information.

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Page 5 of 257 EAL Groups, Categories and Subcategories EAL Group/Category EAL Subcategory Any Operating Mode:

A - Abnormal Rad Levels / Rad Effluent 1 - Radiological Effluent 2 - Irradiated Fuel Event 3 - Area Radiation Levels H - Hazards and Other Conditions Affecting Plant Safety 1 - Security 2 - Seismic Event 3 - Natural or Technological Hazard 4 - Fire 5 - Hazardous Gas 6 - Control Room Evacuation 7 - Emergency Director Judgment E - Independent Spent Fuel Storage Installation (ISFSI) 1 - Confinement Boundary Hot Conditions:

S - System Malfunction 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 - Containment Failure 9 - Hazardous Event Affecting Safety Systems F - Fission Product Barrier Degradation None Cold Conditions:

C - Cold Shutdown / Refueling System Malfunction 1 - RCS Level 2 - Loss of Essential AC Power 3 - RCS Temperature 4 - Loss of Vital DC Power 5 - Loss of Communications 6 - Hazardous Event Affecting Safety Systems

2.5 Technical

Bases Information

EAL technical bases are provided in Attachment 1 for each EAL according to EAL group (Any, Hot, Cold), EAL category (A, C, E, F, H and S) 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:

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Page 6 of 257 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 (A, C, E, F, H or S) 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. If an ANO Unit 2 EAL threshold value differs from Unit 1, the Unit 2 threshold is enclosed in brackets. For example, in the EAL threshold "RVLMS Levels 1 through 8 indicate DRY [RVLMS Levels 1 through 5 indicate DRY]", "RVLMS Levels 1 through 5 indicate DRY" apply only to Unit 2.

Mode Applicability One or more of the following plant operating conditions comprise the mode to which each EAL is applicable: 1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown, 5 - Cold Shutdown, 6 - Refueling, DEF - Defueled, or All. (See Section 2.6 for operating mode definitions).

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Page 7 of 257 Definitions:

If the EAL wording contains a defined term, the definition of the term is included in this section. These definitions can also be found in Section 5.1.

Basis: An EAL basis section that provides ANO-relevant information concerning the EAL as well as a description of the rationale for the EAL as provided in NEI 99-01 Rev. 6.

Reference(s):

Source documentation from which the EAL is derived.

2.6 Operating

Mode Applicability

Unit 1 (ref. 4.1.6):

1 Power Operation K eff 0.99, reactor power > 5%

2 Startup K eff 0.99, reactor power 5% 3 Hot Standby K eff < 0.99, reactor coolant temperature 280°F 4 Hot Shutdown K eff < 0.99, reactor coolant temperature 280°F > Tavg > 200°F and all reactor vessel head closure bolts fully tensioned 5 Cold Shutdown K eff < 0.99, reactor coolant temperature 200°F and all reactor vessel head closure bolts fully tensioned 6 Refueling One or more reactor vessel head closure bolts less than fully tensioned DEF Defueled All fuel assemblies have been removed from Containment and placed in the spent fuel pool.

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Page 8 of 257 Unit 2 (ref. 4.1.6):

1 Power Operation K eff 0.99, reactor power > 5%, average coolant temperature 300°F 2 Startup K eff 0.99, reactor power 5%, average coolant temperature 300°F 3 Hot Standby K eff < 0.99, average coolant temperature 300°F 4 Hot Shutdown K eff < 0.99, average coolant temperature 300°F > Tavg > 200°F 5 Cold Shutdown K eff < 0.99, average coolant temperature 200°F 6 Refueling K eff 0.95, average coolant temperature 140°F, reactor vessel head unbolted or removed, and fuel in the vessel.

DEF Defueled All fuel assemblies have been removed from Containment and placed in the spent fuel pool.

The plant operating mode that exists at the time t hat 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.

3.0 GUIDANCE

ON MAKING EMERGENCY CLASSIFICATIONS

3.1 General

Considerations

When making an emergency classification, the Emergency Director 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.

EAL matrices should be read from left to right, from General Emergency to Unusual Event, and top to bottom. Declaration decisions should be independently verified before declaration is made except when gaining this verification would exceed the 15 minute declaration requirement. Place keeping should be used on all EAL matrices.

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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 provi ded guidance on implementing this requirement in NSIR/DPR-ISG-01, "Interim Staff Guidance, Emergency Planning for Nuclear Power Plants" (ref. 4.1.8).

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 the indicator's operability, the condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment.

3.1.3 Imminent

Conditions

For ICs and EALs that have a stipulated time duration (e.g., 15 minutes, 30 minutes, etc.), the Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the condition has exceeded, or will likely exceed, the applicable time. If an ongoing radiological release is detected and the release start time is unknown, it should be assumed that the release duration specified in the IC/EAL has been exceeded, absent data to the contrary.

3.1.4 Planned

vs. Unplanned Events

A planned work activity that results in an expected event or condition which meets or exceeds an EAL does not warrant an emergency declaration provided that: 1) the activity proceeds as planned, and 2) the plant remains within the limits imposed by the operating license. Such activities include planned work to test, manipulate, repair, maintain or modify a system or component. In these cases, the controls associated with the planning, preparation and execution of the work will ensure that compliance is maintained with all aspects of the operating license provided that the activity proceeds and concludes as expected. Events or conditions of this type may be subject to the reporting requirements of 10 CFR 50.72 (ref. 4.1.4).

3.1.5 Classification

Based on Analysis

The assessment of some EALs is based on the results of analyses that are necessary to ascertain whether a specific EAL threshold has been exceeded (e.g., dose assessments, chemistry sampling, RCS leak rate calculation, etc.). For these EALs, the EAL wording or the to 0CAN091801

Page 10 of 257 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 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 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 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. 4.1.8).

3.2.1 Classification

of Multiple Events and Conditions

When multiple emergency events or conditions are present, the user will identify all met or exceeded EALs. The highest applicable ECL identified during this review is declared. For example: If an Alert EAL and a Site Area Emergency EAL are met, whether at one unit or at two units, a Site Area Emergency should be declared.

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

If two Alert EALs are met, whether at one unit or at two units, an Alert should be declared.

Related guidance concerning classification of rapidly escalating events or conditions is provided in Regulatory Issue Summary (RIS) 2007-02, Clarification of NRC Guidance for Emergency Notifications During Quickly Changing Events (ref. 4.1.2).

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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 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, 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 Termination An ECL may be terminated when the event or condition that meets the classified IC and EAL no longer exists, and other site-specific termination requirements are met.

As noted above, guidance concerning classification of rapidly escalating events or conditions is provided in RIS 2007-02 (ref. 4.1.2).

3.2.5 Classification

of Short-Lived Events Event-based ICs and EALs define a variety of specific occurrences that have potential or actual safety significance. By their nature, some of these events may be short-lived and, thus, over before the emergency classification assessment can be completed. If an event occurs that meets or exceeds an EAL, the associated ECL must be declared regardless of its continued presence at the time of declaration. Examples of such events include an earthquake or a failure of the reactor protection system to automatically trip the reactor followed by a successful manual trip.

3.2.6 Classification

of Transient Conditions Many of the ICs and/or EALs employ time-based criteria. These criteria will require that the IC/EAL conditions be present for a defined period of time before an emergency declaration is warranted. In cases where no time-based criterion is specified, it is recognized that some transient conditions may cause an EAL to be met for a brief period of time (e.g., a few seconds to a few minutes). The following guidance should be applied to the classification of these conditions. to 0CAN091801

Page 12 of 257 EAL momentarily met during expected plant response - In instances in which 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. 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 completing the review and steps necessary to make the emergency declaration. This provision is included to ensure that any public protective actions resulting from the emergency classification are truly warranted by the plant conditions.

3.2.7 After-the-Fact Discovery of an Emergency Event or Condition

In some cases, an EAL may be met but the emergency classification was not made at the time of the event or condition. This situation can occur when personnel discover that an event or condition existed which met an EAL, but no emergency was declared, and the event or condition no longer exists at the time of discovery. This may be due to the event or condition not being recognized at the time or an error that was made in the emergency classification process. In these cases, no emergency declaration is warranted; however, the guidance contained in NUREG-1022 (ref. 4.1.3) is applicable. Specifically, the event should be reported to the NRC in accordance with 10 CFR § 50.72 (ref. 4.1.4) within one hour of the discovery of the undeclared event or condition. The licensee should also notify appropriate State and local agencies in accordance with the agreed upon arrangements.

3.2.8 Retraction

of an Emergency Declaration

Guidance on the retraction of an emergency declaration reported to the NRC is discussed in NUREG-1022 (ref. 4.1.3).

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4.0 REFERENCES

4.1 Developmental

4.1.1 NEI 99-01 Revision 6, Methodology for the Development of Emergency Action Levels for Non-Passive Reactors, ADAMS Accession Number ML12326A805 4.1.2 RIS 2007-02 Clarification of NRC Guidance for Emergency Notifications During Quickly Changing Events, February 2, 2007. 4.1.3 NUREG-1022 Event Reporting Guidelines: 10CFR50.72 and 50.73 4.1.4 10 § CFR 50.72 Immediate Notification Requirements for Operating Nuclear Power Reactors 4.1.5 10 § CFR 50.73 License Event Report System 4.1.6 Unit 1[2] Technical Specifications Table 1.1-1[1.1], Modes[Operational Modes]

4.1.7 Arkansas

Nuclear One Offsite Dose Calculation Manual (ODCM) 4.1.8 NSIR/DPR-ISG-01 Interim Staff Guidance, Emergency Planning for Nuclear Power Plants 4.1.9 Arkansas Nuclear One Emergency Plan 4.1.10 1015.008 Unit 2 SDC Control

4.2 Implementing

4.2.1 1903.010 Emergency Action Level Classification 4.2.2 NEI 99-01 Rev. 6 to ANO EAL Comparison Matrix 4.2.3 ANO EAL Matrix 5.0 DEFINITIONS, ACRONYMS & ABBREVIATIONS

5.1 Definitions

(ref. 4.1.1 except as noted)

Selected terms used in Initiating Condition, Emergency Action Level statements and EAL bases are set in all capital letters (e.g., ALL CAPS). These words are defined terms that have specific meanings as used in this document. The definitions of these terms are provided below.

Alert Events are in progress, or have occurred, which involve an actual or potential substantial degradation of the level of safety of the plant or a security event that involves probable life threatening risk to site personnel or damage to site equipment because of HOSTILE ACTION.

Any releases are expected to be limited to small fractions of the EPA Protective Action Guideline exposure levels.

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Page 14 of 257 Confinement Boundary The barrier(s) between spent fuel and the environment once the spent fuel is processed for dry

storage. As related to the ANO ISFSI, the Confinement Boundary is comprised of either the Multi-assembly Sealed Basket (MSB) (SNC System) or Multi-Purpose Canister (MPC) (Holtec System). Containment Closure

The procedurally defined action s taken to secure primary containment and its associated structures, systems, and components as a functional barrier to fission product release under shutdown existing plant conditions (ref. 4.1.10). As applied to ANO, Containment Closure must be capable of being set within 30 minutes.

Containment Closure is set when the penetrations are isolated by manual or automatic isolation valve, blind flange, or equivalent.

Emergency Action Level (EAL)

A pre-determined, site-specific, observable threshold for an INITIATING CONDITION that, when met or exceeded, places the plant in a given emergency classification level.

Emergency Classification Level (ECL)

One of a set of names or titles established by the US Nuclear Regulatory Commission (NRC) for grouping off-normal events or conditions according to (1) potential or actual effects or consequences, and (2) resulting onsite and offsite response actions. The emergency classification levels, in ascending order of severity, are:

Notification of Unusual Event (NO UE) Alert Site Area Emergency (SAE)

General Emergency (GE)

Explosion

A rapid, violent and catastrophic failure of a piece of equipment due to combustion, chemical reaction or overpressurization. A release of steam (from high energy lines or components) or

an electrical component failure (caused by short circuits, grounding, arcing, etc.) should not automatically be considered an explosion. Such events require a post-event inspection to determine if the attributes of an explosion are present.

Faulted The term applied to a steam generator that has a steam leak on the secondary side of sufficient size to cause an uncontrolled drop in steam generator pressure or the steam generator to become completely depressurized.

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Page 15 of 257 Fire Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred but is not required if large quantities of smoke and heat are observed.

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

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

General Emergency Events are in 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 PAG exposure levels offsite for more than the immediate site area.

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

Hostile Action An act toward a NPP ANO 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 the NPPANO. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the SECURITY OWNER CONTROLLED AREA).

Hostile Force One or more individuals who are engaged in a determined assault, overtly or by stealth and deception, equipped with suitable weapons capable of killing, maiming, or causing destruction.

Imminent The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions.

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Page 16 of 257 Impede(d)

Personnel access to a room or area is hindered to an extent that extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., requiring use of protective equipment, such as SCBAs, that is not routinely employed).

Independent Spent Fuel Storage Installation (ISFSI)

A complex that is designed and constructed for the interim storage of spent nuclear fuel and other radioactive materials associated with spent fuel storage.

Initiating Condition (IC)

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

Normal Levels As applied to radiological IC/EALs, the highest reading in the past twenty-four hours excluding the current peak value.

Owner Controlled Area (OCA)

Projectile

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

Protected Area An area clearly demarcated by a fence or building wall with an entrance portal that is regulated by Security Personnel to control access.

RCS Intact The RCS should be considered intact when the RCS pressure boundary is in its normal condition for the cold shutdown mode of operation (e.g., no freeze seals or nozzle dams).

Refueling Pathway All the cavities, tubes, canals and pools through which irradiated fuel may be moved, but not including the reactor vessel.

Ruptured The condition of a steam generator in which primary-to-secondary leakage is of sufficient magnitude to require a safety injection (automatic or manual). to 0CAN091801

Page 17 of 257 Safety System A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2):

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

(1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.

Security Condition Any security event as listed in the approved security contingency plan that constitutes a threat/compromise to site security, threat/risk to site personnel, or a potential degradation to the level of safety of the plant. A Security Condition does not involve a HOSTILE ACTION.

Security Owner Controlled Area (SOCA)

The SOCA is the area demarcated as a Vehicle Barrier System (VBS) consisting of passive elements including a series of large concrete blocks on the inside of a delay fence with early warning capabilities. The SOCA is the area between the SOCA Fence and the PROTECTED AREA Boundary.

Site Area Emergency Events are in progress or have occurred which involve actual or likely major failures of plant functions needed for protection of the public or HOSTILE ACTION that results in intentional damage or malicious acts; (1) toward site personnel or equipment that could lead to the likely failure of or; (2) that prevent effective access to equipment needed for the protection of the

public. Any releases are not expected to result in exposure levels which exceed EPA PAG exposure levels beyond the SITE BOUNDARY.

Site Boundary That boundary defined by a 1046 meter (0.65 mile) radius around the plant (ref. 4.1.7).

Unisolable

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

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

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Page 18 of 257 Unusual Event Events are in progress or have occurred which indicate a potential degradation in the level of safety of the plant or indicate a security threat to facility protection has been initiated. No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of SAFETY SYSTEMS occurs.

Valid An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the condition's existence, or the report's accuracy is removed.

Implicit in this definition is the need for timely assessment.

Visible Damage Damage to a SAFETY SYSTEM train that is readily observable without measurements, testing, or analysis.

The visual impact of the damage is sufficient to cause concern regarding the operability or reliability of the affected SAFETY SYSTEM train.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.

5.2 Abbreviations/Acronyms

°F .................................................................................................................... Degrees Fahrenheit

° ...............................................................................................................................

.......... Degrees AC ..................................................................................................................... Altern ating Current ANO ............................................................................................................ Arkansas Nucle ar One AOP .............................................................................................. Abnormal Operating Procedure ATWS .................................................................................... Anticipated Transient Without Scram BMS ................................................................................................... Boron Management System BWST ................................................................................................ Borated Water Storage Tank CDE ................................................................................................... Committed Dose Equivale nt CET .......................................................................................................... Core Exit Thermo couple CFR ................................................................................................... Code of Federal Regulations CIAS .................................................................................. Containment Isolation Actuation Signal CMT, CNTMT, CTMT .................................................................................................. Containment CNB ................................................................................................................ Containmen t Barrier DBA ............................................................................................................. Design Basis Accident DBE ......................................................................................................... Design Basis Earthqua ke to 0CAN091801

Page 19 of 257 DC ............................................................................................................................. Direct Current DEF ...........................................................................................................................

........ Defueled D/G ....................................................................................................................... Diesel Generator DHR .............................................................................................................. Decay Heat Removal DROPS ............................................................ Diverse Reactor Overpressure Protection System DSC ............................................................................................................. Dry Shielded Canister DSS ............................................................................................................ Diverse Scram Syst em EAL .......................................................................................................... Emergency Action Level ECCS ......................................................................................... Emergency Core Cooling System ECL ............................................................................................... Emergency Classification Le vel DEF ...........................................................................................................................

........ Defueled ENS ............................................................................................... Emergency Notification System EOF ................................................................................................ Emergency Operations Facility EOP ........................................................................................... Emergency Operating Procedure EPA ............................................................................................ Environmental Protection Agenc y ERG ............................................................................................ Emergency Response Guideline EPIP ............................................................................. Emergency Plan Implementing Procedure ESAS ........................................................................... Engineered Safeguards Actuation System ESF ...................................................................................................... Engineered Safety Fe ature ESFAS .................................................................. Engineered Safety Features Actuation System FAA ............................................................................................... Federal Aviation Administration FBI ................................................................................................ Federal Bureau of Investigation FCB ...................................................................................................................... Fuel Clad Barrier FEMA ............................................................................ Federal Emergency Management Agency GE ................................................................................................................... General Emergency HPI ............................................................................................................. High Pressure Injection IC ...................................................................................................................... Initiating Condition IPEEE ............................. Individual Plant Examination of External Events (Generic Letter 88-20)

ISFSI ......................................................................... Independent Spent Fuel Storage Installation

K eff ...................................................................................... Effective Neutron Multiplication Factor LCO ................................................................................................

Limiting Condition of Operation LER .............................................................................................................

Licensee Event Report LNEPR ........................................................................... Lower North Electrical Penetration Room LOCA ...................................................................................................... Loss of Coolant Accident to 0CAN091801

Page 20 of 257 LRW .................................................................................................................... Liquid Rad Waste LTOP ............................................................................................ Low Temperature Overpressure LWR ................................................................................................................ Light Wate r Reactor MCC .............................................................................................................. Motor Contro l Center MPC ................................................ Maximum Permissible Concentration/Multi-Purpose Canister mR, mRem, mrem, mREM ............................................................ milli-Roentgen Equivalent Man MSL ...................................................................................................................... Main Ste am Line MTS ................................................................................................................ Margin to Saturation MW .................................................................................................................................. Megawatt NDTT ...................................................................................... Nil Ductility Transition Temperatur e NEI ............................................................................................................

Nuclear Energy Institute NEIC ................................................................................ National Earthquake Information Center NESP ................................................................................ National Environmental Studies Project NORAD ................................................................ North American Aerospace Defense Command NOT .............................................................................................. Normal Operating Temperature (NO)UE ............................................................................................. Notification of Unusual Event NPP ................................................................................................................ Nuclear Power Plant NRC ............................................................................................ Nuclear Regulatory Commission NSSS ............................................................................................. Nuclear Steam Supply System OBE ................................................................................................... Operating Basis Earthqu ake ODCM ........................................................................................ Off-site Dose Calculation Manual ORO ............................................................................................... Offsite Response Organization PA ...........................................................................................................................

Protected Area PAG ..................................................................................................... Protective Action Gui deline PRA/PSA .................................. Probabilistic Risk Assessment / Probabilistic Safety Assessment P-T .............................................................................................................. Pressure-Temperature PTS ..................................................................................................... Pressurized Thermal Shock PWR ..................................................................................................... Pressurized Water Reactor PSIG ............................................................................................ Pounds per Square Inch Gauge

R .............................................................................................................................

........ Roentgen RB .........................................................................................................................

Reactor Building RCC ......................................................................................................... Reactor Control C onsole RCB ............................................................................................. Reactor Coolant System Barrie r RCP ............................................................................................................ Reactor Coolan t Pump to 0CAN091801

Page 21 of 257 RCS ......................................................................................................... Reactor Coolant System Rem, rem, REM ..................................................................................... Roentgen Equivalent Man Rep CET ....................................................................... Representative Core Exit Thermocouples RETS ...................................................................... Radiological Effluent Technical Specifications RPS ...................................................................................................... Reactor Protection Syste m RV ...........................................................................................................................

Reactor Vessel RVLMS ........................................................................... Reactor Vessel Level Monitoring System RWT ............................................................................................................. Refueling Wat er Tank SAR ............................................................................................................. Safety Analysis Report SBO ...................................................................................................................... Stat ion Blackout SCBA ................................................................................... Self-Contained Breathing Apparatus SDC ................................................................................................................... Shutdown Cooling SOCA ........................................................................................... Security Owner Controlled Area SG ....................................................................................................................... Steam Generator SI ............................................................................................................................ Sa fety Injection SPDS ........................................................................................ Safety Parameter Display System SPING ..................................................................................... Super Particulate Iodine Noble Gas SRO ......................................................................................................... Senior Reactor Op erator TEDE ............................................................................................ Total Effective Dose Equivale nt TOAF ................................................................................................................. Top of A ctive Fuel TSC ........................................................................................................ Technical Support Center UNEPR .......................................................................... Upper North Electrical Penetration Room USGS .......................................................................................... United States Geological Survey VBS ............................................................................................................ Vehicle Barrie r System to 0CAN091801

Page 22 of 257 6.0 ANO-TO-NEI 99-01 REV. 6 EAL CROSS-REFERENCE This cross-reference is provided to facilitate association and location of an ANO EAL within the NEI 99-01 IC/EAL identification scheme. Further information regarding the development of the ANO EALs based on the NEI guidance can be found in the EAL Comparison Matrix.

ANO NEI 99-01 Rev. 6 EAL IC Example EAL AU1.1 AU1 1, 2 AU1.2 AU1 3 AU2.1 AU2 1 AA1.1 AA1 1 AA1.2 AA1 2 AA1.3 AA1 3 AA1.4 AA1 4 AA2.1 AA2 1 AA2.2 AA2 2 AA2.3 AA2 3 AA3.1 AA3 1 AA3.2 AA3 2 AS1.1 AS1 1 AS1.2 AS1 2 AS1.3 AS1 3 AS2.1 AS2 1 AG1.1 AG1 1 AG1.2 AG1 2 AG1.3 AG1 3 AG2.1 AG2 1 CU1.1 CU1 1 CU1.2 CU1 2 CU2.1 CU2 1 CU3.1 CU3 1 CU3.2 CU3 2 CU4.1 CU4 1 to 0CAN091801

Page 23 of 257 ANO NEI 99-01 Rev. 6 EAL IC Example EAL CU5.1 CU5 1, 2, 3 CA1.1 CA1 1 CA1.2 CA1 2 CA2.1 CA2 1 CA3.1 CA3 1, 2 CA6.1 CA6 1 CS1.1 CS1 1 CS1.2 CS1 2 CS1.3 CS1 3 CG1.1 CG1 1 CG1.2 CG1 2 EU1.1 EU1 1 FA1.1 FA1 1 FS1.1 FS1 1 FG1.1 FG1 1 HU1.1 HU1 1, 2, 3 HU2.1 HU2 1 HU3.1 HU3 1 HU3.2 HU3 2 HU3.3 HU3 3 HU3.4 HU3 4 HU4.1 HU4 1 HU4.2 HU4 2 HU4.3 HU4 3 HU4.4 HU4 4 HU7.1 HU7 1 HA1.1 HA1 1, 2 HA5.1 HA5 1 HA6.1 HA6 1 HA7.1 HA7 1 HS1.1 HS1 1 to 0CAN091801

Page 24 of 257 ANO NEI 99-01 Rev. 6 EAL IC Example EAL HS6.1 HS6 1 HS7.1 HS7 1 HG7.1 HG7 1 SU1.1 SU1 1 SU3.1 SU2 1 SU4.1 SU3 1 SU4.2 SU3 2 SU5.1 SU4 1, 2, 3 SU6.1 SU5 1 SU6.2 SU5 2 SU7.1 SU6 1, 2, 3 SU8.1 SU7 1, 2 SA1.1 SA1 1 SA3.1 SA2 1 SA6.1 SA5 1 SA9.1 SA9 1 SS1.1 SS1 1 SS2.1 SS8 1 SS6.1 SS5 1 SG1.1 SG1 1 SG1.2 SG8 1

7.0 ATTACHMENTS

7.1 Attachment

1, Emergency Action Level Technical Bases

7.2 Attachment

2, Safe Operation & Shutdown Areas Tables 1[2]A-3 & 1[2]H-2 Bases

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Page 25 of 257 Attachment 1 - Emergency Action Level Technical Bases

Category A - Abnormal Rad Levels / Rad Effluent EAL Group: ANY (EALs in this category are applicable to any plant condition, hot or cold.)

Many EALs are based on actual or potential degradation of fission product barriers because of the elevated potential for offsite radioactivity release. Degradation of fission product barriers though is not always apparent via non-radiological symptoms. Therefore, direct indication of elevated radiological effluents or area radiation levels are appropriate symptoms for emergency classification.

At lower levels, abnormal radioactivity releases may be indicative of a failure of containment systems or precursors to more significant releases. At higher release rates, offsite radiological conditions may result which require offsite protective actions. Elevated area radiation levels in plant may also be indicative of the failure of containment systems or preclude access to plant vital equipment necessary to ensure plant safety.

Events of this category pertain to the following subcategories:

1. Radiological Effluent Direct indication of effluent radiation monitoring systems provides a rapid assessment mechanism to determine releases in excess of classifiable limits. Projected offsite doses, actual offsite field measurements or measured release rates via sampling indicate doses or dose rates above classifiable limits.
2. Irradiated Fuel Event Conditions indicative of a loss of adequate shielding or damage to irradiated fuel may preclude access to vital plant areas or result in radiological releases that warrant emergency classification.
3. Area Radiation Levels Sustained general area radiation levels which may preclude access to areas requiring continuous occupancy also warrant emergency classification.

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Page 26 of 257 Attachment 1 - Emergency Action Level Technical Bases

Category: A - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent

Initiating Condition: Release of gaseous or liquid radioactivity greater than 2 times the ODCM limits for 60 minutes or longer

EAL: AU1.1 Unusual Event Reading on any Table 1[2]A-1 effluent radiation monitor > column "UE" for 60 min. (Notes 1, 2, 3)

Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is 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 due to actions to isolate the release path, then the effluent monitor reading is no longer VALID for classification purposes. Table 1A-1 Unit 1 Effluent Monitor Classification Thresholds (2 min. avg reading)

Release Point Monitor GE SAE Alert UE Gaseous Containment Purge RX-9820 (SPING 1) 4.15E+01 Ci/cc 4.15E+00 µCi/cc4.15E-01 µCi/cc 1.21E-03 µCi/cc Radwaste Area RX-9825 (SPING 2) 2.67E+01 Ci/cc 2.67E+00 µCi/cc2.67E-01 µCi/cc 4.94E-04 µCi/cc Fuel Handling Area RX-9830 (SPING 3) 6.20E+02 Ci/cc 6.20E+01 µCi/cc6.20E+00 µCi/cc 5.44E-04 µCi/cc Emergency Penetration Room RX-9835 (SPING 4) 6.55E+02 Ci/cc 6.55E+01 µCi/cc6.55E+00 µCi/cc 1.21E-02 µCi/cc Liquid Liquid Radwaste RE-4642 ----


---- 2.46E+05 cpm to 0CAN091801

Page 27 of 257 Attachment 1 - Emergency Action Level Technical Bases

Table 2A-1 Unit 2 Effluent Monitor Classification Thresholds (2 min. avg reading)

Release Point Monitor GE SAE Alert UE Gaseous Containment Purge 2RX-9820 (SPING 5) 1.88E+01 Ci/cc 1.88E+00 µCi/cc1.88E-01 µCi/cc 5.48E-04 µCi/cc Radwaste Area 2RX-9825 (SPING 6) 2.35E+01 Ci/cc 2.35E+00 µCi/cc2.35E-01 µCi/cc 4.35E-04 µCi/cc Fuel Handling Area 2RX-9830 (SPING 7) 6.86E+02 Ci/cc 6.86E+01 µCi/cc6.86E+00 µCi/cc 6.04E-04 µCi/cc Emergency Penetration Room 2RX-9835 (SPING 8) 5.88E+02 Ci/cc 5.88E+01 µCi/cc5.88E+00 µCi/cc 1.09E-02 µCi/cc Liquid BMS Liquid Discharge 2RE-2330 ----


---- 2.45E+04 cpm Regenerative Waste Discharge 2RE-4423 ----


---- 2.45E+05 cpm Mode Applicability:

All Definition(s):

VALID - An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment.

Basis: This IC addresses a potential decrease reduction 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. to 0CAN091801

Page 28 of 257 Attachment 1 - Emergency Action Level Technical Bases

Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more 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 #1 - This EAL addresses normally occurring continuous radioactivity releases from monitored gaseous or liquid effluent pathways . EAL #2 - This EAL addressesas well as radioactivity releases that cause effluent radiation monitor readings to exceed 2 times the limit established by a radioactivity discharge permit. This EAL will Such releases are 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 AA1.

Reference(s):

1. OP-1604.051 Eberline Radiation Monitor System 2. Offsite Dose Calculation Manual
3. EP-CALC-ANO-1701 Radiological Effluent EAL Values
4. NEI 99-01 AU1 to 0CAN091801

Page 29 of 257 Attachment 1 - Emergency Action Level Technical Bases

Category: A - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent

Initiating Condition: Release of gaseous or liquid radioactivity greater than 2 times the ODCM limits for 60 minutes or longer

EAL: AU1.2 Unusual Event Sample analysis for a gaseous or liquid release indicates a concentration or release rate 2 x ODCM limits for 60 min. (Notes 1, 2)

Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded.

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

Mode Applicability:

All

Definition(s):

None Basis: This IC addresses a potential decrease reduction 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.

to 0CAN091801

Page 30 of 257 Attachment 1 - Emergency Action Level Technical Bases

Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer VALID for classification purposes.

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 #1 - This EAL addresses normally occurring continuous radioactivity releases from monitored gaseous or liquid effluent pathways.

EAL #2 - This EAL addresses radioactivity releases that cause effluent radiation monitor readings to exceed 2 times the limit establishe d 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 AA1.

Reference(s):

1. Offsite Dose Calculation Manual
2. NEI 99-01 AU1

to 0CAN091801

Page 31 of 257 Attachment 1 - Emergency Action Level Technical Bases

Category: A - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent

Initiating Condition: Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid CDE

EAL: AA1.1 Alert Reading on any Table 1[2]A-1 effluent radiation monitor > column "ALERT" for 15 min. (Notes 1, 2, 3, 4)

Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is 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 due to actions to isolate the release path, then the effluent monitor reading is no longer VALID for classification purposes. Note 4 The pre-calculated effluent monitor values presented in EALs AA1.1, AS1.1 and AG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.

Table 1A-1 Unit 1 Effluent Monitor Classification Thresholds (2 min. avg reading)

Release Point Monitor GE SAE Alert UE Gaseous Containment Purge RX-9820 (SPING 1) 4.15E+01 Ci/cc 4.15E+00 µCi/cc4.15E-01 µCi/cc 1.21E-03 µCi/cc Radwaste Area RX-9825 (SPING 2) 2.67E+01 Ci/cc 2.67E+00 µCi/cc2.67E-01 µCi/cc 4.94E-04 µCi/cc Fuel Handling Area RX-9830 (SPING 3) 6.20E+02 Ci/cc 6.20E+01 µCi/cc6.20E+00 µCi/cc 5.44E-04 µCi/cc Emergency Penetration Room RX-9835 (SPING 4) 6.55E+02 Ci/cc 6.55E+01 µCi/cc6.55E+00 µCi/cc 1.21E-02 µCi/cc Liquid Liquid Radwaste RE-4642 ----


---- 2.46E+05 cpm to 0CAN091801

Page 32 of 257 Attachment 1 - Emergency Action Level Technical Bases

Table 2A-1 Unit 2 Effluent Monitor Classification Thresholds (2 min. avg reading)

Release Point Monitor GE SAE Alert UE Gaseous Containment Purge 2RX-9820 (SPING 5) 1.88E+01 Ci/cc 1.88E+00 µCi/cc1.88E-01 µCi/cc 5.48E-04 µCi/cc Radwaste Area 2RX-9825 (SPING 6) 2.35E+01 Ci/cc 2.35E+00 µCi/cc2.35E-01 µCi/cc 4.35E-04 µCi/cc Fuel Handling Area 2RX-9830 (SPING 7) 6.86E+02 Ci/cc 6.86E+01 µCi/cc6.86E+00 µCi/cc 6.04E-04 µCi/cc Emergency Penetration Room 2RX-9835 (SPING 8) 5.88E+02 Ci/cc 5.88E+01 µCi/cc5.88E+00 µCi/cc 1.09E-02 µCi/cc Liquid BMS Liquid Discharge 2RE-2330 ----


---- 2.45E+04 cpm Regenerative Waste Discharge 2RE-4423 ----


---- 2.45E+05 cpm Mode Applicability:

All Definition(s):

VALID - An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment.

Basis: 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. to 0CAN091801

Page 33 of 257 Attachment 1 - Emergency Action Level Technical Bases

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

Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer VALID for classification purposes.

Escalation of the emergency classification level would be via IC AS1.

Reference(s):

1. OP-1604.051 Eberline Radiation Monitor System 2. EP-CALC-ANO-1701 Radiological Effluent EAL Values
3. NEI 99-01 AA1

to 0CAN091801

Page 34 of 257 Attachment 1 - Emergency Action Level Technical Bases

Category: A - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent

Initiating Condition: Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid CDE

EAL: AA1.2 Alert Dose assessment using actual meteorology indicates doses > 10 mrem TEDE or 50 mrem thyroid CDE at or beyond the SITE BOUNDARY. (Note 4)

Note 4: The pre-calculated effluent monitor values presented in EALs AA1.1, AS1.1 and AG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.

Mode Applicability:

All

Definition(s):

SITE BOUNDARY - That boundary defined by a 1046 meter (0.65 mile) radius around the plant.

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. to 0CAN091801

Page 35 of 257 Attachment 1 - Emergency Action Level Technical Bases

Escalation of the emergency classification level would be via IC AS1.

Reference(s):

1. OP-1904.002 Offsite Dose Projections
2. NEI 99-01 AA1 to 0CAN091801

Page 36 of 257 Attachment 1 - Emergency Action Level Technical Bases

Category: A - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent

Initiating Condition: Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid CDE

EAL: AA1.3 Alert Analysis of a liquid effluent sample indicates a concentration or release rate that would result in doses > 10 mrem TEDE or 50 mrem thyroid CDE at or beyond the SITE BOUNDARY for 60 min. of exposure. (Notes 1, 2)

Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.

Mode Applicability:

All

Definition(s):

SITE BOUNDARY - That boundary defined by a 1046 meter (0.65 mile) radius around the plant.

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.

to 0CAN091801

Page 37 of 257 Attachment 1 - Emergency Action Level Technical Bases

Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.This EAL is assessed per the ODCM (ref. 2).

Escalation of the emergency classification level would be via IC AS1.

Reference(s):

1. OP-1904.002 Offsite Dose Projections
2. Offsite Dose Calculation Manual
3. NEI 99-01 AA1 to 0CAN091801

Page 38 of 257 Attachment 1 - Emergency Action Level Technical Bases

Category: A - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent

Initiating Condition: Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid CDE

EAL: AA1.4 Alert Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY: Closed window dose rates > 10 mR/hr expected to continue for 60 min. Analyses of field survey samples indicate thyroid CDE > 50 mrem for 60 min. of inhalation. (Notes 1, 2)

Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.

Mode Applicability:

All Definition(s):

SITE BOUNDARY - That boundary defined by a 1046 meter (0.65 mile) radius around the plant.

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.

to 0CAN091801

Page 39 of 257 Attachment 1 - Emergency Action Level Technical Bases

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

Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.

Escalation of the emergency classification level would be via IC AS1.

Reference(s):

1. OP-1905.002 Offsite Emergency Monitoring 2. NEI 99-01 AA1

to 0CAN091801

Page 40 of 257 Attachment 1 - Emergency Action Level Technical Bases

Category: A - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent

Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem thyroid CDE

EAL: AS1.1 Site Area Emergency Reading on any Table 1[2]A-1 effluent radiation monitor > column "SAE" for 15 min. (Notes 1, 2, 3, 4)

Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is 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 due to actions to isolate the release path, then the effluent monitor reading is no longer VALID for classification purposes. Note 4: The pre-calculated effluent monitor values presented in EALs AA1.1, AS1.1 and AG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.

Table 1A-1 Unit 1 Effluent Monitor Classification Thresholds (2 min. avg reading)

Release Point Monitor GE SAE Alert UE Gaseous Containment Purge RX-9820 (SPING 1) 4.15E+01 Ci/cc 4.15E+00 µCi/cc4.15E-01 µCi/cc 1.21E-03 µCi/cc Radwaste Area RX-9825 (SPING 2) 2.67E+01 Ci/cc 2.67E+00 µCi/cc2.67E-01 µCi/cc 4.94E-04 µCi/cc Fuel Handling Area RX-9830 (SPING 3) 6.20E+02 Ci/cc 6.20E+01 µCi/cc6.20E+00 µCi/cc 5.44E-04 µCi/cc Emergency Penetration Room RX-9835 (SPING 4) 6.55E+02 Ci/cc 6.55E+01 µCi/cc6.55E+00 µCi/cc 1.21E-02 µCi/cc Liquid Liquid Radwaste RE-4642 ----


---- 2.46E+05 cpm to 0CAN091801

Page 41 of 257 Attachment 1 - Emergency Action Level Technical Bases

Table 2A-1 Unit 2 Effluent Monitor Classification Thresholds (2 min. avg reading)

Release Point Monitor GE SAE Alert UE Gaseous Containment Purge 2RX-9820 (SPING 5) 1.88E+01 Ci/cc 1.88E+00 µCi/cc1.88E-01 µCi/cc 5.48E-04 µCi/cc Radwaste Area 2RX-9825 (SPING 6) 2.35E+01 Ci/cc 2.35E+00 µCi/cc2.35E-01 µCi/cc 4.35E-04 µCi/cc Fuel Handling Area 2RX-9830 (SPING 7) 6.86E+02 Ci/cc 6.86E+01 µCi/cc6.86E+00 µCi/cc 6.04E-04 µCi/cc Emergency Penetration Room 2RX-9835 (SPING 8) 5.88E+02 Ci/cc 5.88E+01 µCi/cc5.88E+00 µCi/cc 1.09E-02 µCi/cc Liquid BMS Liquid Discharge 2RE-2330 ----


---- 2.45E+04 cpm Regenerative Waste Discharge 2RE-4423 ----


---- 2.45E+05 cpm Mode Applicability:

All Definition(s):

VALID - An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment.

Basis: 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.

to 0CAN091801

Page 42 of 257 Attachment 1 - Emergency Action Level Technical Bases

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 AG1.

Reference(s):

1. OP-1604.051 Eberline Radiation Monitor System 2. EP-CALC-ANO-1701 Radiological Effluent EAL Values
3. NEI 99-01 AS1

to 0CAN091801

Page 43 of 257 Attachment 1 - Emergency Action Level Technical Bases

Category: A - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent

Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem thyroid CDE

EAL: AS1.2 Site Area Emergency Dose assessment using actual meteorology indicates doses > 100 mrem TEDE or 500 mrem thyroid CDE at or beyond the SITE BOUNDARY. (Note 4)

Note 4 The pre-calculated effluent monitor values presented in EALs AA1.1, AS1.1 and AG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.

Mode Applicability:

All

Definition(s):

SITE BOUNDARY - That boundary defined by a 1046 meter (0.65 mile) radius around the plant).

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 AG1. to 0CAN091801

Page 44 of 257 Attachment 1 - Emergency Action Level Technical Bases

Reference(s):

1. OP-1904.002 Offsite Dose Projections 2. NEI 99-01 AS1

to 0CAN091801

Page 45 of 257 Attachment 1 - Emergency Action Level Technical Bases

Category: A - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent

Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem thyroid CDE

EAL: AS1.3 Site Area Emergency Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY: Closed window dose rates > 100 mR/hr expected to continue for 60 min. Analyses of field survey samples indicate thyroid CDE > 500 mrem for 60 min. of inhalation. (Notes 1, 2)

Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.

Mode Applicability:

All Definition(s):

SITE BOUNDARY - That boundary defined by a 1046 meter (0.65 mile) radius around the plant.

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.

to 0CAN091801

Page 46 of 257 Attachment 1 - Emergency Action Level Technical Bases

Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.

Escalation of the emergency classification level would be via IC AG1.

Reference(s):

1. OP-1905.002 Offsite Emergency Monitoring
2. NEI 99-01 AS1 to 0CAN091801

Page 47 of 257 Attachment 1 - Emergency Action Level Technical Bases

Category: A - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent

Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid CDE

EAL: AG1.1 General Emergency Reading on any Table 1[2]A-1 effluent radiation monitor > column "GE" for 15 min. (Notes 1, 2, 3, 4)

Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is 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 due to actions to isolate the release path, then the effluent monitor reading is no longer VALID for classification purposes. Note 4: The pre-calculated effluent monitor values presented in EALs AA1.1, AS1.1 and AG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.

Table 1A-1 Unit 1 Effluent Monitor Classification Thresholds (2 min. avg reading)

Release Point Monitor GE SAE Alert UE Gaseous Containment Purge RX-9820 (SPING 1) 4.15E+01 Ci/cc 4.15E+00 µCi/cc4.15E-01 µCi/cc 1.21E-03 µCi/cc Radwaste Area RX-9825 (SPING 2) 2.67E+01 Ci/cc 2.67E+00 µCi/cc2.67E-01 µCi/cc 4.94E-04 µCi/cc Fuel Handling Area RX-9830 (SPING 3) 6.20E+02 Ci/cc 6.20E+01 µCi/cc6.20E+00 µCi/cc 5.44E-04 µCi/cc Emergency Penetration Room RX-9835 (SPING 4) 6.55E+02 Ci/cc 6.55E+01 µCi/cc6.55E+00 µCi/cc 1.21E-02 µCi/cc Liquid Liquid Radwaste RE-4642 ----


---- 2.46E+05 cpm to 0CAN091801

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Table 2A-1 Unit 2 Effluent Monitor Classification Thresholds (2 min. avg reading)

Release Point Monitor GE SAE Alert UE Gaseous Containment Purge 2RX-9820 (SPING 5) 1.88E+01 Ci/cc 1.88E+00 µCi/cc1.88E-01 µCi/cc 5.48E-04 µCi/cc Radwaste Area 2RX-9825 (SPING 6) 2.35E+01 Ci/cc 2.35E+00 µCi/cc2.35E-01 µCi/cc 4.35E-04 µCi/cc Fuel Handling Area 2RX-9830 (SPING 7) 6.86E+02 Ci/cc 6.86E+01 µCi/cc6.86E+00 µCi/cc 6.04E-04 µCi/cc Emergency Penetration Room 2RX-9835 (SPING 8) 5.88E+02 Ci/cc 5.88E+01 µCi/cc5.88E+00 µCi/cc 1.09E-02 µCi/cc Liquid BMS Liquid Discharge 2RE-2330 ----


---- 2.45E+04 cpm Regenerative Waste Discharge 2RE-4423 ----


---- 2.45E+05 cpm Mode Applicability:

All Definition(s):

VALID - An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment.

Basis: 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.

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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.

Reference(s):

1. OP-1604.051 Eberline Radiation Monitor System 2. EP-CALC-ANO-1701 Radiological Effluent EAL Values 3. NEI 99-01 AG1

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Category: A - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent

Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid CDE

EAL: AG1.2 General Emergency Dose assessment using actual meteorology indicates doses > 1,000 mrem TEDE or 5,000 mrem thyroid CDE at or beyond the SITE BOUNDARY (Note 4)

Note 4 The pre-calculated effluent monitor values presented in EALs AA1.1, AS1.1 and AG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.

Mode Applicability:

All

Definition(s):

SITE BOUNDARY - That boundary defined by a 1046 meter (0.65 mile) radius around the plant).

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.

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Reference(s):

1. OP-1904.002 Offsite Dose Projections 2. NEI 99-01 AG1

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Category: A - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent

Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid CDE

EAL: AG1.3 General Emergency Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY: Closed window dose rates > 1,000 mR/hr expected to continue for 60 min. Analyses of field survey samples indicate thyroid CDE > 5,000 mrem for 60 min. of inhalation (Notes 1, 2)

Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.

Mode Applicability:

All Definition(s):

SITE BOUNDARY - That boundary defined by a 1046 meter (0.65 mile) radius around the plant.

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.

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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.

Reference(s):

1. OP-1905.002 Offsite Emergency Monitoring 2. NEI 99-01 AG1

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Category: A - Abnormal Rad Levels / Rad Effluent Subcategory: 2 - Irradiated Fuel Event

Initiating Condition: UNPLANNED loss of water level above irradiated fuel EAL: AU2.1 Unusual Event UNPLANNED water level drop in the REFUELING PATHWAY as indicated by low water level alarm, visual observation, or BWST[RWT] level drop due to makeup demands AND UNPLANNED rise in corresponding area radiation levels as indicated by any of the following radiation monitors: Unit 1 o RE-8009 Spent Fuel Area o RE-8017 Fuel Handling Area Unit 2 o 2RE-8914 Spent Fuel Area o 2RE-8915 Spent Fuel Area o 2RE-8916 Spent Fuel Area o 2RE-8912 Containment Incore Instrumentation 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 - All the cavities, tubes, canals and pools through which irradiated fuel may be moved, but not including the reactor vessel.

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Basis:

This IC addresses a decrease drop 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 drop 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 a rise 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 rise 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 AA2.

Reference(s):

1. OP-1203.050 Unit 1 Spent Fuel Pool Emergencies 2. OP-2203.002 Spent Fuel Pool Emergencies
3. 1SAR 11.2.5 Area Radiation Monitoring Systems Table 11-15 Area Radiation Monitors
4. 2SAR 12.1.4 Area Radiation Monitoring System 5. NEI 99-01 AU2

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Category: A - Abnormal Rad Levels / Rad Effluent Subcategory: 2 - Irradiated Fuel Event

Initiating Condition: Significant lowering of water level above, or damage to, irradiated fuel EAL: AA2.1 Alert IMMINENT uncovery of irradiated fuel in the REFUELING PATHWAY.

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 ANO ISFSI, the Confinement Boundary is comprised of either the Multi-assembly Sealed Basket (MSB) (SNC System) or Multi-Purpose Canister (MPC) (Holtec System).

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.

REFUELING PATHWAY - All the cavities, tubes, canals and pools through which irradiated fuel may be moved, but not including the reactor vessel.

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 poolREFUELING PATHWAY (see Developer 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 applies to irradiated fuel that is licensed for dry storage up to the point that the loaded storage cask is sealed. Once sealed, damage to a loaded cask causing loss of the CONFINEMENT BOUNDARY is classified in accordance with IC E-HU1. Escalation of the emergency would be based on either Recognition Category A or C ICs.

EAL #1 to 0CAN091801

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This EAL escalates from AU2 A U2.1 in that the loss of level, in the affected portion of the REFUELING PATHWAY, is of sufficient magnitude to have resulted in uncovery of irradiated fuel. Indications of irradiated fuel uncovery may include direct or indirect visual observation (e.g., reports from personnel or camera images), as well as significant changes in water and radiation levels, or other plant parameters. Computational aids may also be used (e.g., a boil-off curve). Classification of an event using this EAL should be based on the totality of available indications, reports and observations.

While an area radiation monitor could detect an increasea rise 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 with 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 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 #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 IC s AS1 AS1or AS2 (see AS2 Developer Notes

). Reference(s):

1. NEI 99-01 AA2

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Category: A - Abnormal Rad Levels / Rad Effluent Subcategory: 2 - Irradiated Fuel Event

Initiating Condition: Significant lowering of water level above, or damage to, irradiated fuel EAL: AA2.2 Alert Damage to irradiated fuel resulting in a release of radioactivity AND High alarm on any Table 1[2]A-2 radiation monitor.

Table 1A-2 Unit 1 Fuel Damage Radiation Monitors RE-8009 Spent Fuel Area RE-8017 Fuel Handling RE-8060 Containment High Range Radiation Monitor RE-8061 Containment High Range Radiation Monitor RX-9820 (SPING 1) Containment Purge RX-9825 (SPING 2) Radwaste Area RX-9830 (SPING 3) Fuel Handling Area Table 2A-2 Unit 2 Fuel Damage Radiation Monitors 2RE-8905 Containment Equipment Hatch Area 2RE-8909 Containment Personnel Access Area 2RE-8912 Containment Incore Inst. 2RE-8914 Spent Fuel Area 2RE-8915 Spent Fuel Area 2RE-8916 Spent Fuel Area 2RE-8925-1 Containment High Range Radiation Monitor 2RE-8925-2 Containment High Range Radiation Monitor 2RX-9820 (SPING 5) Containment Purge 2RX-9825 (SPING 6) Radwaste Area 2RX-9830 (SPING 7) Fuel Handling Area to 0CAN091801

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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 ANO ISFSI, the Confinement Boundary is comprised of either the Multi-assembly Sealed Basket (MSB) (SNC System) or Multi-Purpose Canister (MPC) (Holtec System).

VALID - An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment.

Basis: This IC EAL 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 Developer 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 EAL applies to irradiated fuel that is licensed for dry storage up to the point that the loaded storage cask is sealed. Once sealed, damage to a loaded cask causing loss of the CONFINEMENT BOUNDARY is classified in accordance with IC E-HU1. EAL #This EAL escalates from AU2 in that the loss of level, in the affected portion of the REFUELING PATHWAY, is of sufficient magnitude to have resulted in uncovery of irradiated fuel. Indications of irradiated fuel uncovery may include direct or indirect visual observation (e.g., reports from personnel or camera images), as well as significant changes in water and radiation levels, or other plant parameters. Computational aids may also be used (e.g., a boil-off curve). Classification of an event using this EAL should be based on the totality of available indications, reports and observations.

While an area radiation monitor could detect an increase in a dose rate due to a lowering of water level in some portion of the REFUELING PATHWAY, the reading may not be a reliable indication of whether or not the fuel is actually uncovered. To the degree possible, readings should be considered in combination with other available indications of inventory loss.

A drop in water level above irradiated fuel within the reactor vessel may be classified in accordance Recognition Category C during the Cold Shutdown and Refueling modes.

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 to 0CAN091801

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considered in conjunction with in-plant reports or observations of a potential fuel damaging event (e.g., a fuel handling accident). EAL #3Spent fuel pool water level at this value is within the lower end of the level range necessary to prevent significant dose consequences from direct gamma radiation to personnel performing operations in the vicinity of the spent fuel pool. This condition reflects a significant loss of spent fuel pool water inventory and thus it is also a precursor to a loss of the ability to adequately cool the irradiated fuel assembles stored in the pool. Escalation of the emergency classification level would be via IC s AS1 or AS2 (see AS2 Developer Notes

). Reference(s):

1. OP-1203.050 Unit 1 Spent Fuel Pool Emergencies 2. OP-1305.001 Radiation Monitoring System Check and Test
3. OP-2203.002 Spent Fuel Pool Emergencies
4. OP-1604.051 Eberline Radiation Monitoring System
5. OP-2304.133 Containment High Range Radiation Monitor Calibration 6. Offsite Dose Calculation Manual 7. NEI 99-01 AA2

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Category: A - Abnormal Rad Levels / Rad Effluent Subcategory: 2 - Irradiated Fuel Event

Initiating Condition: Significant lowering of water level above, or damage to, irradiated fuel EAL: AA2.3 Alert Lowering of spent fuel pool level to 387.0 ft. [389.5 ft.] (Alarm 2) on LIT-2020-3(4)

[2LIT-2020-1(2)]

Mode Applicability:

All Definition(s):

None Basis: This IC EAL 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 Developer 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 applies to irradiated fuel that is licensed for dry storage up to the point that the loaded storage cask is sealed. Once sealed, damage to a loaded cask causing loss of the CONFINEMENT BOUNDARY is classified in accordance with IC E-HU1.

Escalation of the emergency would be based on either Recognition Category A or C ICs. EAL #This EAL escalates from AU2 in that the loss of level, in the affected portion of the REFUELING PATHWAY, is of sufficient magnitude to have resulted in uncovery of irradiated fuel. Indications of irradiated fuel uncovery may include direct or indirect visual observation (e.g., reports from personnel or camera images), as well as significant changes in water and radiation levels, or other plant parameters. Computational aids may also be used (e.g., a boil-off curve). Classification of an event using this EAL should be based on the totality of available indications, reports and observations.

While an area radiation monitor could detect an increase in a dose rate due to a lowering of water level in some portion of the REFUELING PATHWAY, the reading may not be a reliable indication of whether or not the fuel is actually uncovered. To the degree possible, readings should be considered in combination with other available indications of inventory loss.

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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.

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 pool water level at this value is within the lower end of the level range necessary to prevent significant dose consequences from direct gamma radiation to personnel performing operations in the vicinity of the spent fuel pool. This condition reflects a significant loss of spent fuel pool water inventory and thus it is also a precursor to a loss of the ability to adequately cool the irradiated fuel assembles stored in the pool.

Escalation of the emergency classification level would be via IC s AS1 or AS2 (see AS2 Developer Notes

). Post-Fukushima order EA-12-051 required the installation of reliable SFP level indication capable of identifying normal level (Level 1), SFP level 10 ft. above the top of the fuel racks (Level 2) and SFP level at the top of the fuel racks (Level 3) (ref. 1, 2).

Reference(s):

1. MOHR-ANO-1, ANO-1 SFPI (Level) Configuration, Sheet 1, Revision 0 2. MOHR-ANO-2, ANO-2 SFPI (Level) Configuration, Sheet 1, Revision 0
3. NEI 99-01 AA2

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Category: A - Abnormal Rad Levels / Rad Effluent Subcategory: 2 - Irradiated Fuel Event

Initiating Condition: Spent fuel pool level at the top of the fuel racks EAL: AS2.1 Site Area Emergency Lowering of spent fuel pool level to 377.0 ft. [379.5 ft.] (Alarm 3) on LIT-2020-3(4)

[2LIT-2020-1(2)]

Mode Applicability:

All Definition(s):

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

Basis:

This IC EAL addresses a significant loss of spent fuel pool inventory control and makeup capability leading to IMMINENT fuel damage. This condition entails major failures of plant functions needed for protection of the public and thus warrant a Site Area Emergency declaration.

It is recognized that this IC would likely not be met until well after another Site Area Emergency IC was met; however, it is included to provide classification diversity.

Escalation of the emergency classification level would be via IC AG1 or AG2 A. Post-Fukushima order EA-12-051 required the installation of reliable SFP level indication capable of identifying normal level (Level 1), SFP level 10 ft. above the top of the fuel racks (Level 2) and SFP level at the top of the fuel racks (Level 3) (ref. 1, 2).

Reference(s):

1. MOHR-ANO-1, ANO-1 SFPI (Level) Configuration, Sheet 1, Revision 0 2. MOHR-ANO-2, ANO-2 SFPI (Level) Configuration, Sheet 1, Revision 0
3. NEI 99-01 AS2

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Category: A - Abnormal Rad Levels / Rad Effluent Subcategory: 2 - Irradiated Fuel Event

Initiating Condition: Spent fuel pool level cannot be restored to at least the top of the fuel racks for 60 minutes or longer

EAL: AG2.1 General Emergency Spent fuel pool level cannot be restored to at least 377.0 ft. [379.5 ft.] (Alarm 3) on LIT-2020-3(4) [2LIT-2020-1(2)] for 60 min. (Note 1)

Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded.

Mode Applicability:

All

Definition(s):

None Basis: This IC EAL addresses a significant loss of spent fuel pool inventory control and makeup capability leading to a prolonged uncovery of spent fuel. This condition will lead to fuel damage and a radiological release to the environment.

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

Post-Fukushima order EA-12-051 required the installation of reliable SFP level indication capable of identifying normal level (Level 1), SFP level 10 ft. above the top of the fuel racks (Level 2) and SFP level at the top of the fuel racks (Level 3) (ref. 1, 2).

Reference(s):

1. MOHR-ANO-1, ANO-1 SFPI (Level) Configuration, Sheet 1, Revision 0 2. MOHR-ANO-2, ANO-2 SFPI (Level) Configuration, Sheet 1, Revision 0 3. NEI 99-01 AG2 to 0CAN091801

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Category: A - Abnormal Rad Levels / Rad Effluent Subcategory: 3 - Area Radiation Levels

Initiating Condition: Radiation levels that IMPEDE access to equipment necessary for normal plant operations, cooldown or shutdown

EAL: AA3.1 Alert Dose rate > 15 mR/hr in EITHER of the following areas: Control Room Central Alarm Station (by survey)

Mode Applicability:

All Definition(s):

IMPEDE(D) - Personnel access to a room or area is hindered to an extent that extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., requiring use of protective equipment, such as SCBAs, that is not routinely employed).

Basis: Areas that meet this threshold include the Control Room (CR) and the Central Alarm Station (CAS). The Control Room envelope (Unit 1 and Unit 2) is monitored for excessive radiation by five detectors. These radiation detectors are RE-8001, 2RE-8001A, 2RE-8001B, 2RE-8750-1A, and 2RE-8750-1B (ref. 1). The CAS is included in this EAL because of its importance to permitting access to areas required to assure safe plant operations. There are no permanently installed area radiation monitors in CAS that may be used to assess this EAL threshold. Therefore, this threshold is evaluated using local radiation survey for this area.

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 should consider the cause of the increased rise in radiation levels and determine if another IC may be applicable. For EAL #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 to 0CAN091801

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affected room/area (e.g., installing temporary shielding, requiring use of non-routine protective equipment, requesting an extension in dose limits beyond normal administrative limits).

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

The plant is in an operating mode different than the mode specified for the affected room/area (i.e., entry is not required during the operating mode in effect at the time of the elevated radiation levels). For example, the plant is in Mode 1 when the radiation increase occurs, and the procedures used for normal operation, cooldown and shutdown do not require entry into the affected room until Mode 4.

The increased radiation levels are a result of a planned activity that includes compensatory measures which address the temporary inaccessibility of a room or area (e.g., radiography, spent filter or resin transfer, etc.).

The action for which room/area entry is required is of an administrative or record keeping nature (e.g., normal rounds or routine inspections).

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

Escalation of the emergency classification level would be via Recognition Category A, C or F

ICs. Reference(s):

1. STM 1-62 Radiation Monitoring 2. NEI 99-01 AA3 to 0CAN091801

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Category: A - Abnormal Rad Levels / Rad Effluent Subcategory: 3 - Area Radiation Levels

Initiating Condition: Radiation levels that IMPEDE access to equipment necessary for normal plant operations, cooldown or shutdown

EAL: AA3.2 Alert An UNPLANNED event results in radiation levels that prohibit or IMPEDE access to any Table 1[2]A-3 room or area (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 1A-3 Unit 1 Safe Operation & Shutdown Rooms/Areas Room/Area Mode A-4 Switchgear Room 3, 4 Upper North Electrical Penetration Room 3, 4 Lower South Electrical Equipment Room 3, 4 Table 2A-3 Unit 2 Safe Operation & Shutdown Rooms/Areas Room/Area Mode Aux Building 317' Emergency Core Cooling Rooms 3, 4 Aux Building 317' Tendon Gallery Access 3, 4 Aux Building 335' Charging Pumps / MCC 2B-52 3, 4 Aux Building 354' MCC 2B-62 Area 3, 4 Emergency Diesel Generator Corridor 3, 4 Lower South Piping Penetration Room 3, 4 Aux Building 386' Containment Hatch 3, 4 to 0CAN091801

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Mode Applicability:

3 - Hot Standby, 4 - Hot Shutdown Definition(s):

IMPEDE(D) - Personnel access to a room or area is hindered to an extent that extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., requiring use of protective equipment, such as SCBAs, that is not routinely employed).

UNPLANNED

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

Basis: 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 should consider the cause of the increased rise in radiation levels and determine if another IC may be applicable.

For EAL #2 A A3.2 , 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 higher radiation levels. Access should be considered as IMPEDED if extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., installing temporary shielding, requiring use of non-routine protective equipment, requesting an extension in dose limits beyond normal administrative limits).

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

The plant is in an operating mode different than the mode specified for the affected room/area (i.e., entry is not required during the operating mode in effect at the time of the elevated radiation levels). For example, the plant is in Mode 1 when the radiation increase rise occurs, and the procedures used for normal operation, cooldown and shutdown do not require entry into the affected room until Mode

43. The increased higher radiation levels are a result of a planned activity that includes compensatory measures which address the tem porary 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. to 0CAN091801

Page 69 of 257 Attachment 1 - Emergency Action Level Technical Bases

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

If the equipment in the listed room or area was already inoperable, or out-of-service, before the event occurred, then no emergency should be declared since the event will have no adverse impact beyond that already allowed by Technical Specifications at the time of the event.

The list of plant rooms or areas with entry-related mode applicability identified specify those rooms or areas that contain equipment which require a manual/local action as specified in operating procedures used for normal plant operation, cooldown and shutdown.

Rooms or areas in which actions of a contingent or emergency nature would be performed (e.g., an action to address an off-normal or emergency condition such as emergency repairs, corrective measures or emergency operations) are not included.

In addition, the list specifies the plant mode(s) during which entry would be required for each room or area (ref. 1).

EAL AA3.2 mode applicability has been limited to the mode limitations of Table 1[2]A-3 (Modes 3 and 4 only). Reference(s):

1. Attachment 2 Safe Operation & Shutdown Areas Tables 1[2]A-3 & 1[2]H-2 Bases 2. NEI 99-01 AA3 to 0CAN091801

Page 70 of 257 Attachment 1 - Emergency Action Level Technical Bases

Category C - Cold Shutdown / Refueling System Malfunction EAL Group: Cold Conditions (RCS temperature 200°F); EALs in this category are applicable only in one or more cold operating modes.

Category C EALs are directly associated with cold shutdown or refueling system safety functions. Given the variability of plant configurations (e.g., systems out-of-service for maintenance, containment open, reduced AC power redundancy, time since shutdown) during these periods, the consequences of any given init iating event can vary greatly. For example, a loss of decay heat removal capability that occurs at the end of an extended outage has less significance than a similar loss occurring during the first week after shutdown. Compounding these events is the likelihood that instrumentation necessary for assessment may also be inoperable. The cold shutdown and refueling system malfunction EALs are based on performance capability to the extent possible with consideration given to RCS integrity, CONTAINMENT CLOSURE, and fuel clad integrity for the applicable operating modes (5 - Cold Shutdown, 6 - Refueling, DEF - Defueled).

The events of this category pertain to the following subcategories:

1. RCS Level RCS water level is directly related to the status of adequate core cooling and, therefore, fuel clad integrity.
2. Loss of Vital AC Power Loss of vital 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 vital buses.
3. RCS Temperature Uncontrolled or inadvertent temperature or pressure rises are indicative of a potential loss of

safety functions.

4. Loss of Vital DC Power Loss of emergency plant electrical power can compromise plant SAFETY SYSTEM operability including decay heat removal and emergency core cooling systems which may be necessary to ensure fission product barrier integrity. This category includes loss of power to or degraded voltage on the 125V DC vital buses.
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. to 0CAN091801

Page 71 of 257 Attachment 1 - Emergency Action Level Technical Bases

Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - RCS Level

Initiating Condition: UNPLANNED loss of RCS inventory EAL: CU1.1 Unusual Event UNPLANNED loss of reactor coolant results in RCS water level less than a required lower limit for 15 min. (Note 1)

Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded.

Mode Applicability:

5 - Cold Shutdown, 6 - Refueling

Definition(s):

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

Basis: This IC EAL 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 (reactor vessel/

RCS [PWR] or RPV [BWR]) 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 decreaselower RCS water inventory are carefully planned and controlled. An UNPLANNED event that results in water level decreasinglowering below a procedurally required limit warrants the declaration of an Unusual Event due to the reduced water inventory that is available to keep the core covered.

This EAL #1 recognizes that the minimum required (reactor vessel/

RCS [PWR] or RPV [

BWR]) 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.

to 0CAN091801

Page 72 of 257 Attachment 1 - Emergency Action Level Technical Bases

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 addresses a condition where all means to determine (reactor vessel/RCS [

PWR] or RPV [BWR]) 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 vessel/RCS [

PWR] or RPV [

BWR]). Continued loss of RCS inventory may result in escalation to the Alert emergency classification level via either IC CA1 or CA3.

Reference(s):

1. OP-1015.002 Decay Heat Removal and LTOP System 2. OP-1015.008 Unit 2 SDC Control
3. NEI 99-01 CU1 to 0CAN091801

Page 73 of 257 Attachment 1 - Emergency Action Level Technical Bases

Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - RCS Level

Initiating Condition: UNPLANNED loss of RCS inventory EAL: CU1.2 Unusual Event RCS level cannot be monitored AND EITHER:

UNPLANNED rise in any Table 1[2]C-1 sump/tank level due to loss of RCS inventory Visual observation of UNISOLABLE RCS leakage Table 1C-1 Unit 1 Sumps / Tanks Reactor Building Sump Reactor Drain Tank Aux. Building Equipment Drain Tank Aux. Building Sump Quench Tank Table 2C-1 Unit 2 Sumps / Tanks CNTMT Sump Reactor Drain Tank LRW Waste Tank (2T-20) Holdup Tank Aux. Building Sump Quench Tank Mode Applicability:

5 - Cold Shutdown, 6 - Refueling to 0CAN091801

Page 74 of 257 Attachment 1 - Emergency Action Level Technical Bases

Definition(s):

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

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

Basis: 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 (reactor vessel/

RCS [PWR] or RPV [BWR]) 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 lower RCS water inventory are carefully planned and controlled. An UNPLANNED event that results in water level decreasing lowering 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 recognizes that the minimum required (reactor vessel/RCS [PWR] or RPV [

BWR]) 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.

This EAL #2 addresses a condition where all means to determine (reactor vessel/

RCS [PWR] or RPV [BWR]) level have been lost. In this condition, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels (Table C-1). Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the (reactor vessel/

RCS [PWR] or RPV [BWR])

. Continued loss of RCS inventory may result in escalation to the Alert emergency classification level via either IC CA1 or CA3.

Reference(s):

1. OP-1203.039 Excess RCS Leakage 2. OP-2203.016 Excess RCS Leakage
3. NEI 99-01 CU1 to 0CAN091801

Page 75 of 257 Attachment 1 - Emergency Action Level Technical Bases

Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - RCS Level

Initiating Condition: Significant Loss of RCS inventory EAL: CA1.1 Alert Loss of RCS inventory as indicated by EITHER: RVLMS Levels 1 through 8 [1 through 5] indicate DRY Reactor vessel level < 370.2 ft. (LT-1195/LT-1196) [< 24 in. (L4791/L4792)] (minimum level for DHR operation @ 1000 gpm)[(minimum level for SDC operation)]

Mode Applicability:

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

None

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 #1 , a lowering of RPV water level below (site-specific level) ft the specified level indicates that operator actions have not been successful in restoring and maintaining RCS (reactor vessel/RCS [PWR] or RPV [BWR]) water level. The heat-up rate of the coolant will increase rise as the available water inventory is reduced. A continuing decrease drop in water level will lead to core uncovery.

Although related, EAL

  1. 1 is concerned with the loss of RPV inventory and not the potential concurrent effects on systems needed for decay heat removal (e.g., loss of a Residual Decay Heat Removal suction point). A n increase rise in RCS temperature caused by a loss of decay heat removal capability is evaluated under IC CA3.

For EAL #2, the inability to monitor (reactor vessel/RCS [PWR] or RPV [BWR]) 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 (reactor vessel/RCS [PWR] or RPV [BWR]). to 0CAN091801

Page 76 of 257 Attachment 1 - Emergency Action Level Technical Bases

The 15-minute duration for the loss of level indication was chosen because it is half of the EAL duration specified in IC CS1 If RCS the (reactor vessel/RCS [PWR] or RPV [BWR]) inventory water level continues to lower, then escalation to Site Area Emergency would be via IC CS1.

A loss of DHR/SDC will occur at approximately RLVMS Level 8 (Unit 1) or RVLMS Level 5 (Unit 2). However, RVLMS may not be available in the cold shutdown modes. Redundant means of level indication is provided in these modes and included in this EAL. The point at which a loss of DHR/SDC is likely to occur is 370.2 ft. (Unit 1) or 24 in. (Unit 2) as indicated in the respective Control Rooms. The value selected for ANO-1 is based on 1000 gpm DHR flow which is the flow rate at which the low flow alarm is received. The ANO-2 value is the proceduralized minimum value. Below these levels, a loss of suction to decay heat removal systems will occur (ref. 1, 2, 3). The inability to restore and maintain level after reaching this value would be indicative of a failure of the RCS barrier.

Reference(s):

1. OP-1104.004 Decay Heat Removal Operating Procedure 2. OP-1105.008 Inadequate Core Cooling Monitor and Display
3. OP-2105.003 Reactor Vessel Level Monitoring System Operations
4. OP-2203.029 Loss of Shutdown Cooling 5. Calculation No. 90-E-0116-01 ANO-2 EOP Setpoint Basis Document, Setpoints R.3 and R.9 6. NEI 99-01 CA1

to 0CAN091801

Page 77 of 257 Attachment 1 - Emergency Action Level Technical Bases

Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - RCS Level

Initiating Condition: Significant Loss of RCS inventory EAL: CA1.2 Alert RCS level cannot be monitored for 15 min. (Note 1) AND EITHER:

UNPLANNED rise in any Table 1[2]C-1 Sump / Tank level due to a loss of RCS inventory Visual observation of UNISOLABLE RCS leakage Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded.

Table 1C-1 Unit 1 Sumps / Tanks Reactor Building Sump Reactor Drain Tank Aux. Building Equipment Drain Tank Aux. Building Sump Quench Tank Table 2C-1 Unit 2 Sumps / Tanks CNTMT Sump Reactor Drain Tank LRW Waste Tank (2T-20) Holdup Tank Aux. Building Sump Quench Tank to 0CAN091801

Page 78 of 257 Attachment 1 - Emergency Action Level Technical Bases

Mode Applicability:

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

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

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

Basis: 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 level below (site-specific level) indicates that operator actions have not been successful in restoring and maintaining (reactor vessel/RCS [

PWR] or RPV [BWR]) water level. The heat-up rate of the coolant will increase as the available water inventory is reduced. A continuing decrease in water level will lead to core uncovery.

Although related, EAL #1 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 CA3.

For this EAL #2, the inability to monitor RCS (reactor vessel/RCS [PWR] or RPV [BWR])

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 (reactor vessel/

RCS [PWR] or RPV [BWR])

. The 15-minute duration for the loss of level indication was chosen because it is half of the EAL duration specified in IC CS1

. If the (reactor vessel/

RCS [PWR] or RPV

[BWR]) inventory level continues to lower, then escalation to Site Area Emergency would be via IC CS1.

Reference(s):

1. OP-1203.039 Excess RCS Leakage 2. OP-2203.016 Excess RCS Leakage 3. NEI 99-01 CA1 to 0CAN091801

Page 79 of 257 Attachment 1 - Emergency Action Level Technical Bases

Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - RCS Level

Initiating Condition: Loss of RCS inventory affecting core decay heat removal capability EAL: CS1.1 Site Area Emergency CONTAINMENT CLOSURE not established AND RVLMS Levels 1 through 9 [1 through 6] indicate DRY Mode Applicability:

5 - Cold Shutdown, 6 - Refueling

Definition(s):

CONTAINMENT CLOSURE - The action to secure primary containment and its associated structures, systems, and components as a functional barrier to fission product release under existing plant conditions.

As applied to ANO, CONTAINMENT CLOSURE must be capable of being set within 30 minutes. CONTAINMENT CLOSURE is set when the penetrations are isolated by manual or automatic isolation valve, blind flange, or equivalent.

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

Basis: This IC addresses a significant and prolonged loss of (reactor vessel/RCS [PWR] or RPV

[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 RCS/reactor vessel level cannot be restored, fuel damage is probable.

Outage/shutdown contingency plans typically provide for re-establishing or verifying CONTAINMENT CLOSURE following a loss of heat removal or RCS inventory control functions.

The difference in the specified RCS/reactor vessel levels of EALs 1.b and 2.b reflect the fact to 0CAN091801

Page 80 of 257 Attachment 1 - Emergency Action Level Technical Bases

that with CONTAINMENT CLOSURE established, there is a lower probability of a fission product release to the environment.

In EAL 3.a, the 30-minute criterion is tied to a readily recognizable event start time (i.e., the total loss of ability to monitor level), and allows sufficient time to monitor, assess and correlate reactor and plant conditions to determine if core uncovery has actually occurred (i.e., to account for various accident progression and instrumentation uncertainties). It also allows sufficient time for performance of actions to terminate leakage, recover inventory control/makeup equipment and/or restore level monitoring.

The inability to monitor (reactor vessel/RCS [PWR] or RPV [BWR]) 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 (reactor vessel/RCS [PWR] or RPV [BWR]).

Th isese EAL s address es concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal; SECY 91-283, Evaluation of Shutdown and Low Power Risk Issues; NUREG-1449, Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the United States

and NUMARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management.

Escalation of the emergency classification level would be via IC CG1 or AG1.

Reference(s):

1. OP-1105.008 Inadequate Core Cooling Monitor and Display 2. OP-2105.003 Reactor Vessel Level Monitoring System Operations
3. NEI 99-01 CS1 to 0CAN091801

Page 81 of 257 Attachment 1 - Emergency Action Level Technical Bases

Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - RCS Level

Initiating Condition: Loss of RCS inventory affecting core decay heat removal capability EAL: CS1.2 Site Area Emergency [RVLMS Levels 1 through 7 indicate DRY OR] RCS level cannot be monitored for 30 min. (Note 1) AND Core uncovery is indicated by any of the following: UNPLANNED rise in any Table 1[2]C-1 sump/tank level of sufficient magnitude to indicate core uncovery Containment high range radiation monitor RE-8060/8061 [2RE-8925-1/8925-2] reading

> 10 R/hr Erratic Source Range Monitor indication Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded.

Table 1C-1 Unit 1 Sumps / Tanks Reactor Building Sump Reactor Drain Tank Aux. Building Equipment Drain Tank Aux. Building Sump Quench Tank to 0CAN091801

Page 82 of 257 Attachment 1 - Emergency Action Level Technical Bases

Table 2C-1 Unit 2 Sumps / Tanks CNTMT Sump Reactor Drain Tank LRW Waste Tank (2T-20) Holdup Tank Aux. Building Sump Quench Tank Mode Applicability:

5 - Cold Shutdown, 6 - Refueling

Definition(s):

CONTAINMENT CLOSURE - The action to secure primary containment and its associated structures, systems, and components as a functional barrier to fission product release under existing plant conditions.

As applied to ANO, CONTAINMENT CLOSURE must be capable of being set within 30 minutes. CONTAINMENT CLOSURE is set when the penetrations are isolated by manual or automatic isolation valve, blind flange, or equivalent.

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.

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

Basis: When in service, the Unit 2 RVLMS can measure RCS level below the top of active fuel. Level 7 DRY on this system is an indication of core uncovery.

This IC addresses a significant and prolonged loss of (reactor vessel/RCS RCS [PWR] or RPV

[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.

to 0CAN091801

Page 83 of 257 Attachment 1 - Emergency Action Level Technical Bases

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

Outage/shutdown contingency plans typically provide for re-establishing or verifying CONTAINMENT CLOSURE following a loss of heat removal or RCS inventory control functions.

The difference in the specified RCS

/reactor vessel levels of EALs 1.b and 2.b CS1.1 and CS1.2 reflect the fact that with CONTAINMENT CLOSURE established, there is a lower probability of a fission product release to the environment.

In EAL 3.a, 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 (reactor vessel/

RCS [PWR] or RPV [BWR]) 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 (reactor vessel/

RCS [PWR] or RPV [BWR])

. These This EAL s address es 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. Containment High Range Radiation Monitors RE-8060/8061 [2RE-8925-1/8925-2] are the site-specific radiation monitors that would be indicative of possible core uncovery in the Refueling mode. The dose rate due to core shine when the top of the core becomes uncovered should result in dose rates > 10 R/hr. Escalation of the emergency classification level would be via IC CG1 or AG1.

Reference(s):

1. OP-1203.039 Excess RCS Leakage 2. OP-2203.016 Excess RCS Leakage 3. OP-2105.003 Reactor Vessel Level Monitoring System Operations 4. 1SAR Table 7-11
5. 2SAR 12.1.4.2
6. NEI 99-01 CS1 to 0CAN091801

Page 84 of 257 Attachment 1 - Emergency Action Level Technical Bases

Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - RCS Level

Initiating Condition: Loss of RCS inventory affecting fuel clad integrity with containment challenged

EAL: CG1.1 General Emergency - UNIT 2 ONLY RVLMS Levels 1 through 7 indicate DRY AND Any Containment Challenge indication, Table 1[2]C-2 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 1[2]C-2 Containment Challenge Indications CONTAINMENT CLOSURE not established (Note 6) Containment hydrogen concentration > 4% UNPLANNED rise in containment pressure Mode Applicability:

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

CONTAINMENT CLOSURE - The action to secure primary containment and its associated structures, systems, and components as a functional barrier to fission product release under existing plant conditions.

As applied to ANO, CONTAINMENT CLOSURE must be capable of being set within 30 minutes. CONTAINMENT CLOSURE is set when the penetrations are isolated by manual or automatic isolation valve, blind flange, or equivalent.

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.

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. to 0CAN091801

Page 85 of 257 Attachment 1 - Emergency Action Level Technical Bases

Basis: When in service, the Unit 2 RVLMS can measure RCS level below the top of active fuel. Level 7 DRY on this system is an indication of core uncovery.

This IC addresses the inability to restore and maintain reactor vessel level above the top of active fuel with containment challenged. This condition represents actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity. Releases can be reasonably expected to exceed EPA PAG exposure levels offsite for more than the immediate site area.

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

With CONTAINMENT CLOSURE not established, there is a high potential for a direct and unmonitored release of radioactivity to the environment. If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, then declaration of a General Emergency is not required.

The existence of an explosive mixture means, at a minimum, that the containment atmospheric hydrogen concentration is sufficient to support a hydrogen burn (i.e., at the lower deflagration limit). A hydrogen burn will raise containment pressure and could result in collateral equipment damage leading to a loss of containment integrity. It therefore represents a challenge to Containment integrity.

In the early stages of a core uncovery event, it is unlikely that hydrogen buildup due to a core uncovery could result in an explosive gas mixture in containment. If all installed hydrogen gas monitors are out-of-service during an event leading to fuel cladding damage, it may not be possible to obtain a containment hydrogen gas concentration reading as ambient conditions within the containment will preclude personnel access. During periods when installed containment hydrogen gas monitors are out-of-service, operators may use the other listed indications to assess whether or not containment is challenged.

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 (reactor vessel/RCS [PWR] or RPV [

BWR]) 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 (reactor vessel/RCS [

PWR] or RPV [

BWR]). to 0CAN091801

Page 86 of 257 Attachment 1 - Emergency Action Level Technical Bases

This EAL addresses concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal

SECY 91-283, Evaluation of Shutdown and Low Power Risk Issues; NUREG-1449, Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the United States
and NUMARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management. Reference(s)
1. OP-1203.039 Excess RCS Leakage 2. OP-2203.016 Excess RCS Leakage
3. OP-2105.003 Reactor Vessel Level Monitoring System Operations
4. 1SAR Table 7-11 5. 2SAR 12.1.4.2 6. Unit 1 SAMG Figure III-1B
7. Unit 2 SAMG Phase 1 Instructions, Containment Flowchart
8. NEI 99-01 CG1

to 0CAN091801

Page 87 of 257 Attachment 1 - Emergency Action Level Technical Bases

Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - RCS Level

Initiating Condition: Loss of RCS inventory affecting fuel clad integrity with containment challenged

EAL: CG1.2 General Emergency RCS level cannot be monitored for 30 min. (Note 1) AND Core uncovery is indicated by any of the following: UNPLANNED rise in any Table 1[2]C-1 sump/tank level of sufficient magnitude to indicate core uncovery Containment High Range Radiation Monitor RE-8060/8061 [2RE-8925-1/8925-2] reading > 10 R/hr Erratic Source Range Monitor indication AND Any Containment Challenge indication, Table 1[2]C-2 Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is 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 1C-1 Unit 1 Sumps / Tanks Reactor Building Sump Reactor Drain Tank Aux. Building Equipment Drain Tank Aux. Building Sump Quench Tank to 0CAN091801

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Table 2C-1 Unit 2 Sumps / Tanks CNTMT Sump Reactor Drain Tank LRW Waste Tank (2T-20) Holdup Tank Aux. Building Sump Quench Tank Table 1[2]C-2 Containment Challenge Indications CONTAINMENT CLOSURE not established (Note 6) Containment hydrogen concentration > 4% UNPLANNED rise in containment pressure Mode Applicability:

5 - Cold Shutdown, 6 - Refueling

Definition(s):

CONTAINMENT CLOSURE - The action to secure primary containment and its associated structures, systems, and components as a functional barrier to fission product release under existing plant conditions.

As applied to ANO, CONTAINMENT CLOSURE must be capable of being set within 30 minutes. CONTAINMENT CLOSURE is set when the penetrations are isolated by manual or automatic isolation valve, blind flange, or equivalent.

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.

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

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Basis:

When in service, the Unit 2 RVLMS can measure RCS level below the top of active fuel. Level 7 DRY on this system is an indication of core uncovery.

This IC addresses the inability to restore and maintain reactor vessel level above the top of active fuel with containment challenged. This condition represents actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity. Releases can be reasonably expected to exceed EPA PAG exposure levels offsite for more than the immediate site area.

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

With CONTAINMENT CLOSURE not established, there is a high potential for a direct and unmonitored release of radioactivity to the environment. If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, then declaration of a General Emergency is not required.

The existence of an explosive mixture means, at a minimum, that the containment atmospheric hydrogen concentration is sufficient to support a hydrogen burn (i.e., at the lower deflagration limit). A hydrogen burn will raise containment pressure and could result in collateral equipment damage leading to a loss of containment integrity. It therefore represents a challenge to Containment integrity.

In the early stages of a core uncovery event, it is unlikely that hydrogen buildup due to a core uncovery could result in an explosive gas mixture in containment. If all installed hydrogen gas monitors are out-of-service during an event leading to fuel cladding damage, it may not be possible to obtain a containment hydrogen gas concentration reading as ambient conditions within the containment will preclude personnel access. During periods when installed containment hydrogen gas monitors are out-of-service, operators may use the other listed indications to assess whether or not containment is challenged.

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 (reactor vessel/

RCS [PWR] or RPV [

BWR]) 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 (reactor vessel/

RCS [PWR] or RPV [

BWR]). to 0CAN091801

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Containment High Range Radiation Monitor RE-8060/8061 [2RE-8925-1/8925-2] are the site-specific radiation monitors that would be indicative of possible core uncovery in the Refueling mode. The dose rate due to core shine when the top of the core becomes uncovered should result in dose rates > 10 R/hr. This EAL addresses concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal

SECY 91-283, Evaluation of Shutdown and Low Power Risk Issues; NUREG-1449, Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the United States
and NUMARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management. Reference(s)
1. OP-1203.039 Excess RCS Leakage 2. OP-2203.016 Excess RCS Leakage
3. OP-2105.003 Reactor Vessel Level Monitoring System Operations
4. 1SAR Table 7-11
5. 2SAR 12.1.4.2
6. Unit 1 SAMG Figure III-1B 7. Unit 2 SAMG Phase 1 Instructions, Containment Flowchart 8. NEI 99-01 CG1

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Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 2 - Loss of Vital AC Power

Initiating Condition: Loss of all but one AC power source to vital buses for 15 minutes or longer EAL: CU2.1 Unusual Event AC power capability, Table 1[2]C-3, to vital 4.16 KV buses A3 [2A3] and A4 [2A4] reduced to a single power source for 15 min. (Note 1) AND Any additional single power source failure will result in loss of all AC power to SAFETY SYSTEMS Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded.

Table 1C-3 Unit 1 AC Power Sources Offsite Startup Transformer No. 1 Startup Transformer No. 2 Unit Auxiliary Transformer (from 22 KV switchyard)

Onsite DG1 DG2 AAC Gen to 0CAN091801

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Table 2C-3 Unit 2 AC Power Sources Offsite Startup Transformer No. 3 Startup Transformer No. 2 Unit Auxiliary Transformer (backfed from main transformer)

Onsite 2DG1 2DG2 AAC Gen Mode Applicability:

5 - Cold Shutdown, 6 - Refueling, DEF - 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 10 CFR 50.2):

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: 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 greater 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. to 0CAN091801

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An "AC power source" is a source recognized in AOPs and EOPs, and capable of supplying required power to a vital bus. Some examples of this condition are presented below.

A loss of all offsite power with a concurrent failure of all but one emergency vital power source (e.g., an onsite diesel generator). A loss of all offsite power and loss of all emergency vital power sources (e.g., onsite diesel generators) with a single train of emergency vital buses being back-fed from the unit main generator. A loss of emergency vital power sources (e.g., onsite diesel generators) with a single train of emergency vital 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.

This EAL is the cold condition equivalent of the hot condition EAL SA1.1.

Reference(s):

1. 1SAR Figure 8-1 Station Single Line Diagram 2. OP-1202.007 Degraded Power
3. OP-1202.008 Blackout 4. OP-2104.037 Alternate AC Diesel Generator Operations 5. 2SAR Figure 8.3-1 Station Single Line Diagram
6. OP-2202.007 Loss of Off-Site Power
7. OP-2202.008 Station Blackout
8. OP-2107.006 Backfeed of Unit Auxiliary Transformer 9. NEI 99-01 CU2

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Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 2 - Loss of Vital AC Power

Initiating Condition: Loss of all offsite and all onsite AC power to vital buses for 15 minutes or longer

EAL: CA2.1 Alert Loss of all offsite and all onsite AC power to vital 4.16 KV buses A3 [2A3] and A4 [2A4] for 15 min. (Note 1)

Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded.

Mode Applicability:

5 - Cold Shutdown, 6 - Refueling, DEF - 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 10 CFR 50.2):

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:

Although the AAC may be considered available, it will not prevent declaration of this EAL unless it is powering a vital bus within the 15-minute time period of the EAL. 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. Mitigative strategies using non-safety related power sources (FLEX generators, etc.) may be effective in supplying power to these buses. These power sources must be controlled in to 0CAN091801

Page 95 of 257 Attachment 1 - Emergency Action Level Technical Bases

accordance with abnormal or emergency operating procedures, or beyond design basis accident response guidelines (e.g., FLEX support guidelines) and must be capable (alone or in combination) of supplying power for long term decay heat removal systems.

When in the cold shutdown, refueling, or defueled mode, this condition is not classified as a Site Area Emergency because of the increased greater time available to restore a n emergency vital bus to service. Additional time is available due to the reduced core decay heat load, and the lower temperatures and pressures in various plant systems. Thus, when in these modes, this condition represents an actual or potential substantial degradation of the level of safety of the plant.

Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.

Escalation of the emergency classification level would be via IC CS1 or AS1.

This EAL is the cold condition equivalent of the hot condition EAL SS1.1.

Reference(s):

1. 1SAR Figure 8-1 Station Single Line Diagram 2. OP-1202.007 Degraded Power
3. OP-1202.008 Blackout
4. OP-2104.037 Alternate AC Diesel Generator Operations 5. 2SAR Figure 8.3-1 Station Single Line Diagram 6. OP-2202.007 Loss of Off-Site Power
7. OP-2202.008 Station Blackout
8. OP-2107.006 Backfeed of Unit Auxiliary Transformer
9. NEI 99-01 CU2 to 0CAN091801

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Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 3 - RCS Temperature

Initiating Condition: UNPLANNED rise in RCS temperature EAL: CU3.1 Unusual Event UNPLANNED rise in RCS temperature to > 200°F Mode Applicability:

5 - Cold Shutdown, 6 - Refueling°

Definition(s):

CONTAINMENT CLOSURE - The action to secure primary containment and its associated structures, systems, and components as a functional barrier to fission product release under existing plant conditions.

As applied to ANO, CONTAINMENT CLOSURE must be capable of being set within 30 minutes. CONTAINMENT CLOSURE is set when the penetrations are isolated by manual or automatic isolation valve, blind flange, or equivalent.

UNPLANNED -

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

Basis: This IC addresses an UNPLANNED increase rise in RCS temperature above the Technical Specification cold shutdown temperature limit or the inability to determine RCS temperature and level,and represents a potential degradation of the level of safety of the plant. If the RCS is not intact and CONTAINMENT CLOSURE is not established during this event, the Emergency Director should also refer to IC EAL CA3.1. 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 #1This EALThis 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.

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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 loweredreduced inventory may result in a rapid increase rise in reactor coolant temperature depending on the time after shutdown.

EAL #2 reflects a condition where there has been a significant loss of instrumentation capability necessary to monitor RCS conditions and operators would be unable to monitor key parameters necessary to assure core decay heat removal. During this condition, there is no immediate threat of fuel damage because the core decay heat load has been reduced since the cessation of power operation.

Fifteen minutes was selected as a threshold to exclude transient or momentary losses of indication.

Escalation to Alert would be via IC CA1 based on an inventory loss or IC CA3 based on exceeding plant configuration-specific time criteria.

Reference(s):

1. Unit 1 and Unit 2 Technical Specifications Table 1.1-1 2. NEI 99-01 CU3 to 0CAN091801

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Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 3 - RCS Temperature

Initiating Condition: UNPLANNED rise in RCS temperature EAL: CU3.2 Unusual Event Loss of all RCS temperature and RCS level indication for 15 min. (Note 1)

Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded.

Mode Applicability:

5 - Cold Shutdown, 6 - Refueling

Definition(s):

CONTAINMENT CLOSURE - The action to secure primary containment and its associated structures, systems, and components as a functional barrier to fission product release under existing plant conditions.

As applied to ANO, CONTAINMENT CLOSURE must be capable of being set within 30 minutes. CONTAINMENT CLOSURE is set when the penetrations are isolated by manual or automatic isolation valve, blind flange, or equivalent.

UNPLANNED -

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

Basis: This IC EALEAL addresses an UNPLANNED increase in RCS temperature above the Technical Specification cold shutdown temperature limit, or the inability to determine RCS temperature and level, andand 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 should also refer to IC EAL CA3.1. 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 #1 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 to 0CAN091801

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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 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.

EAL #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 parameters necessary to assure core decay heat removal. During this condition, there is no immediate threat of fuel damage because the core decay heat load has been reduced since the cessation of power operation.

Fifteen minutes was selected as a threshold to exclude transient or momentary losses of indication.

Escalation to Alert would be via IC CA1 based on an inventory loss or IC CA3 based on exceeding plant configuration-specific time criteria.

Reference(s):

1. NEI 99-01 CU3

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Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 3 - RCS Temperature

Initiating Condition: Inability to maintain plant in cold shutdown EAL: CA3.1 Alert UNPLANNED rise in RCS temperature to > 200°F for > Table 1[2]C-4 duration (Note 1)

OR UNPLANNED RCS pressure rise > 10 psig (this EAL does not apply during water-solid plant conditions)

Note 1: The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded.

Table 1[2]C-4 RCS Heat-up Duration Thresholds RCS Status CONTAINMENT CLOSURE Status Heat-up Duration Intact (but not lowered inventory) N/A 60 min.*

Not intact OR lowered inventory established 20 min.*

not established 0 min.

  • If an RCS heat removal system is in operation within this time frame and RCS temperature is being reduced, the EAL is not applicable.

Mode Applicability:

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

CONTAINMENT CLOSURE - The action to secure primary containment and its associated structures, systems, and components as a functional barrier to fission product release under existing plant conditions.

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As applied to ANO, CONTAINMENT CLOSURE must be capable of being set within 30 minutes. CONTAINMENT CLOSURE is set when the penetrations are isolated by manual or automatic isolation valve, blind flange, or equivalent.

UNPLANNED -

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

Basis: In the absence of reliable RCS temperature indication, classification should be based on the RCS pressure rise criteria when the RCS is intact in Mode 5 or based on time to boil data when in Mode 6 or the RCS is not intact in Mode 5.

This IC EAL 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 increasea rise in RCS temperature when CONTAINMENT CLOSURE is established but the RCS is not intact, or RCS inventory is reduced (e.g., lowered inventory operationmid-loop operation in PWRs

). The 20-minute criterion was included to allow time for operator action to address the temperature increaserise. The RCS Heat-up Duration Thresholds table also addresses an increasea rise 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 rise without a substantial degradation in plant safety.

Finally, in the case where there is a n increase rise in RCS temperature, the RCS is not intact or is at loweredreduced inventory

[PWR], and CONTAINMENT CLOSURE is not established, no heat-up duration is allowed (i.e., 0 minutes). This is because 1) the evaporated reactor coolant may be released directly into the containment atmosphere and subsequently to the environment, and 2) there is reduced reactor coolant inventory above the top of irradiated fuel.

EAL #2The RCS pressure rise 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 AS1.

Reference(s):

1. Unit 1 and Unit 2 Technical Specifications Table 1.1-1 2. NEI 99-01 CA3 to 0CAN091801

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Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 4 - Loss of Vital DC Power

Initiating Condition: Loss of Vital DC power for 15 minutes or longer EAL: CU4.1 Unusual Event Indicated voltage is < 105 VDC on vital 125 VDC buses for 15 min. (Note 1)

Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded.

Mode Applicability:

5 - Cold Shutdown, 6 - Refueling

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 10 CFR 50.2).

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: Unit 1 batteries D06 and D07 and Unit 2 batteries 2D11 and 2D12 contain 58 cells each with a minimum cell voltage of 1.81 V or 105 VDC.

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 raise 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.

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As used in this EAL, "required" means the vital DC buses necessary to support operation of the in-service, or operable, train or trains of SAFETY SYSTEM equipment. For example, if Train A is out-of-service (inoperable) for scheduled outage maintenance work and Train B is in-service (operable), then a loss of Vital DC power affecting Train B would require the declaration of an Unusual Event. A loss of Vital DC power to Train A would not warrant an emergency classification.

Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.

Depending upon the event, escalation of the emergency classification level would be via IC CA1 or CA3, or an IC in Recognition Category A.

This EAL is the cold condition equivalent of the hot condition EAL SS2.1.

Reference(s):

1. 1SAR 8.3.2.1.1 Batteries 2. 2SAR 8.3.2.1.1 Batteries 3. NEI 99-01 CU4

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Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 5 - Loss of Communications

Initiating Condition: Loss of all onsite or offsite communications capabilities EAL: CU5.1 Unusual Event Loss of all Table 1[2]C-5 onsite communication methods OR Loss of all Table 1[2]C-5 State and local agency communication methods OR Loss of all Table 1[2]C-5 NRC communication methods Table 1[2]C-5 Communication Methods System Onsite ORO NRC Station radio system X ANO plant phone system X Gaitronics X Telephone Systems: Commercial Microwave Satellite VOIP X X Emergency Notification System (ENS) X Mode Applicability:

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

None to 0CAN091801

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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 State and local agencieORO s 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 #1The 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 State and local agenciesOROs of an emergency declaration. The State and local agenciesOROs referred to here are the Arkansas Department of Health, Arkansas Department of Emergency Management, Pope, Yell, Johnson, and Logan County offsite agencies (see Developer Notes). EAL #3The third EAL addresses a total loss of the communications methods used to notify the NRC of an emergency declaration.

This EAL is the cold condition equivalent of the hot condition EAL SU7.1.

Reference(s):

1. OP-1903.062 Communications System Operating Procedure 2. NEI 99-01 CU5 to 0CAN091801

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Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 6 - Hazardous Event Affecting Safety Systems

Initiating Condition: Hazardous event affecting SAFETY SYSTEMS needed for the current operating mode

EAL: CA6.1 Alert The occurrence of any Table 1[2]C-6 hazardous event AND Event damage has caused indications of degraded performance on one train of a SAFETY SYSTEM needed for the current operating mode AND EITHER:

Event damage has caused indications of degraded performance to the second train of the SAFETY SYSTEM needed for the current operating mode Event damage has resulted in VISIBLE DAMAGE to the second train of the SAFETY SYSTEM needed for the current operating mode (Notes 10, 11)

Note 10: If the affected SAFETY SYSTEM train was already inoperable or out of service before the hazardous event occurred, then emergency classification is not warranted.

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

Table 1[2]C-6 Hazardous Events Seismic event (earthquake)

Internal or external FLOODING event High winds or tornado strike FIRE EXPLOSION Other events with similar hazard characteristics as determined by the Shift Manager to 0CAN091801

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Mode Applicability:

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

EXPLOSION - A rapid, violent and catastrophic failure of a piece of equipment due to combustion, chemical reaction or overpressurization. A release of steam (from high energy lines or components) or an electrical component failure (caused by short circuits, grounding, arcing, etc.) should not automatically be considered an explosion. Such events require a post-event inspection to determine if the attributes of an explosion are present.

FIRE - Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred but is not required if large quantities of smoke and heat are observed.

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 10 CFR 50.2).

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

(1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.

VISIBLE DAMAGE - Damage to a SAFETY SYSTEM train that is readily observable without measurements, testing, or analysis. The visual impact of the damage is sufficient to cause concern regarding the operability or reliability of the affected SAFETY SYSTEM train.

Basis: This IC addresses a hazardous event that causes damage to SAFETY SYSTEMS needed for the current operating mode. In order to provide the appropriate context for consideration of an ALERT classification, the hazardous event must have caused indications of degraded SAFETY SYSTEM performance in one train, and there must be either indications of performance issues with the second SAFETY SYSTEM train or VISIBLE DAMAGE to the second train such that the potential exists for this second SAFETY SYSTEM train to have performance issues. In other words, in order for this EAL to be classified, the hazardous event must occur, at least one SAFETY SYSTEM train must have indications of degraded performance, and the second SAFETY SYSTEM train must have indications of degraded performance or VISIBLE DAMAGE such that the potential exists for performance issues. Note that this second SAFETY SYSTEM to 0CAN091801

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

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

VISIBLE DAMAGE addresses damage to a SAFETY SYSTEM train that is not in service/operation and that potentially could cause performance issues. 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. This VISIBLE DAMAGE should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train.

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.1 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 1.b.2 addresses damage to a SAFETY SYSTEM component that is not in service/operation or readily apparent through indications alone, or to a structure containing SAFETY SYSTEM components. Operators will make this determination based on the totality of available event and damage report information. This is intended to be a brief assessment not requiring lengthy analysis or quantification of the damage.

Escalation of the emergency classification level would be via IC CS1 or AS1.

This EAL is the cold condition equivalent of the hot condition EAL SA9.1.

Reference(s):

1. EP FAQ 2016-002 2. NEI 99-01 CA6 to 0CAN091801

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Category E - Independent Spent Fuel Storage Installation (ISFSI)

EAL Group: ANY (EALs in this category are applicable to any plant condition, hot or cold.)

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

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

The ANO ISFSI is located wholly within the plant PROTECTED AREA. Therefore any security event related to the ISFSI is classified under Category H1 security event related EALs.

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Category: E - ISFSI Subcategory: Confinement Boundary

Initiating Condition: Damage to a loaded cask CONFINEMENT BOUNDARY EAL: EU1.1 Unusual Event Damage to a loaded cask CONFINEMENT BOUNDARY as indicated by an on-contact radiation reading on the surface of a loaded spent fuel cask (VSC-24 VCC or HI-STORM

overpack) > any Table 1[2]-E-1 value Table 1[2]E-1 ISFSI Dose Rates VSC-24 VCC HI-STORM 200 mrem/hr on the sides 400 mrem/hr on the top 700 mrem/hr at the air inlet 200 mrem/hr at the air outlet 60 mrem/hr (gamma + neutron) on the top or outlet vent 600 mrem/hr (gamma + neutron) on the side of the overpack (excluding inlet and outlet ducts)

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 ANO ISFSI, the Confinement Boundary is comprised of either the Multi-assembly Sealed Basket (MSB) (SNC System) or Multi-Purpose Canister (MPC) (Holtec System).

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.

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 to 0CAN091801

Page 111 of 257 Attachment 1 - Emergency Action Level Technical Bases

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 specified EAL threshold values correspond to 2 times the cask technical specification values (ref. 1, 2). The technical specification (licensing bases document) multiple of "2 times", which is also used in Recognition Category A IC AU1, is used here to distinguish between non-emergency and emergency conditions. The emphasis for this classification is the degradation in the level of safety of the spent fuel cask and not the magnitude of the associated dose or dose rate. It is recognized that in the case of extreme damage to a loaded cask, the fact that the "on-contact" dose rate limit is

exceeded may be determined based on measurement of a dose rate at some distance from the cask.

Security-related events for ISFSIs are covered under ICs HU1 and HA1.

Reference(s):

1. Certificate of Compliance Appendix A Technical Specifications for the HI-STORM 100 Cask System Section 5.7.4 2. VSC-24 Storage Cask Final Safety Analysis Report Section 1.2.4 Maximum External Surface Dose Rate 3. NEI 99-01 E-HU1 to 0CAN091801

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Category F - Fission Product Barrier Degradation EAL Group: Hot Conditions (RCS temperature > 200°F); EALs in this category are applicable only in one or more hot operating modes.

EALs in this category represent threats to the defense in depth design concept that precludes the release of highly radioactive fission products to the environment. This concept relies on multiple physical barriers any one of which, if maintained intact, precludes the release of significant amounts of radioactive fission products to the environment. The primary fission product barriers are:

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

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

C. Containment Barrier (CNB): The Containment Barrier includes the Reactor Building and connections up to and including the outermost containment isolation valves. This barrier also includes the main steam, feedwater, and blowdown line extensions outside the Reactor Building up to and including the outermost secondary side isolation valve. Containment Barrier thresholds are used as criteria for escalation of the Emergency Classification Level (ECL) from an Alert to a Site Area Emergency or a General Emergency.

The EALs in this category require evaluation of the loss and potential loss thresholds listed in the fission product barrier matrix of Table F-1. "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 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 third barrier to 0CAN091801

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The logic used for emergency classification based on fission product barrier monitoring should reflect the following considerations:

The Fuel Clad Barrier and the RCS Barrier are weighted more heavily than the Containment Barrier. Unusual Event ICs associated with RCS and Fuel Clad Barriers are addressed under System Malfunction ICs. For accident conditions involving a radiological release, evaluation of the fission product barrier thresholds will need to be performed in conjunction with dose assessments to ensure correct and timely escalation of the emergency classification. For example, an evaluation of the fission product barrier thresholds may result in a Site Area Emergency classification while a dose assessment may indicate that an EAL for General Emergency IC AG1 has been exceeded. The fission product barrier thresholds specified within a scheme reflect plant-specific ANO 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 containment, an interfacing system, or outside of the containment. The release of liquid or steam mass from the RCS due to the as-designed/expected operation of a relief valve is not considered to be RCS leakage. At the Site Area Emergency level, EAL users should maintain cognizance of how far present conditions are from meeting a threshold that would require a General Emergency declaration. For example, if the Fuel Clad and RCS fission product barriers were both lost, then there should be frequent assessments of containment radioactive inventory and integrity. Alternatively, if both the Fuel Clad and RCS fission product barriers were potentially lost, the Emergency Director would have more assurance that there was no immediate need to escalate to a General Emergency.

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Category: F - Fission Product Barrier Degradation Subcategory: N/A Initiating Condition:

Any loss or any potential loss of either Fuel Clad or RCS EAL: FA1.1 Aler t Any loss or any potential loss of either Fuel Clad or RCS barrier (Table 1[2]F-1)

Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown

Definition(s):

None Basis: Fuel Clad, RCS and Containment comprise the fission product barriers. Table 1[2]F-1 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.

Reference(s):

1. NEI 99-01 FA1 to 0CAN091801

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Category: F - Fission Product Barrier Degradation Subcategory: N/A Initiating Condition: Loss or potential loss of any two barriers EAL: FS1.1 Site Area Emergency Loss or potential loss of any two barriers (Table 1[2]F-1)

Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown

Definition(s):

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

Basis: Fuel Clad, RCS and Containment comprise the fission product barriers. Table 1[2]F-1 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 would have greater assurance that escalation to a General Emergency is less IMMINENT.

Reference(s):

1. NEI 99-01 FS1

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Category: F - Fission Product Barrier Degradation Subcategory: N/A Initiating Condition: Loss of any two barriers and loss or potential loss of third barrier EAL: FG1.1 General Emergency Loss of any two barriers AND Loss or potential loss of the third barrier (Table 1[2]F-1)

Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown

Definition(s):

None Basis:

Fuel Clad, RCS and Containment comprise the fission product barriers. Table 1[2]F-1 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 Reference(s):

1. NEI 99-01 FG1 to 0CAN091801

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Table 1[2]F-1 Fission Product Barrier Threshold Matrix & Bases

Table 1[2]F-1 lists the threshold conditions that define the Loss and Potential Loss of the three fission product barriers (Fuel Clad, Reactor Coolant System, and Containment). The table is structured so that each of the three barriers occupies adjacent columns. Each fission product barrier column is further divided into two columns; one for Loss thresholds and one for Potential Loss thresholds.

The first column of the table (to the left of the Fuel Clad Loss column) lists the categories (types) of fission product barrier thresholds. The fission product barrier categories are:

A. RCS or S/G Tube Leakage B. Inadequate Heat removal C. Containment Radiation / RCS Activity D. Containment Integrity or Bypass E. Emergency Director Judgment Each category occupies a row in Table 1[2]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 barrier column beginning with number one (ex., FCB1, FCB2-FCB9).

If a cell in Table 1[2]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 1[2]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 1[2]F-1, locates the likely category and then reads across the fission product barrier Loss and Potential Loss thresholds in that category to determine if a threshold has been exceeded. If a threshold has not been exceeded, the EAL-user proceeds to the next likely category and continues review of the thresholds in the new

category.

If the EAL-user determines that any threshold has been exceeded, by definition, the barrier is lost or potentially lost - even if multiple thresholds in the same barrier column are exceeded, only that one barrier is lost or potentially lost. The EAL-user must examine each of the three fission product barriers to determine if other barrier thresholds in the category are lost or potentially lost. For example, if containment radiation is sufficiently high, a Loss of the Fuel to 0CAN091801

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Clad 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,-, E.

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Table 1[2]F-1 Fission Product Barrier Threshold Matrix Fuel Clad Barrier (FCB)

Reactor Coolant System Barrier (RCB)

Containment Barrier (CNB)

Category Loss Potential Loss Loss Potential Loss Loss Potential Loss A RCS or S/G Tube Leakage None FCB1 RVLMS Levels 1 through 9 [1 through 7] indicate DRY RCB1 An automatic or manual ESAS [ESFAS] actuation required by EITHER: UNISOLABLE RCS leakage S/G tube RUPTURE RCB2 UNISOLABLE RCS leakage or S/G tube leakage > 50[44] gpm excluding normal reductions in RCS inventory (e.g., letdown, RCP seal leakoff)

RCB3 Unit 1: PTS limits apply (RT14) AND RCS pressure and temperature are left of the NDTT/LTOP limit lines on EOP Figure 3 (Note 12) Unit 2: Uncontrolled RCS cooldown (> 50 °F step change or > 100 °F change in less than a one-hour period) AND RCS pressure and temperature are to the left of line B (200 degrees MTS),

Standard Attachment 1, P-T Limits (Note 12) CNB1 A S/G that is leaking > 50[44] gpm (excluding normal reductions in RCS inventory) or that is RUPTURED is also FAULTED outside of containment None B Inadequate Heat Removal FCB2 CETs > 1200°F FCB3 CETs > 700°F FCB4 RCS heat removal cannot be established using steam generators AND An on-shift SRO has determined that the procedure conditions are met to commence initiation of HPI[Once Through] cooling None RCB4 RCS heat removal cannot be established using steam generators AND An on-shift SRO has determined that the procedure conditions are met to commence initiation of HPI[Once Through] cooling None CNB2 CETs > 1200°F AND Restoration procedures not effective within 15 min. (Note 1)

C CTMT Radiation /

RCS Activity FCB5 Containment High Range Radiation Monitor RE-8060/8061

[2RE-8925-1/ 8925-2]

> 750 [700] R/hr FCB6 Coolant activity

> 300 Ci/gm dose equivalent I-131 None RCB5 Containment High Range Radiation Monitor RE-8060/8061 [2RE-8925-1/8925-2] > 40 [50] R/hr None None CNB3 Containment High Range Radiation Monitor RE-8060/8061 [2RE-8925-1/8925-2] > 10,000 [12,000] R/hr to 0CAN091801

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Table 1[2]F-1 Fission Product Barrier Threshold Matrix Fuel Clad Barrier (FCB)

Reactor Coolant System Barrier (RCB)

Containment Barrier (CNB)

Category Loss Potential Loss Loss Potential Loss Loss Potential Loss D CTMT Integrity or Bypass None None None None CNB4 Containment isolation is required AND EITHER: Containment integrity has been lost based on Emergency Director judgment UNISOLABLE pathway from Containment to the environment exists CNB5 Indications of RCS leakage outside of Containment CNB6 Containment pressure > 73.7 psia CNB7 Containment hydrogen concentration

> 4% CNB8 Containment pressure > 44.7 psia [23.3 psia] with < one full train of containment heat removal systems (Note 9) operating per design for 15 min. (Note 1)

E Emergency Director Judgment FCB7 Any condition in the opinion of the Emergency Director that indicates loss of the Fuel Clad barrier FCB8 Any condition in the opinion of the Emergency Director that indicates potential loss of the Fuel Clad barrier RCB6 Any condition in the opinion of the Emergency Director that indicates loss of the RCS barrier RCB7 Any condition in the opinion of the Emergency Director that indicates potential loss of the RCS barrier CNB9 Any condition in the opinion of the Emergency Director that indicates loss of the Containment barrier CNB10 Any condition in the opinion of the Emergency Director that indicates potential loss of the Containment barrier to 0CAN091801

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Barrier: Fuel Clad Category: A - RCS or S/G Tube Leakage

Degradation Threat: Loss Threshold:

None

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Barrier: Fuel Clad Category: A - RCS or S/G Tube Leakage

Degradation Threat: Potential Loss Threshold:

FCB1 RVLMS Levels 1 through 9 [1 through 7] indicate DRY Definition(s):

None Basis: This reading indicates a reduction in reactor vessel water level sufficient to allow the onset of heat-induced cladding damage.

There is no Fuel Clad Barrier Loss threshold associated with RCS or S/G Tube Leakage.

Reference(s):

1. ULD-1-SYS-24 Unit 1 Inadequate Core Cooling System 2. Calculation 84-EQ-0080-02 Loop Error Analysis for Reactor Vessel Level Monitoring System 3. ULD-2-SYS-24 Unit 2 Inadequate Core Cooling Monitoring System
4. Calculation 90-E-0116-01 Unit 2 EOP Setpoint Document, Setpoint R.3 5. NEI 99-01 RCS or SG Tube Leakage Potential Loss 1.A

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Barrier: Fuel Clad Category: B - Inadequate Heat Removal

Degradation Threat: Loss Threshold:

FCB2 CETs > 1200 °F Definition(s):

None Basis: This reading indicates temperatures within the core are sufficient to cause significant superheating of reactor coolant.

Reference(s):

1. NEI 99-01 Inadequate Heat Removal Loss 2.A to 0CAN091801

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Barrier: Fuel Clad Category: B - Inadequate Heat Removal

Degradation Threat: Potential Loss Threshold:

FCB3 CETs > 700 °F Definition(s):

None Basis: This reading indicates a reduction in reactor vessel water level sufficient to allow the onset of heat-induced cladding damage.

Reference(s):

1. NEI 99-01 Inadequate Heat Removal Potential Loss 2.A to 0CAN091801

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Barrier: Fuel Clad Category: B - Inadequate Heat Removal

Degradation Threat: Potential Loss Threshold:

FCB4 RCS heat removal cannot be established using steam generators AND An on-shift SRO has determined that the procedure conditions are met to commence initiation of HPI[Once Through] cooling Definition(s):

None

Basis: This condition indicates an extreme challenge to the ability to remove RCS heat using the steam generators (i.e., loss of an effective secondary-side heat sink). This condition represents a potential loss of the Fuel Clad Barrier. In accordance with EOPs, there may be unusual accident conditions during which operators intentionally reduce the heat removal capability of the steam generators; during these conditions, classification using this threshold is not warranted.

In combination with Potential Loss RCB4, meeting this threshold results in a Site Area Emergency.

Reference(s):

1. OP-1202.004 Overheating 2. OP-1202.013 Figure 4, Core Exit Thermocouple for Inadequate Core Cooling 3. OP-2202.006 Loss of Feedwater 4. OP-2202.009 Functional Recovery, Safety Function Status Check 5
5. NEI 99-01 Inadequate Heat Removal Potential Loss 2.B

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Barrier: Fuel Clad Category: C - CTMT Radiation / RCS Activity

Degradation Threat: Loss Threshold:

FCB5 Containment High Range Radiation Monitor RE-8060/8061 [2RE-8925-1/8925-2] > 750 [700] R/hr Definition(s):

None Basis:

The containment radiation monitor reading (768[682] R/hr rounded to 750[700] R/hr for readability) corresponds to an instantaneous release of all reactor coolant mass into the containment, assuming that reactor coolant activity equals 300 µCi/gm dose equivalent I-131. Reactor coolant activity above this level is greater than that expected for iodine spikes and

corresponds to 51.49[1.13]an approximate range of 2% to 3% fuel clad damage. Since this condition indicates that a significant amount of fuel clad damage has occurred, it represents a loss of the Fuel Clad Barrier.

The radiation monitor reading in this threshold is higher than that specified for RCS Barrier Loss

threshold C.1RCB5 since it indicates a loss of both the Fuel Clad Barrier and the RCS Barrier. Note that a combination of the two monitor readings appropriately escalates the ECL to a Site Area Emergency.

There is no Potential Loss threshold associated with CTMT Radiation/

RCS Activity

/Containment Radiation. Basis Reference(s):

1. EP-CALC-ANO-1702 Containment High Range Radiation Monitor EAL Values 2. NEI 99-01 RCS Activity/Containment Radiation FC Loss 3.A to 0CAN091801

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Barrier: Fuel Clad Category: C - CTMT Radiation / RCS Activity

Degradation Threat: Loss Threshold:

FCB6 Coolant activity > 300 Ci/gm dose equivalent I-131 Definition(s):

None Basis: This threshold indicates that RCS radioactivity concentration is greater than 300 µCi/gm dose equivalent I-131. Reactor coolant activity above this level is greater than that expected for iodine spikes and corresponds to an approximate range of 2% to 5% fuel clad damage. Since this condition indicates that a significant amount of fuel clad damage has occurred, it represents a loss of the Fuel Clad Barrier.

It is recognized that sample collection and analysis of reactor coolant with highly elevated activity levels could require several hours to complete. Nonetheless, a sample-related threshold is included as a backup to other indications.

There is no Potential Loss threshold associated with CTMT Radiation/

RCS Activity

/Containment Radiation. Reference(s):

1. NEI 99-01 RCS Activity/Containment Radiation Fuel Clad Loss 3.B

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Barrier: Fuel Clad Category: C - CTMT Radiation / RCS Activity

Degradation Threat: Potential Loss Threshold:

None

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Barrier: Fuel Clad Category: D - CTMT Integrity or Bypass

Degradation Threat: Loss Threshold:

None

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Barrier: Fuel Clad Category: D - CTMT Integrity or Bypass

Degradation Threat: Potential Loss Threshold:

None

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Barrier: Fuel Clad Category: E - Emergency Director Judgment

Degradation Threat: Loss Threshold:

FCB7 Any condition in the opinion of the Emergency Director that indicates loss of the Fuel Clad barrier Definition(s):

None Basis:

This threshold addresses any other factors that are to be used by the Emergency Director in determining whether the Fuel Clad barrier is lost.

Reference(s):

1. NEI 99-01 Emergency Director Judgment Fuel Clad Loss 6.A

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Barrier: Fuel Clad Category: E - Emergency Director Judgment

Degradation Threat: Potential Loss Threshold:

FCB8 Any condition in the opinion of the Emergency Director that indicates potential loss of the Fuel Clad barrier Definition(s):

None Basis:

This threshold addresses any other factors that are to be used by the Emergency Director in determining whether the Fuel Clad barrier is potentially lost. The Emergency Director should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored.

Reference(s):

1. NEI 99-01 Emergency Director Judgment Potential Fuel Clad Loss 6.A

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Barrier: Reactor Coolant System Category: A - RCS or S/G Tube Leakage

Degradation Threat: Loss Threshold:

RCB1 An automatic or manual ESAS [ESFAS] actuation required by EITHER: UNISOLABLE RCS leakage S/G tube RUPTURE Definition(s):

FAULTED - The term applied to a steam generator that has a steam leak on the secondary side of sufficient size to cause an uncontrolled drop in steam generator pressure or the steam generator to become completely depressurized.

RUPTURED - The condition of a steam generator in which primary-to-secondary leakage is of sufficient magnitude to require a safety injection (automatic or manual).

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

Basis: Failure to isolate the leak (from the Control Room or locally), within 15 minutes or if known that the leak cannot be isolated within 15 minutes, from the start of the leak requires immediate classification.

This threshold is based on an UNISOLABLE RCS leak of sufficient size to require an automatic or manual actuation of the Emergency Core Cooling System (ECCS). This condition clearly represents a loss of the RCS Barrier.

This threshold is applicable to unidentified and pressure boundary leakage, as well as identified leakage. It is also applicable to UNISOLABLE RCS leakage through an interfacing system.

The mass loss may be into any location - inside containment, to the secondary-side (i.e., steam generator tube leakage) or outside of containment.

A steam generator with primary-to-secondary leakage of sufficient magnitude to require a safety injection is considered to be RUPTURED. If a RUPTURED steam generator is also FAULTED outside of containment, the declaration escalates to a Site Area Emergency since the Containment Barrier Loss threshold 1.ACNB1 will also be met.

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Reference(s):

1. OP-1202.010 ESAS 2. OP-2202.003 Loss of Coolant Accident
3. NEI 99-01 RCS or SG Tube Leakage Reactor Coolant System Loss 1.A

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Barrier: Reactor Coolant System Category: A - RCS or S/G Tube Leakage

Degradation Threat: Potential Loss Threshold:

RCB2 UNISOLABLE RCS leakage or S/G tube leakage > 50[44] gpm excluding normal reductions in RCS inventory (e.g., letdown, RCP seal leakoff)

Definition(s):

FAULTED - The term applied to a steam generator that has a steam leak on the secondary side of sufficient size to cause an uncontrolled drop in steam generator pressure or the steam generator to become completely depressurized.

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

Basis: Failure to isolate the leak (from the Control Room or locally), within 15 minutes or if known that the leak cannot be isolated within 15 minutes, from the start of the leak requires immediate classification.

This threshold is based on an UNISOLABLE RCS leak that results in the inability to maintain pressurizer level within specified limits by operation of a normally used makeup [charging] (makeup) pump, but an ECCS (SI)ESAS [ESFAS] actuation has not occurred.

The threshold is met when letdown has been isolated and an operating procedure, or operating crew supervision, directs that a standby charging (makeup)makeup [charging] pump be placed in service to restore and maintain pressurizer level.

This threshold is applicable to unidentified and pressure boundary leakage, as well as identified leakage. It is also applicable to UNISOLABLE RCS leakage through an interfacing system. The mass loss may be into any location - inside containment, to the secondary-side (i.e., steam generator tube leakage) or outside of containment.

If a leaking steam generator is also FAULTED outside of containment, the declaration escalates to a Site Area Emergency since the Containment Barrier Loss threshold 1.ACNB1 will also be met. to 0CAN091801

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Reference(s):

1. 1SAR 9.1 Makeup and Purification System 2. 2SAR 9.3.4 Chemical and Volume Control System
3. NEI 99-01 RCS or SG Tube Leakage Reactor Coolant System Potential Loss 1.A

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Barrier: Reactor Coolant System Category: A - RCS or S/G Tube Leakage

Degradation Threat: Potential Loss Threshold:

RCB3 Unit 1: PTS limits apply (RT14) AND RCS pressure and temperature are left of the NDTT/LTOP limit lines, on EOP Figure 3 (Note 12)

Unit 2: Uncontrolled RCS cooldown (> 50 °F step change or > 100 °F change in less than a one-hour period)

AND RCS pressure and temperature are to the left of line B (200 degrees MTS), Standard , P-T Limits (Note 12)

Note 12: Once PTS limits are first invoked, if RCS temperature and pressure are not brought within the limits within 15 minutes, this threshold is met and an immediate declaration is warranted. This threshold is met immediately upon exceeding the limits after this initial 15 minute period until PTS limits no longer apply.

Definition(s):

None Basis: This condition indicates an extreme challenge to the integrity of the RCS pressure boundary due to pressurized thermal shock - a transient that causes rapid RCS cooldown while the RCS is in Mode 3 or higher (i.e., hot and pressurized).

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Reference(s):

1. OP-1202.012 Repetitive Task 14 Control RCS Pressure 2. OP-1202.013 EOP Figures, Figure 3 RCS Pressure vs Temperature Limits
3. OP-1202.011 HPI Cooldown
4. Calculation No: 90-E-0116-01 ANO- EOP Setpoint Basis Document OP Setpoint P.2, RCS Pressure-Temperature 5. OP-2202.010 Standard Attachments, Attachment 1, P-T Limits 6. NEI 99-01 RCS or SG Tube Leakage Reactor Coolant System Potential Loss 1.B

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Barrier: Reactor Coolant System Category: B - Inadequate Heat Removal

Degradation Threat: Loss Threshold:

None

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Barrier: Reactor Coolant System Category: B - Inadequate Heat Removal

Degradation Threat: Potential Loss Threshold:

RCB4 RCS heat removal cannot be established using steam generators AND An on-shift SRO has determined that the procedure conditions are met to commence initiation of HPI[Once Through] cooling Definition(s):

None Basis: This condition indicates an extreme challenge to the ability to remove RCS heat using the steam generators (i.e., loss of an effective secondary-side heat sink). This condition represents a potential loss of the RCS Barrier. In accordance with EOPs, there may be unusual accident conditions during which operators intentionally reduce the heat removal capability of the steam generators; during these conditions, classification using this threshold is not warranted.

In combination with Potential Loss FCB4, meeting this threshold results in a Site Area Emergency.

Meeting this threshold results in a Site Area Emergency because this threshold is identical to Fuel Clad Barrier Potential Loss threshold 2.BFCB4; both will be met. This condition warrants a Site Area Emergency declaration because inadequate RCS heat removal may result in fuel heat-up sufficient to damage the cladding and increaseraise RCS pressure to the point where mass will be lost from the system.

There is no RCS barrier Loss threshold associated with Inadequate Heat Removal.

Reference(s):

1. OP-1202.004 Overheating 2. OP-1202.013 Figure 4, Core Exit Thermocouple for Inadequate Core Cooling
3. OP-2202.006 Loss of Feedwater 4. OP-2202.009 Functional Recovery, Safety Function Status Check 5 5. NEI 99-01 Inadequate Heat Removal RCS Potential Loss 2.B to 0CAN091801

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Barrier: Reactor Coolant System Category: C - CTMT Radiation/ RCS Activity

Degradation Threat: Loss Threshold:

RCB5 Containment High Range Radiation Monitor RE-8060/8061 [2RE-8925-1/8925-2] > 40 [50] R/hr Definition(s):

None Basis: NRC Information Notice 97-045

, Supplement 1

, identifies the potential for erratic indications from the high range radiation monitors (HRRMs) as a result of thermally induced currents (TIC) which may cause the HRRM to read falsely high (for approximately 15 minutes) on a rapid temperature rise, and fail low intermittently on a rapid temperature fall.

Because of this phenomenon, any trends or alarms on the HRRM's should be validated by comparison to the containment low range/area radiation monitors and Air Monitoring Systems trends before actions are taken.

The containment radiation monitor reading (42.8[50.4] R/hr rounded to 40[50] R/hr for readability) corresponds to an instantaneous release of all reactor coolant mass into the containment, assuming that reactor coolant activity equals Technical Specification allowable limits. This value is lower than that specified for Fuel Clad Barrier Loss threshold 3.AFCB5 since it indicates a loss of the RCS Barrier only.

There is no Potential Loss threshold associated with RCS Activity / ContainmentCTMT Radiation/RCS Activity. Reference(s):

1. EP-CALC-ANO-1702 Containment High Range Radiation Monitor EAL Values 2. NEI 99-01 CMT Radiation / RCS Activity RCS Loss 3.A to 0CAN091801

Page 142 of 257 Attachment 1 - Emergency Action Level Technical Bases

Barrier: Reactor Coolant System Category: B - CTMT Radiation/ RCS Activity

Degradation Threat: Potential Loss Threshold:

None

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Barrier: Reactor Coolant System Category: D - CTMT Integrity or Bypass

Degradation Threat: Loss Threshold:

None

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Barrier: Reactor Coolant System Category: D - CTMT Integrity or Bypass

Degradation Threat: Potential Loss Threshold:

None

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Barrier: Reactor Coolant System Category: E - Emergency Director Judgment

Degradation Threat: Loss Threshold:

RCB6 Any condition in the opinion of the Emergency Director that indicates loss of the RCS barrier Definition(s):

None Basis: This threshold addresses any other factors that may be used by the Emergency Director in determining whether the RCS Barrier is lost.

Reference(s):

1. NEI 99-01 Emergency Director Judgment RCS Loss 6.A to 0CAN091801

Page 146 of 257 Attachment 1 - Emergency Action Level Technical Bases

Barrier: Reactor Coolant System Category: E - Emergency Director Judgment

Degradation Threat: Potential Loss Threshold:

RCB7 Any condition in the opinion of the Emergency Director that indicates potential loss of the RCS barrier Definition(s):

None Basis:

This threshold addresses any other factors that may be used by the Emergency Director in determining whether the RCS Barrier is potentially lost. The Emergency Director should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored.

Reference(s):

1. NEI 99-01 Emergency Director Judgment RCS Potential Loss 6.A

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Barrier: Containment Category: A - RCS or S/G Tube Leakage

Degradation Threat: Loss Threshold:

CNB1 A S/G that is leaking > 50[44] gpm (excluding normal reductions in RCS inventory) or that is RUPTURED S/G is also FAULTED outside of containment Definition(s):

FAULTED - The term applied to a steam generator that has a steam leak on the secondary side of sufficient size to cause an uncontrolled drop in steam generator pressure or the steam generator to become completely depressurized.

RUPTURED - The condition of a steam generator in which primary-to-secondary leakage is of sufficient magnitude to require a safety injection (automatic or manual).

Basis:

This threshold addresses a leaking or RUPTURED Steam Generator (SG) that is also FAULTED outside of containment. The condition of the SG, whether leaking or RUPTURED, is determined in accordance with the thresholds for RCS Barrier Potential Loss 1.A RCB2 and Loss 1.ARCB1, respectively. This condition represents a bypass of the containment barrier.

FAULTED is a defined term within the NEI 99-01 methodology; this determination is not necessarily dependent upon entry into, or diagnostic steps within, an EOP. For example, if the pressure in a steam generator is droppecreasing uncontrollably

([part of the FAULTED definition

)] and the FAULTED steam generator isolation procedure is not entered because EOP user rules are dictating implementation of another procedure to address a higher priority condition, the steam generator is still considered FAULTED for emergency classification purposes.

The FAULTED criterion establishes an appropriate lower bound on the size of a steam release that may require an emergency classification. Steam releases of this size are readily observable with normal Control Room indications. The lower bound for this aspect of the containment barrier is analogous to the lower bound criteria specified in IC SU4 for the fuel clad barrier (i.e., RCS activity values) and IC SU5 for the RCS barrier (i.e., RCS leak rate values).

This threshold also applies to prolonged steam releases necessitated by operational considerations such as the forced steaming of a leaking or RUPTURED steam generator directly to atmosphere to cooldown the plant, or to drive an auxiliary (emergency) feed water pump. These types of conditions will result in a significant and sustained release of radioactive to 0CAN091801

Page 148 of 257 Attachment 1 - Emergency Action Level Technical Bases

steam to the environment (and are thus similar to a FAULTED condition). The inability to isolate the steam flow without an adverse effect on plant cooldown meets the intent of a loss of containment.

Steam releases associated with the expected operation of a SG power operated relief valve or safety relief valve do not meet the intent of this threshold. Such releases may occur intermittently for a short period of time following a reactor trip as operators process through emergency operating procedures to bring the plant to a stable condition and prepare to initiate a plant cooldown. Steam releases associated with the unexpected operation of a valve (e.g., a stuck-open safety valve) do meet this threshold.

Following a n SG tube leak or rupture, there may be minor radiological releases through a secondary-side system component (e.g., air ejectors, glad seal exhausters, valve packing, etc.). These types of releases do not constitute a loss or potential loss of containment but should be evaluated using the Recognition Category A ICs.

The emergency classification levelECLs resulting from primary-to-secondary leakage, with or without a steam release from the FAULTED SG, are summarized below.

Affected SG is FAULTED Outside of Containment?

P-to-S Leak Rate Yes No Less than or equal to 25 gpm No classification No classification Greater than 25 gpm Unusual Event per SU4SU5.1 Unusual Event per SU4SU5.1 Greater than 50[44] gpm (RCS Barrier Potential Loss

) Site Area Emergency per FS1.1 Alert per FA1

.1 Requires an automatic or manual ESAS [ESFAS]ECCS (SIAS) actuation (RCS Barrier Loss

) Site Area Emergency per FS1.1 Alert per FA1

.1 There is no Potential Loss threshold associated with RCS or S

/G Tube Leakage.

Reference(s):

1. NEI 99-01 RCS or SG Tube Leakage Containment Loss 1.A

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Barrier: Containment Category: A - RCS or S/G Tube Leakage

Degradation Threat: Potential Loss Threshold:

None

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Page 150 of 257 Attachment 1 - Emergency Action Level Technical Bases

Barrier: Containment Category: B - Inadequate Heat Removal

Degradation Threat: Loss Threshold:

None

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Barrier: Containment Category: B - Inadequate Heat Removal

Degradation Threat: Potential Loss Threshold:

CNB2 CETs > 1200°F AND Restoration procedures not effective within 15 min. (Note 1)

Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded.

Definition(s):

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

Basis: The restoration procedure is considered "effective" if core exit thermocouple readings are droppdecreasing and/or if reactor vessel level is risincreasing. Whether or not the procedure(s) will be effective should be apparent within 15 minutes. The Emergency Director should escalate the emergency classification level as soon as it is determined that the procedure(s) will not be

effective.

This condition represents an IMMINENT core melt sequence which, if not corrected, could lead to vessel failure and an increased potential for containment failure. For this condition to occur, there must already have been a loss of the RCS Barrier and the Fuel Clad Barrier. If implementation of a procedure(s) to restore adequate core cooling is not effective (successful) within 15 minutes, it is assumed that the event trajectory will likely lead to core melting and a subsequent challenge of the Containment Barrier.

Severe accident analyses (e.g., NUREG-1150) have concluded that function restoration procedures can arrest core degradation in a significant fraction of core damage scenarios, and that the likelihood of containment failure is very small in these events. Given this, it is appropriate to provide 15 minutes beyond the required entry point to determine if procedural actions can reverse the core melt sequence.

Reference(s):

1. NEI 99-01 Inadequate Heat Removal Containment Potential Loss 2.A to 0CAN091801

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Barrier: Containment Category: C - CTMT Radiation/RCS Activity

Degradation Threat: Loss Threshold:

None

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Barrier: Containment Category: C - CTMT Radiation/RCS Activity

Degradation Threat: Potential Loss Threshold:

CNB3 Containment High Range Radiation Monitor RE-8060/8061 [2RE-8925-1/8925-2]

> 10,000 [12,000] R/hr Definition(s):

None Basis:

The containment radiation monitor reading (10,300[12,100] R/hr rounded to 10,000[12,000] R/hr for readability) corresponds to an instantaneous release of all reactor coolant mass into the containment, assuming that 20% of the fuel cladding has failed. This level of fuel clad failure is well above that used to determine the analogous Fuel Clad Barrier Loss and RCS Barrier Loss thresholds.

NUREG-1228, Source Estimations During Incident Response to Severe Nuclear Power Plant Accidents, indicates the fuel clad failure must be greater than approximately 20% in order for there to be a major release of radioactivity requiring offsite protective actions. For this condition to exist, there must already have been a loss of the RCS Barrier and the Fuel Clad Barrier. It is therefore prudent to treat this condition as a potential loss of containment which would then escalate the emergency classification levelECL to a General Emergency.

There is no Loss threshold associated with RCS Activity/ContainmentCTMT Radiation

/RCS Activity. Reference(s):

1. EP-CALC-ANO-1702 Containment High Range Radiation Monitor EAL Values 2. NEI 99-01 CTMT Radiation / RCS Activity Containment Potential Loss 3.A to 0CAN091801

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Barrier: Containment Category: D - CTMT Integrity or Bypass

Degradation Threat: Loss Threshold:

CNB4 Containment isolation is required AND EITHER:

Containment integrity has been lost based on Emergency Director judgment UNISOLABLE pathway from Containment to the environment exists Definition(s):

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

Basis: Failure to isolate the leak (from the Control Room or locally), within 15 minutes or if known that the leak cannot be isolated within 15 minutes, from the start of the leak requires immediate classification.

The status of the containment barrier during an event involving steam generator tube leakage is assessed using Loss Threshold 1.ACNB1. These thresholds address a situation where containment isolation is required and one of two conditions exists as discussed below. Users are reminded that there may be accident and release conditions that simultaneously meet both bulleted thresholds 4.A.1 and 4.A.2. 4.A.1First Threshold - Containment integrity has been lost, i.e., the actual containment atmospheric leak rate likely exceeds that associated with allowable leakage (or sometimes referred to as design leakage). Following the release of RCS mass into containment, containment pressure will fluctuate based on a variety of factors; a loss of containment integrity condition may (or may not) be accompanied by a noticeable drop in containment pressure.

Recognizing the inherent difficulties in determining a containment leak rate during accident conditions, it is expected that the Emergency Director will assess this threshold using judgment, and with due consideration given to current plant conditions, and available operational and radiological data (e.g., containment pressure, readings on radiation monitors outside containment, operating status of containment pressure control equipment, etc.).

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Refer to the middle piping run of Figure 9-F-41. Two simplified examples are provided. One is leakage from a penetration and the other is leakage from an in-service system valve. Depending upon radiation monitor locations and sensitivities, the leakage could be detected by any of the four monitors depicted in the figure.

Another example would be a loss or potential loss of the RCS barrier, and the simultaneous occurrence of two FAULTED locations on a steam generator where one fault is located inside containment (e.g., on a steam or feedwater line) and the other outside of containment. In this case, the associated steam line provides a pathway for the containment atmosphere to escape to an area outside the containment.

Following the leakage of RCS mass into containment and a rise in containment pressure, there may be minor radiological releases associated with allowable (design) containment leakage through various penetrations or system components. These releases do not constitute a loss or potential loss of containment but should be evaluated using the Recognition Category A ICs.

4.A.2Second Threshold - Conditions are such that there is an UNISOLABLE pathway for the migration of radioactive material from the containment atmosphere to the environment. As used here, the term "environment" includes the atmosphere of a room or area, outside the containment, that may, in turn, communicate with the outside-the-plant atmosphere (e.g.,

through discharge of a ventilation system or atmospheric leakage). Depending upon a variety of factors, this condition may or may not be accompanied by a noticeable drop in containment pressure.

Refer to the top piping run of Figure 9-F-41. In this simplified example, the inboard and outboard isolation valves remained open after a containment isolation was required (i.e., containment isolation was not successful). There is now an UNISOLABLE pathway from the containment to the environment.

The existence of a filter is not considered in the threshold assessment. Filters do not remove fission product noble gases. In addition, a filter could become ineffective due to iodine and/or particulate loading beyond design limits (i.e., retention ability has been exceeded) or water saturation from steam/high humidity in the release stream.

Leakage between two interfacing liquid systems, by itself, does not meet this threshold.

Refer to the bottom piping run of Figure 9-F-41. In this simplified example, leakage in an RCP seal cooler is allowing radioactive material to enter the Auxiliary Building. The radioactivity would be detected by the Process Monitor. If there is no leakage from the closed water cooling system to the Auxiliary Building, then no threshold has been met. If the pump developed a leak that allowed steam/water to enter the Auxiliary Building, then second threshold 4.B would be met. Depending upon radiation monitor locations and sensitivities, this leakage could be detected by any of the four monitors depicted in the figure and cause the first threshold 4.A.1 to be met as well.

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Following the leakage of RCS mass into containment and a rise in containment pressure, there may be minor radiological releases associated with allowable containment leakage through various penetrations or system components. Minor releases may also occur if a containment isolation valve(s) fails to close but the containment atmosphere escapes to an enclosed system. These releases do not constitute a loss or potential loss of containment but should be evaluated using the Recognition Category A ICs.

Reference(s):

1. NEI 99-01 CTMT Integrity or Bypass Containment Loss 4.A

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Barrier: Containment Category: D - CTMT Integrity or Bypass

Degradation Threat: Loss Threshold:

CNB5 Indications of RCS leakage outside of Containment Definition(s):

None Basis: To ensure proper escalation of the emergency classification, the RCS leakage outside of containment must be related to the mass loss that is causing the RCS Loss RCB1 and/or Potential Loss RCB2 threshold 1.A to be met.

The status of the containment barrier during an event involving steam generator tube leakage is assessed using Containment Loss Threshold CNB1.

Containment sump, temperature, pressure and/or radiation levels will riincrease if reactor coolant mass is leaking into the containment. If these parameters have not risenincreased , then the reactor coolant mass may be leaking outside of containment (i.e., a containment bypass sequence).

RiIncreases in sump, temperature, pressure, flow and/or radiation level readings outside of the containment may indicate that the RCS mass is being lost outside of containment.

Unexpected elevated readings and alarms on radiation monitors with detectors outside

containment should be corroborated with other available indications to confirm that the source is a loss of RCS mass outside of containment. If the fuel clad barrier has not been lost, radiation monitor readings outside of containment may not riincrease significantly; however, other unexpected changes in sump levels, area temperatures or pressures, flow rates, etc. should be sufficient to determine if RCS mass is being lost outside of the containment.

Refer to the middle piping run of Figure 9-F-41. In this simplified example, a leak has occurred at a reducer on a pipe carrying reactor coolant in the Auxiliary Building. Depending upon radiation monitor locations and sensitivities, the leakage could be detected by any of the four monitors depicted in the figure and cause threshold CNB4 to be met as well.

Reference(s):

1. NEI 99-01 CTMT Integrity or Bypass Containment Loss 4.B to 0CAN091801

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Figure 1: Containment Integrity or Bypass Examples

Open valve Open valve Open valve Open valve Open valve Open valvePenetrationDamperDamper Interface leakage RCP Seal Cooling Inside Reactor Building Auxiliary Building 1 s t Threshold - Airborne 1 s t Threshold - Airborne release from penetration 2 n d Threshold - Airborne release from pathway 2 n d Threshold - RCS leakage outside AB Airborne Monitor Area Monitor Process Monitor Effluent MonitorClosed Cooling Pump to 0CAN091801

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Barrier: Containment Category: D - CTMT Integrity or Bypass

Degradation Threat: Potential Loss

Threshold:

CNB6 Containment pressure > 73.7 psia Definition(s):

None Basis:

If containment pressure exceeds the design pressure, there exists a potential to lose the Containment Barrier. To reach this level, there must be an inadequate core cooling condition for an extended period of time; therefore, the RCS and Fuel Clad barriers would already be lost.

Thus, this threshold is a discriminator between a Site Area Emergency and General Emergency since there is now a potential to lose the third barrier.

Reference(s):

1. 1SAR 1.4.43 Criterion 50 - Containment Design Basis 2. 2SAR Table 6.2-7 Principle Containment Design Parameters
3. NEI 99-01 CTMT Integrity or Bypass Containment Potential Loss 4.A to 0CAN091801

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Barrier: Containment Category: D - CTMT Integrity or Bypass

Degradation Threat: Potential Loss Threshold:

CNB7 Containment hydrogen concentration > 4%

Definition(s):

None Basis: 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). The 4% hydrogen concentration is generally considered the lower limit for hydrogen deflagrations. A hydrogen burn will raise containment pressure and could result in collateral equipment damage leading to a loss of containment integrity. It therefore represents a potential loss of the Containment Barrier.

Reference(s):

1. Unit 1 SAMG Figure III-1B 2. Unit 2 SAMG Phase 1 Instructions, Containment Flowchart
3. NEI 99-01 CTMT Integrity or Bypass Containment Potential Loss 4.B to 0CAN091801

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Barrier: Containment Category: D - CTMT Integrity or Bypass

Degradation Threat: Potential Loss Threshold:

CNB8 Containment pressure > 44.7 psia [23.3 psia] with < one full train of containment heat removal systems (Note 9) operating per design for 15 min. (Note 1)

Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Note 9: One full train of containment heat removal systems consists of one train of RB [Containment] Spray and one train of RB [Containment] Cooling System.

Definition(s):

None Basis:

This threshold describes a condition where containment pressure is greater than the setpoint at which containment energy (heat) removal systems are designed to automatically actuate, and less than one full train of equipment is capable of operating per design. The 15-minute criterion is included to allow operators time to manually start equipment that may not have automatically started, if possible. This threshold represents a potential loss of containment in that containment heat removal/depressurization systems (e.g., containment sprays , ice condenser fans, etc., but not including containment venting strategies) are either lost or performing in a degraded manner.

Reference(s):

1. 1SAR 6.2 Reactor Building Spray System 2. 1SAR 6.3 Reactor Building Cooling System
3. OP-2202.003 Loss of Coolant Accident
4. OP-2202.010 Standard Attachments, Attachment 22
5. 2SAR 6.2.2 Containment Heat Removal Systems 6. 2SAR 7.3.1.1.11.2 Containment Spray System 7. NEI 99-01 CTMT Integrity or Bypass Containment Potential Loss 4.C to 0CAN091801

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Barrier: Containment Category: E - Emergency Director Judgment

Degradation Threat: Loss Threshold:

CNB9 Any condition in the opinion of the Emergency Director that indicates loss of the Containment barrier Definition(s):

None Basis:

This threshold addresses any other factors that may be used by the Emergency Director in determining whether the Containment Barrier is lost.

Reference(s):

1. NEI 99-01 Emergency Director Judgment Containment Loss 6.A

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Barrier: Containment Category: E - Emergency Director Judgment

Degradation Threat: Potential Loss Threshold:

CNB10 Any condition in the opinion of the Emergency Director that indicates potential loss of the Containment barrier Definition(s):

None Basis:

This threshold addresses any other factors that may be used by the Emergency Director in determining whether the Containment Barrier is potentially lost. The Emergency Director should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored.

Reference(s):

1. NEI 99-01 Emergency Director Judgment Containment Potential Loss 6.A

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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 Technological 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 plant 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 Events that are indicative of loss of Control Room habitability. If the Control Room must be evacuated, additional support for monitoring and controlling plant functions is necessary through the emergency response facilities.
7. Emergency Director Judgment The EALs defined in other categories specify the predetermined symptoms or events that are indicative of emergency or potential emergency conditions and thus warrant classification.

While these EALs have been developed to address the full spectrum of possible emergency conditions which may warrant classification and subsequent implementation of the Emergency Plan, a provision for classification of emergencies based on operator/management experience and judgment is still necessary. The EALs of this category provide the Emergency Director the latitude to classify emergency conditions consistent with the established classification criteria based upon Emergency Director judgment. to 0CAN091801

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Category: H - Hazards Subcategory: 1 - Security

Initiating Condition: Confirmed SECURITY CONDITION or threat EAL: HU1.1 Unusual Event A SECURITY CONDITION that does not involve a HOSTILE ACTION as reported by ANO 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):

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 ANO 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 ANO. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the SECURITY OWNER CONTROLLED AREA).

PROJECTILE - An object directed toward a Nuclear Power Plant that could cause concern for its continued operability, reliability, or personnel safety.

PROTECTED AREA - An area clearly demarcated by a fence or building wall with an entrance portal that is regulated by Security Personnel to control access.

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 10 CFR 50.2).

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

(1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.

SECURITY CONDITION - Any security event as listed in the approved security contingency plan that constitutes a threat/compromise to site security, threat/risk to site personnel, or a potential degradation to the level of safety of the plant. A security condition does not involve a HOSTILE ACTION.

SECURITY OWNER CONTROLLED AREA - The SOCA is the area demarcated as a Vehicle Barrier System (VBS) consisting of passive elements including a series of large concrete blocks on the inside of a delay fence with early warning capabilities. The SOCA is the area between the SOCA Fence and the PROTECTED AREA Boundary.

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 or 10 CFR 50.72. Security events assessed as HOSTILE ACTIONS are classifiable under ICs HA1 , and 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 O ffsite R esponse Organization

s. Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan [and Independent Spent Fuel Storage Installation Security Program]. The first threshold EAL #1 references the Security Shift Supervision (site-specific security shift supervision)because these are the individuals trained to confirm that a security event is occurring or has occurred. Training on security event confirmation and classification is controlled due to the nature of Safeguards and 10 CFR 2.39 information.

The second threshold EAL #2 addresses the receipt of a credible security threat. The credibility of the threat is assessed in accordance with OP-1203.048 Security Event (site-specific procedure). The third threshold EAL #3 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 to 0CAN091801

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the NRC. Validation of the threat is performed in accordance with 11-S-82-1 Security Contingency Events (ref. 2)(site-specific procedure). 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 Security Plan for ANO (ref. 1). Escalation of the emergency classification level would be via IC HA1.

Reference(s):

1. ANO Security Plan 2. OP-1203.048 Security Event
3. NEI 99-01 HU1

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Category: H - Hazards Subcategory: 1 - Security

Initiating Condition: HOSTILE ACTION within the SECURITY OWNER CONTROLLED AREA or airborne attack threat within 30 minutes

EAL: HA1.1 Alert A HOSTILE ACTION is occurring or has occurred within the SECURITY OWNER CONTROLLED AREA as reported by ANO 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):

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 ANO 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 ANO. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the SECURITY OWNER CONTROLLED AREA).

HOSTILE FORCE - One or more individuals who are engaged in a determined assault, overtly or by stealth and deception, equipped with suitable weapons capable of killing, maiming, or causing destruction.

PROJECTILE - An object directed toward a Nuclear Power Plant that could cause concern for its continued operability, reliability, or personnel safety.

PROTECTED AREA - An area clearly demarcated by a fence or building wall with an entrance portal that is regulated by Security Personnel to control access.

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SECURITY OWNER CONTROLLED AREA - The SOCA is the area demarcated as a Vehicle Barrier System (VBS) consisting of passive elements including a series of large concrete blocks on the inside of a delay fence with early warning capabilities. The SOCA is the area between the SOCA Fence and the PROTECTED AREA Boundary.

Basis:

This IC addresses the occurrence of a HOSTILE ACTION within the SECURITY 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 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 [and Independent Spent Fuel Storage Installation Security Program]. As time and conditions allow, these events require a heightened state of readiness by the plant staff and implementation of onsite protective measures (e.g., evacuation, dispersal or sheltering). The 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 EAL does not apply to incidents that are accidental events, acts of civil disobedience, or otherwise are not a HOSTILE ACTION perpetrated by a HOSTILE FORCE. Examples include the crash of a small aircraft, shots from hunters, physical disputes between employees, etc. Reporting of these types of events is adequately addressed by other EALs, or the requirements of 10 CFR 73.71 or 10 CFR 50.72.

The first threshold EAL #1 is applicable for any HOSTILE ACTION occurring, or that has occurred, in the SECURITY OWNER CONTROLLED AREA. This includes any action directed against an ISFSI that is located outside the plant PROTECTED AREA.

The second threshold EAL #2 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 OP-1203.048 Security Event (ref. 2)(site-specific procedure). 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 SECURITY 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 to 0CAN091801

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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 Security Plan for ANO (ref. 1). Escalation of the emergency classification level would be via IC HS1.

Reference(s):

1. ANO Security Plan 2. OP-1203.048 Security Event
3. NEI 99-01 HA1 to 0CAN091801

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Category: H - Hazards Subcategory: 1 - Security

Initiating Condition: HOSTILE ACTION within the PROTECTED AREA EAL: HS1.1 Site Area Emergency A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by ANO Security Shift Supervision Mode Applicability:

All Definition(s):

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 ANO 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 ANO. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the SECURITY OWNER CONTROLLED AREA).

HOSTILE FORCE - One or more individuals who are engaged in a determined assault, overtly or by stealth and deception, equipped with suitable weapons capable of killing, maiming, or causing destruction.

PROJECTILE - An object directed toward a Nuclear Power Plant that could cause concern for its continued operability, reliability, or personnel safety.

PROTECTED AREA - An area clearly demarcated by a fence or building wall with an entrance portal that is regulated by Security Personnel to control access.

SECURITY OWNER CONTROLLED AREA - The SOCA is the area demarcated as a Vehicle Barrier System (VBS) consisting of passive elements including a series of large concrete blocks on the inside of a delay fence with early warning capabilities. The SOCA is the area between the SOCA Fence and the PROTECTED AREA Boundary.

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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 (ref. 1, 2). Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan [and Independent Spent Fuel Storage Installation Security Program]. As time and conditions allow, these events require a heightened state of readiness by the plant staff and implementation of onsite protective measures (e.g., evacuation, dispersal or sheltering). The Site Area Emergency declaration will mobilize O ffsite R esponse O rganization (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 EAL does not apply to a HOSTILE ACTION directed at an ISFSI Protected Area located outside the PROTECTED AREA; such an attack should be assessed using IC HA1. It also does not apply to incidents that are accidental events, acts of civil disobedience, or otherwise are not a HOSTILE ACTION perpetrated by a HOSTILE FORCE. Examples include the crash of a small aircraft, shots from hunters, physical disputes between employees, etc. Reporting of these types of events is adequately addressed by other EALs, or the requirements of 10 CFR 73.71 or 10 CFR 50.72.

Emergency plans and implementing procedures are public documents; therefore, EALs should not incorporate Security-sensitive information. This includes information that may be advantageous to a potential adversary, such as the particulars concerning a specific threat or threat location. Security-sensitive information should be contained in non-public documents such as the Security Plan for ANO (ref. 1). Escalation of the emergency classification level would be via IC HG1.

Reference(s):

1. ANO Security Plan 2. OP-1203.048 Security Event
3. NEI 99-01 HS1 to 0CAN091801

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Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 2 - Seismic Event

Initiating Condition: Seismic event greater than OBE levels EAL: HU2.1 Unusual Event Seismic event > OBE as indicated by annunciation of the 0.10 g acceleration alarm Mode Applicability:

All Definition(s):

None Basis: This IC addresses a seismic event that results in accelerations at the plant site greater than

those specified for an Operating Basis Eart hquake (OBE). An earthquake greater than an OBE but less than a Safe Shutdown Earthquake (SSE)Design Basis Earthquake (DBE) should have no significant impact on safety-related systems, structures and components; however, some time may be required for the plant staff to ascertain the actual post-event condition of the plant (e.g., performs walk-downs and post-event inspections). Given the time necessary to perform walk-downs and inspections, and fully understand any impacts, this event represents a potential degradation of the level of safety of the plant.

Event verification with external sources should not be necessary during or following an OBE.

Earthquakes of this magnitude should be readily felt by on-site personnel and recognized as a seismic event (e.g., lateral accelerations in excess of 0.08g1g). The Shift Manager or Emergency Director may seek external verification if deemed appropriate (e.g., a call to the USGS, check internet news sources, etc.); however, the verification action must not preclude a timely emergency declaration.

Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via IC CA6 or SA9.

Two strong motion triaxial accelerometers, ACS-8001 and ACS-8003, located at the base slab provide alarms to the Unit 1 control room via the seismic network control center, C529-NCC. One alarm from C529-NCC is triggered when a setpoint of 0.01g has been exceeded. This alarm indicates that an earthquake has occurred and the seismic monitoring system is recording seismic data. Another alarm from C529-NCC is triggered when the pre-determined value of 0.1g, indicating the OBE has been exceeded (ref. 2, 3).

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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 (NEIC)) can confirm that an earthquake has occurred in the area of the plant. Such confirmation should not, however, preclude a timely emergency declaration based on receipt of the OBE alarm. If requested, provide the analyst with the following ANO coordinates: 35º 18' 36" north latitude, 93º 13' 53" west longitude (ref. 4). Alternatively, near real-time seismic activity can be accessed via the NEIC website:

Reference(s):

1. 1SAR 2.2.1 Location 2. 1SAR 2.7.2 Site Seismic Evaluation 3. 1SAR 2.7.6 Time-History Accelerograph
4. OP-1203.025 Natural Emergencies
5. OP-2203.008 Natural Emergencies
6. NEI 99-01 HU2 to 0CAN091801

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Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 3 - Natural or Technological Hazard

Initiating Condition: Hazardous event EAL: HU3.1 Unusual Event A tornado strike within the PROTECTED AREA Mode Applicability:

All Definition(s):

PROTECTED AREA - An area clearly demarcated by a fence or building wall with an entrance portal that is regulated by Security Personnel to control access.

Basis: This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant.

This EALEAL #1 addresses a tornado striking (touching down) within the PROTECTED AREA.

EAL #2 addresses flooding of a building room or area that results in operators isolating power to a SAFETY SYSTEM component due to water level or other wetting concerns. Classification is also required if the water level or related wetting causes an automatic isolation of a SAFETY SYSTEM component from its power source (e.g., a breaker or relay trip). To warrant classification, operability of the affected component must be required by Technical Specifications for the current operating mode.

EAL #3 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 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. to 0CAN091801

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EAL #5 addresses (site-specific description).

Escalation of the emergency classification level would be based on ICs in Recognition Categories A, F, S or C.

If damage is confirmed visually or by other in-plant indications, the event may be escalated to an Alert under EAL CA6.1 or SA9.1.

A tornado striking (touching down) within the PROTECTED AREA warrants declaration of an Unusual Event regardless of the measured wind speed at the meteorological tower. A tornado is defined as a violently rotating column of air in contact with the ground and extending from the base of a thunderstorm.

Reference(s):

1. OP-1203.025 Natural Emergencies 2. OP-2203.008 Natural Emergencies 3. NEI 99-01 HU3

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Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 3 - Natural or Technological Hazard

Initiating Condition: Hazardous event EAL: HU3.2 Unusual Event Internal room or area FLOODING of a magnitude sufficient to require manual or automatic electrical isolation of a SAFETY SYSTEM component required by Technical Specifications 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 10 CFR 50.2).

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: 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 EAL addresses FLOODING of a building room or area that results in operators isolating power to a SAFETY SYSTEM component due to water level or other wetting concerns.

Classification is also required if the water level or related wetting causes an automatic isolation of a SAFETY SYSTEM component from its power source (e.g., a breaker or relay trip). To to 0CAN091801

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warrant classification, operability of the affected component must be required by Technical Specifications for the current operating mode.

EAL #3 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 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.

EAL #5 addresses (site-specific description).

Escalation of the emergency classification level would be based on ICs in Recognition Categories A, F, S or C.

Refer to EAL CA6.1 or SA9.1 for internal FLOODING affecting more than one SAFETY SYSTEM train.

Reference(s):

1. OP-1203.025 Natural Emergencies 2. OP-2203.008 Natural Emergencies
3. NEI 99-01 HU3 to 0CAN091801

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Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 3 - Natural or Technological Hazard

Initiating Condition: Hazardous event EAL: HU3.3 Unusual Event Movement of personnel within the PROTECTE D AREA is IMPEDED due to an event external to the PROTECTED AREA involving hazardous materials (e.g., an offsite chemical spill or toxic gas release)

Mode Applicability:

All Definition(s):

IMPEDE(D) - Personnel access to a room or area is hindered to an extent that extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., requiring use of protective equipment, such as SCBAs, that is not routinely employed).

PROTECTED AREA - An area clearly demarcated by a fence or building wall with an entrance portal that is regulated by Security Personnel to control access.

Basis: This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant.

This EAL EAL #1 addresses a tornado striking (touching down) within the PROTECTED 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. Classification is also required if the water level or related wetting causes an automatic isolation of a SAFETY SYSTEM component from its power source (e.g., a breaker or relay trip). To warrant classification, operability of the affected component must be required by Technical Specifications for the current operating mode.

EAL #3 addresses a hazardous materials event originating at a n offsite location outside the PROTECTED AREA 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 impediment to vehicle movement and significant enough to prohibit the plant staff from accessing the site using personal vehicles. to 0CAN091801

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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.

EAL #5 addresses (site-specific description).

Escalation of the emergency classification level would be based on ICs in Recognition Categories A, F, S or C.

Reference(s):

1. NEI 99-01 HU3

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Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 3 - Natural or Technological Hazard

Initiating Condition: Hazardous event EAL: HU3.4 Unusual Event A hazardous event that results in on-site conditions sufficient to prohibit the plant staff from accessing the site via personal vehicles (Note 7)

Note 7: This EAL does not apply to routine traffic impediments such as fog, snow, ice, or vehicle breakdowns or accidents.

Mode Applicability:

All Definition(s):

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.

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 EAL addresses flooding of a building room or area that results in operators isolating power to a SAFETY SYSTEM component due to water level or other wetting concerns. Classification is also required if the water level or related wetting causes an automatic isolation of a SAFETY SYSTEM component from its power source (e.g., a breaker or relay trip). To warrant classification, operability of the affected component must be required by Technical Specifications for the current operating mode.

EAL #3 addresses a hazardous materials event originating at an offsite location and of sufficient magnitude to impede the movement of personnel within the PROTECTED AREA.

This EAL EAL #4 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.

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This EAL is not intended to 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.

EAL #5 addresses (site-specific description). Escalation of the emergency classification level would be based on ICs in Recognition Categories A, F, S or C.

Reference(s):

1. NEI 99-01 HU3

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Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 4 - Fire

Initiating Condition: FIRE potentially degrading the level of safety of the plant EAL: HU4.1 Unusual Event A FIRE is not extinguished within 15 min. of any of the following FIRE detection indications (Note 1): Report from the field (i.e., visual observation) Receipt of multiple (more than 1) fire alarms or indications (Note 13) Field verification of a single fire alarm AND The FIRE is located within any Table 1[2]H-1 area Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded.

Note 13 Bullet 2 of this EAL (multiple fire alarm indications) is not applicable for LOCAs or MSL breaks in containment.

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Table 1H-1 Unit 1 Fire Areas Reactor Building All elevations Auxiliary Building All elevations including: Penthouse/MSIV Room Exceptions: Boric Acid Mix Tank Room (Chem Add Area), 404' (157-B), EDG Exhaust Fan area on 386' (1-E and 2-E)

Turbine Building All elevations on the west side of Turbine Building and including: Pipechase under ICW Coolers, CRD Pump Pit/T-28 Room/Area under ICW Pumps 372' Elevation from non-vital switchgear area to Auxiliary Building wall at DR 56 Outside Areas Manholes adjacent to Startup #2 XFMR (MH-03/MH-04) Manholes adjacent to Intake Structure (MH-05/MH-06) Intake Structure (354' and 366') Diesel Fuel Vault Diesel Fuel Vault Pump Manholes (MH-09 and MH-10)

Table 2H-1 Unit 2 Fire Areas Reactor Building All elevations

Auxiliary Building All elevations including: MG Set Room, UNEPR, LNEPR, 2B-53 Room Auxiliary Building Extension MSIV Room

Turbine Building All elevations on the west side of Turbine Building and 372' Elevation from non-vital switchgear area to Auxiliary Building wall at DR 340 Outside Areas Intake Structure (354' and 366') Concrete Manhole East, NE of intake (2MH-01) Concrete Manhole East of Turbine Building next to train bay (2MH-03)

Diesel Fuel Vault Diesel Fuel Vault Pump Manholes (MH-09 and MH-10) to 0CAN091801

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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.

VALID - An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment.

Basis: 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 #1 The intent of the 15-minute duration is to size the FIRE and to discriminate against small FIRES that are readily extinguished (e.g., smoldering waste paper basket). In addition to alarms, other indications of a FIRE could be a drop in fire main pressure, automatic activation of a suppression system, etc.

Upon receipt, operators will take prompt actions to confirm the validity of an initial fire alarm, indication, or report. For EAL assessment purposes, the emergency declaration clock starts at the time that the initial alarm, indication, or report was received, and not the time that a subsequent verification action was performed. Similarly, the fire duration clock also starts at the time of receipt of the initial alarm, indication or report.

EAL #2 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.

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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 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.

EAL #3 In addition to a FIRE addressed by EAL #1 or EAL #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 FIRE occurring within the PROTECTED AREA of an ISFSI located outside the plant PROTECTED AREA. [Sentence for plants with an ISFSI outside the plant Protected Area]

EAL #4 If a FIRE within the plant or ISFSI [for plants with an ISFSI 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 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 #2, the 30-minutes to verify a single alarm is well within this worst-case 1-hour time period.

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Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via IC EAL CA6.1 or SA9SA9.1. The 15 minute requirement begins with a credible notification that a FIRE is occurring, or receipt of multiple VALID fire detection system alarms or field validation of a single fire alarm. The alarm is to be validated using available Control Room indications or alarms to prove that it is not spurious, or by reports from the field.

Table 1[2]H-1 Fire Areas are those areas that contain equipment necessary for safe operation and shutdown of the plant (ref. 1, 2).

Reference(s):

1. OP-1203.049 Fires in Areas Affecting Safe Shutdown 2. OP- 2203.049 Fires in Areas Affecting Safe Shutdown
3. NEI 99-01 HU4

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Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 4 - Fire

Initiating Condition: FIRE potentially degrading the level of safety of the plant EAL: HU4.2 Unusual Event Receipt of a single fire alarm (i.e., no other indications of a FIRE) (Note 14) AND The fire alarm is indicating a FIRE within any Table 1[2]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 Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded.

Note 14: During Modes 1 and 2, HU4.2 is not applicable to a single fire alarm in the Reactor Building.

Table 1H-1 Unit 1 Fire Areas Reactor Building All elevations Auxiliary Building All elevations including: Penthouse/MSIV Room Exceptions: Boric Acid Mix Tank Room (Chem Add Area), 404' (157-B), EDG Exhaust Fan area on 386' (1-E and 2-E)

Turbine Building All elevations on the west side of Turbine Building and including: Pipechase under ICW Coolers, CRD Pump Pit/T-28 Room/Area under ICW Pumps 372' Elevation from non-vital switchgear area to Auxiliary Building wall at DR 56 Outside Areas Manholes adjacent to Startup #2 XFMR (MH-03/MH-04) Manholes adjacent to Intake Structure (MH-05/MH-06) Intake Structure (354' and 366')

Diesel Fuel Vault to 0CAN091801

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Diesel Fuel Vault Pump Manholes (MH-09 and MH-10)

Table 2H-1 Unit 2 Fire Areas Reactor Building All elevations Auxiliary Building All elevations including: MG Set Room, UNEPR, LNEPR, 2B-53 Room Auxiliary Building Extension MSIV Room Turbine Building All elevations on the west side of Turbine Building and 372' Elevation from non-vital switchgear area to Auxiliary Building wall at DR 340 Outside Areas Intake Structure (354' and 366') Concrete Manhole East, NE of intake (2MH-01)

Concrete Manhole East of Turbine Building next to train bay (2MH-03) Diesel Fuel Vault Diesel Fuel Vault Pump Manholes (MH-09 and MH-10)

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.

VALID - An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment.

Basis: 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.

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EAL #1 The intent of the 15-minute duration is to size the FIRE and to discriminate against small FIRES that are readily extinguished (e.g., smoldering waste paper basket). In addition to alarms, other indications of a FIRE could be a drop in fire main pressure, automatic activation of a suppression system, etc.

Upon receipt, operators will take prompt actions to confirm the validity of an initial fire alarm, indication, or report. For EAL assessment purposes, the emergency declaration clock starts at the time that the initial alarm, indication, or report was received, and not the time that a subsequent verification action was performed. Similarly, the fire duration clock also starts at the time of receipt of the initial alarm, indication or report.

EAL #2 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.

This EAL is not applicable for the Reactor Building in Modes 1 and 2. The Reactor Building air flow design and Technical Specification requirements for operation of Reactor Building Fan Coolers are such that multiple smoke detectors would be expected to alarm for a fire in the Reactor Building. A fire in the Reactor Building in these modes would therefore be classified under EAL HU4.1.

If an actual FIRE is verified by a report from the field, then HU4.1 EAL #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.

EAL #3 In addition to a FIRE addressed by EAL #1 or EAL #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 FIRE occurring within the PROTECTED AREA of an ISFSI located outside the plant PROTECTED AREA.

[Sentence for plants with an ISFSI outside the plant Protected Area

] EAL #4 Basis-Related Fire Protection Requirements from Appendix R to 0CAN091801

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Appendix R to 10 CFR 50, states in part:

Criterion 3 of 10 CFR 50, Appendix A, states, in part:

"Structures, systems, and components important to safety shall be designed and located to minimize, consistent with other safety requirements, the probability and effect of fires and explosions.

" In this respect, noncombustible and heat resistant materials are used wherever practical throughout the unit, particularly in locations such as the containment and Control Room. Fire detection and fighting systems of appropriate capacity and capability are provided and designed to minimize the adverse effects of fires on SSCs important to safety. Firefighting systems are designed to assure that the rupture or inadvertent operation of a fire system does not significantly impair the safety capability of these structures, systems, and components.

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 is employed(G.2.c). As used in HU4.2EAL #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 EAL CA6.1 or SA9SA9.1. The 30-minute requirement begins upon receipt of a single VALID fire detection system alarm. The alarm is to be validated using available Control Room indications or alarms to prove that it is not spurious, or by reports from the field. Actual field reports must be made within the 30-minute time limit or a classification must be made. If a fire is verified to be occurring by field report, classification shall be made based on EAL HU4.1, with the 15-minute requirement beginning with the verification of the fire by field report.

Table 1[2]H-1 Fire Areas are those areas that contain equipment necessary for safe operation and shutdown of the plant (ref. 1, 2).

Reference(s):

1. OP-1203.049 Fires in Areas Affecting Safe Shutdown to 0CAN091801

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2. OP- 2203.049 Fires in Areas Affecting Safe Shutdown 3. NEI 99-01 HU4 to 0CAN091801

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Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 4 - Fire

Initiating Condition: FIRE potentially degrading the level of safety of the plant EAL: HU4.3 Unusual Event A FIRE within the PROTECTED AREA not extinguished within 60 min. of the initial report, alarm or indication (Note 1)

Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded.

Mode Applicability:

All Definition(s):

FIRE - Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred but is not required if large quantities of smoke and heat are observed.

PROTECTED AREA - An area clearly demarcated by a fence or building wall with an entrance portal that is regulated by Security Personnel to control access.

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 #1 The intent of the 15-minute duration is to size the FIRE and to discriminate against small FIRES that are readily extinguished (e.g., smoldering waste paper basket). In addition to alarms, other indications of a FIRE could be a drop in fire main pressure, automatic activation of a suppression system, etc.

Upon receipt, operators will take prompt actions to confirm the validity of an initial fire alarm, indication, or report. For EAL assessment purposes, the emergency declaration clock starts at the time that the initial alarm, indication, or report was received, and not the time that a subsequent verification action was performed. Similarly, the fire duration clock also starts at the time of receipt of the initial alarm, indication or report. to 0CAN091801

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EAL #2 This EAL addresses receipt of a single fire alarm, and the existence of a FIRE is not verified (i.e., proved or disproved) within 30-minutes of the alarm. Upon receipt, operators will take prompt actions to confirm the validity of a single fire alarm. For EAL assessment purposes, the 30-minute clock starts at the time that the initial alarm was received, and not the time that a subsequent verification action was performed.

A single fire alarm, absent other indication(s) of a FIRE, may be indicative of equipment failure or a spurious activation, and not an actual FIRE. For this reason, additional time is allowed to verify the validity of the alarm. The 30-minute period is a reasonable amount of time to determine if an actual FIRE exists; however, after that time, and absent information to the contrary, it is assumed that an actual FIRE is in progress.

If an actual FIRE is verified by a report from the field, then HU4.1 EAL #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. EAL #3 In addition to a FIRE addressed by EAL HU4.1 #1 or HU4.2EAL #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 FIRE occurring within the PROTECTED AREA of an ISFSI located outside the plant PROTECTED AREA. [

Sentence for plants with an ISFSI outside the plant Protected Area

] EAL #4 If a FIRE within the plant or ISFSI [for plants with an ISFSI 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 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." to 0CAN091801

Page 195 of 257 Attachment 1 - Emergency Action Level Technical Bases

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 #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 EAL CA6.1 or SA9SA8.1. Reference(s):

1. NEI 99-01 HU4

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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 PROTECTED AREA that requires firefighting support by an offsite fire response agency to extinguish Mode Applicability:

All Definition(s):

FIRE - Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred but is not required if large quantities of smoke and heat are observed.

PROTECTED AREA - An area clearly demarcated by a fence or building wall with an entrance portal that is regulated by Security Personnel to control access.

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 #1 The intent of the 15-minute duration is to size the FIRE and to discriminate against small FIRES that are readily extinguished (e.g., smoldering waste paper basket). In addition to alarms, other indications of a FIRE could be a drop in fire main pressure, automatic activation of a suppression system, etc.

Upon receipt, operators will take prompt actions to confirm the validity of an initial fire alarm, indication, or report. For EAL assessment purposes, the emergency declaration clock starts at the time that the initial alarm, indication, or report was received, and not the time that a subsequent verification action was performed. Similarly, the fire duration clock also starts at the time of receipt of the initial alarm, indication or report.

EAL #2 to 0CAN091801

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This EAL addresses receipt of a single fire alarm, and the existence of a FIRE is not verified (i.e., proved or disproved) within 30-minutes of the alarm. Upon receipt, operators will take prompt actions to confirm the validity of a single fire alarm. For EAL assessment purposes, the 30-minute clock starts at the time that the initial alarm was received, and not the time that a subsequent verification action was performed.

A single fire alarm, absent other indication(s) of a FIRE, may be indicative of equipment failure or a spurious activation, and not an actual FIRE. For this reason, additional time is allowed to verify the validity of the alarm. The 30-minute period is a reasonable amount of time to determine if an actual FIRE exists; however, after that time, and absent information to the contrary, it is assumed that an actual FIRE is in progress.

If an actual FIRE is verified by a report from the field, then HU4.1 EAL #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. EAL #3 In addition to a FIRE addressed by EAL #1 or EAL #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 FIRE occurring within the PROTECTED AREA of an ISFSI located outside the plant PROTECTED AREA.

[Sentence for plants with an ISFSI outside the plant Protected Area

] EAL #4 If a FIRE within the plant or ISFSI [for plants with an ISFSI 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 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 to 0CAN091801

Page 198 of 257 Attachment 1 - Emergency Action Level Technical Bases

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 #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 EAL CA6.1 or SA9SA9.1. Reference(s):

1. NEI 99-01 HU4

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Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 5 - Hazardous Gas

Initiating Condition: Gaseous release IMPEDING access to equipment necessary for normal plant operations, cooldown or shutdown

EAL: HA5.1 Alert Release of a toxic, corrosive, asphyxiant or flammable gas into any Table 1[2]H-2 room or area 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 1H-2 Unit 1 Safe Operation & Shutdown Rooms/Areas Room/Area Mode A-4 Switchgear Room 3, 4 Upper North Electrical Penetration Room 3, 4 Lower South Electrical Equipment Room 3, 4 Table 2H-2 Unit 2 Safe Operation & Shutdown Rooms/Areas Room/Area Mode Aux Building 317' Emergency Core Cooling Rooms 3, 4 Aux Building 317' Tendon Gallery Access 3, 4 Aux Building 335' Charging Pumps / MCC 2B-52 3, 4 Aux Building 354' MCC 2B-62 Area 3, 4 Emergency Diesel Generator Corridor 3, 4 Lower South Piping Penetration Room 3, 4 Aux Building 386' Containment Hatch 3, 4 to 0CAN091801

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Mode Applicability:

3 - Hot Standby, 4 - Hot Shutdown Definition(s):

IMPEDE(D) - Personnel access to a room or area is hindered to an extent that extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., requiring use of protective equipment, such as SCBAs, that is not routinely employed).

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 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, 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 r oom 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).

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

If the equipment in the listed room or area was already inoperable, or out-of-service, before the event occurred, then no emergency should be declared since the event will have no adverse impact beyond that already allowed by Technical Specifications at the time of the event.

An asphyxiant is a gas capable of reducing the level of oxygen in the body to dangerous levels.

Most commonly, asphyxiants work by merely displacing air in an enclosed environment. This reduces the concentration of oxygen below the normal level of around 19%, which can lead to breathing difficulties, unconsciousness or even death.

This EAL does not apply to firefighting activities that generate smoke and that automatically or manually activate a fire suppression system in an area , or to intentional inerting of containment. (BWR only). The list of plant rooms or areas with entry-related mode applicability identified specify those rooms or areas that contain equipment which require a manual/local action as specified in operating procedures used for normal plant operation, cooldown and shutdown. Rooms or areas in which actions of a contingent or emergency nature would be performed (e.g., an action to address an off-normal or emergency condition such as emergency repairs, corrective measures or emergency operations) are not included. In addition, the list specifies the plant mode(s) during which entry would be required for each room or area (ref. 1).

Escalation of the emergency classification level would be via Recognition Category A, C or F

ICs. EAL HA5.1 mode applicability has been limited to the mode limitations of Table 1[2]H-2 (Modes 3 and 4 only). Reference(s):

1. Attachment 2 Safe Operation & Shutdown Areas Tables 1[2]A-3 & 1[2]H-2 Bases 2. NEI 99-01 HA5

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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 alternate locations Mode Applicability:

All Definition(s):

None 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.

Transfer of plant control begins when the last licensed operator leaves the Control Room.

Escalation of the emergency classification level would be via IC HS6.

Reference(s):

1. OP-1203.002 Alternate Shutdown 2. OP- 2203.014 Alternate Shutdown
3. NEI 99-01 HA6

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Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 6 - Control Room Evacuation

Initiating Condition: Inability to control a key safety function from outside the Control Room EAL: HS6.1 Site Area Emergency An event has resulted in plant control being transferred from the Control Room to alternate locations AND Control of any of the following key safety functions is not re-established within 15 min. (Note 1): Reactivity (Modes 1, 2 and 3 only) Core cooling RCS heat removal Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded.

Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown, 5 - Cold Shutdown, 6 - Refueling

Definition(s):

None 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 judgment. The Emergency Director is expected to make a reasonable, informed judgment within 15 (the site-specific time for transfer) minutes whether or not the operating staff has control of key safety functions from the remote safe shutdown location(s).

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Transfer of plant control and the time period to establish control begins when the last licensed operator leaves the Control Room.

Escalation of the emergency classification level would be via IC FG1 or CG1

Reference(s):

1. OP-1203.002 Alternate Shutdown 2. OP-2203.014 Alternate Shutdown
3. NEI 99-01 HS6

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Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 7 - Emergency Director Judgment

Initiating Condition: Other conditions exist that in the judgment of the Emergency Director warrant declaration of a UE

EAL: HU7.1 Unusual Event Other conditions exist which, in the judgment of the Emergency Director, 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 10 CFR 50.2).

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: 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 to fall under the emergency classification level description for a n UNUSUAL EVENTNOUE. Reference(s):

1. NEI 99-01 HU7

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Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 7 - Emergency Director Judgment

Initiating Condition: Other conditions exist that in the judgment of the Emergency Director warrant declaration of an ALERT

EAL: HA7.1 Alert Other conditions exist which, in the judgment of the Emergency Director, 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):

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 ANO 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 ANO. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the SECURITY OWNER CONTROLLED AREA).

PROJECTILE - An object directed toward a Nuclear Power Plant that could cause concern for its continued operability, reliability, or personnel safety.

PROTECTED AREA - An area clearly demarcated by a fence or building wall with an entrance portal that is regulated by Security Personnel to control access.

SECURITY OWNER CONTROLLED AREA - The SOCA is the area demarcated as a Vehicle Barrier System (VBS) consisting of passive elements including a series of large concrete blocks on the inside of a delay fence with early warning capabilities. The SOCA is the area between the SOCA Fence and the PROTECTED AREA Boundary.

to 0CAN091801

Page 207 of 257 Attachment 1 - Emergency Action Level Technical Bases

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 to fall under the emergency classification level description for an ALERT.

Reference(s):

1. NEI 99-01 HA7

to 0CAN091801

Page 208 of 257 Attachment 1 - Emergency Action Level Technical Bases

Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 7 - Emergency Director Judgment

Initiating Condition: Other conditions exist that in the judgment of the Emergency Director warrant declaration of a SITE AREA EMERGENCY

EAL: HS7.1 Site Area Emergency Other conditions exist which, in the judgment of the Emergency Director, 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 leve ls which exceed EPA Protective Action Guideline exposure levels beyond the SITE BOUNDARY.

Mode Applicability:

All Definition(s):

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 ANO 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 ANO. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the SECURITY OWNER CONTROLLED AREA).

PROJECTILE - An object directed toward a Nuclear Power Plant that could cause concern for its continued operability, reliability, or personnel safety.

PROTECTED AREA - An area clearly demarcated by a fence or building wall with an entrance portal that is regulated by Security Personnel to control access.

SECURITY OWNER CONTROLLED AREA - The SOCA is the area demarcated as a Vehicle Barrier System (VBS) consisting of passive elements including a series of large concrete blocks on the inside of a delay fence with early warning capabilities. The SOCA is the area between the SOCA Fence and the PROTECTED AREA Boundary. to 0CAN091801

Page 209 of 257 Attachment 1 - Emergency Action Level Technical Bases

SITE BOUNDARY - That boundary defined by a 1046 meter (0.65 mile) radius around the plant.

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 to fall under the emergency classification level description for a SITE AREA EMERGENCY.

Reference(s):

1. NEI 99-01 HS7

to 0CAN091801

Page 210 of 257 Attachment 1 - Emergency Action Level Technical Bases

Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 7 - Emergency Director Judgment

Initiating Condition: Other conditions exist that in the judgment of the Emergency Director warrant declaration of a GENERAL EMERGENCY

EAL: HG7.1 General Emergency Other conditions exist which, in the judgment of the Emergency Director, 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):

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 ANO 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 ANO. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the SECURITY 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.

PROJECTILE - An object directed toward a Nuclear Power Plant that could cause concern for its continued operability, reliability, or personnel safety.

PROTECTED AREA - An area clearly demarcated by a fence or building wall with an entrance portal that is regulated by Security Personnel to control access.

to 0CAN091801

Page 211 of 257 Attachment 1 - Emergency Action Level Technical Bases

SECURITY OWNER CONTROLLED AREA - The SOCA is the area demarcated as a Vehicle Barrier System (VBS) consisting of passive elements including a series of large concrete blocks on the inside of a delay fence with early warning capabilities. The SOCA is the area between the SOCA Fence and the PROTECTED AREA Boundary.

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 to fall under the emergency classification level description for a GENERAL EMERGENCY.

Reference(s):

1. NEI 99-01 HG7

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Page 212 of 257 Attachment 1 - Emergency Action Level Technical Bases

Category S - System Malfunction EAL Group: Hot Conditions (RCS temperature > 200°F); EALs in this category are applicable only in one or more hot operating modes.

Numerous system-related equipment failure event s 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 Vital AC Power Loss of vital 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 vital 4.16 KV buses.
2. Loss of Vital DC Power Loss of emergency electrical power can compromise plant SAFETY SYSTEM operability including decay heat removal and emergency core cooling systems which may be necessary to ensure fission product barrier integrity. This category includes loss of vital plant 125V DC power sources.
3. Loss of Control Room Indications Certain events that degrade plant operator ability to effectively assess plant conditions within the plant warrant emergency classification. Losses of indicators are in this subcategory.
4. RCS Activity During normal operation, reactor coolant fission product activity is very low. Small concentrations of fission products in the coolant are primarily from the fission of tramp uranium in the fuel clad or minor perforations in the clad itself. Any significant rise from these base-line levels (2% - 5% clad failures) is indicative of fuel failures and is covered under the Fission Product Barrier Degradation category. However, lesser amounts of clad damage may result in coolant activity exceeding Technical Specification limits. These fission products will be circulated with the reactor coolant and can be detected by coolant sampling.
5. RCS Leakage The reactor vessel provides a volume for the coolant that covers the reactor core. The reactor pressure vessel and associated pressure piping (reactor coolant system) together provide a barrier to limit the release of radioactive material should the reactor fuel clad integrity fail. Excessive RCS leakage greater than Technical Specification limits indicates potential pipe cracks that may propagate to an extent threatening fuel clad, RCS and containment integrity. to 0CAN091801

Page 213 of 257 Attachment 1 - Emergency Action Level Technical Bases

6. RPS Failure This subcategory includes events related to failure of the Reactor Protection System (RPS) to initiate and complete reactor trips. In the plant licensing basis, postulated failures of the RPS to complete a reactor trip comprise a specific set of analyzed events referred to as Anticipated Transient Without Scram (ATWS) events. For EAL classification, however, ATWS is intended to mean any trip failure event that does not achieve reactor shutdown. If RPS actuation fails to assure reactor shutdown, positive control of reactivity is at risk and could cause a threat to fuel clad, RCS and containment integrity.
7. Loss of Communications Certain events that degrade plant operator ability to effectively communicate with essential personnel within or external to the plant warrant emergency classification.
8. Containment Failure Failure of containment isolation capability (under conditions in which the containment is not currently challenged) warrants emergency classification. Failure of containment pressure control capability also warrants emergency classification.
9. Hazardous Event Affecting Safety Systems Various natural and technological events that result in degraded plant safety system performance or significant VISIBLE DAMAGE warrant emergency classification under this subcategory.

to 0CAN091801

Page 214 of 257 Attachment 1 - Emergency Action Level Technical Bases

Category: S - System Malfunction Subcategory: 1 - Loss of Vital AC Power

Initiating Condition: Loss of all offsite AC power capability to vital buses for 15 minutes or longer EAL: SU1.1 Unusual Event Loss of all offsite AC power capability, Table 1[2]S-1, to vital 4.16 KV buses A3 [2A3] and A4 [2A4] for 15 min. (Note 1)

Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded.

Table 1S-1 Unit 1 AC Power Sources Offsite Startup Transformer No. 1 Startup Transformer No. 2 Unit Auxiliary Transformer (from 22 KV switchyard)

Onsite Unit Auxiliary Transformer (main generator via main transformer) DG1 DG2 AAC Gen to 0CAN091801

Page 215 of 257 Attachment 1 - Emergency Action Level Technical Bases

Table 2S-1 Unit 2 AC Power Sources Offsite Startup Transformer No. 3 Startup Transformer No. 2 Unit Auxiliary Transformer (backfed from main transformer)

Onsite Unit Auxiliary Transformer (main generator via main transformer) 2DG1 2DG2 AAC Gen Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):

None

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 vital buses. This condition represents a potential reduction in the level of safety of the plant.

For emergency classification purposes, "capability" means that an offsite AC power source(s) is available to the emergency vital 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.

Reference(s):

1. 1SAR Figure 8-1 Station Single Line Diagram 2. OP-1202.007 Degraded Power 3. OP-1202.008 Blackout
4. OP-2104.037 Alternate AC Diesel Generator Operations to 0CAN091801

Page 216 of 257 Attachment 1 - Emergency Action Level Technical Bases

5. 2SAR Figure 8.3-1 Station Single Line Diagram 6. OP-2202.007 Loss of Off-Site Power 7. OP-2202.008 Station Blackout 8. OP-2107.006 Backfeed of Unit Auxiliary Transformer
9. NEI 99-01 SU1

to 0CAN091801

Page 217 of 257 Attachment 1 - Emergency Action Level Technical Bases

Category: S - System Malfunction Subcategory: 1 - Loss of Vital AC Power

Initiating Condition: Loss of all but one AC power source to vital buses for 15 minutes or longer EAL: SA1.1 Alert AC power capability, Table 1[2]S-1, to vital 4.16 KV buses A3 [2A3] and A4 [2A4] reduced to a single power source for 15 min. (Note 1) AND Any additional single power source failure will result in loss of all AC power to SAFETY SYSTEMS Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded.

Table 1S-1 Unit 1 AC Power Sources Offsite Startup Transformer No. 1 Startup Transformer No. 2 Unit Auxiliary Transformer (from 22 KV switchyard)

Onsite Unit Auxiliary Transformer (main generator via main transformer) DG1 DG2 AAC Gen to 0CAN091801

Page 218 of 257 Attachment 1 - Emergency Action Level Technical Bases

Table 2S-1 Unit 2 AC Power Sources Offsite Startup Transformer No. 3 Startup Transformer No. 2 Unit Auxiliary Transformer (backfed from main transformer)

Onsite Unit Auxiliary Transformer (main generator via main transformer) 2DG1 2DG2 AAC Gen Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):

SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10 CFR 50.2).

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

related equipment. This IC provides an escalation path from IC SU1.

An "AC power source" is a source recognized in AOPs and EOPs, and capable of supplying required power to a emergency vital bus. Some examples of this condition are presented below. to 0CAN091801

Page 219 of 257 Attachment 1 - Emergency Action Level Technical Bases

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 vital emergency power sources (e.g., onsite diesel generators) with a single train of emergency vital buses being back-fed from the unit main generator.

A loss of vital emergency power sources (e.g., onsite diesel generators) with a single train of vital 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.

This EAL is the hot condition equivalent of the cold condition EAL CU2.1.

Reference(s):

1. 1SAR Figure 8-1 Station Single Line Diagram 2. OP-1202.007 Degraded Power
3. OP-1202.008 Blackout
4. OP-2104.037 Alternate AC Diesel Generator Operations
5. 2SAR Figure 8.3-1 Station Single Line Diagram 6. OP-2202.007 Loss of Off-Site Power 7. OP-2202.008 Station Blackout
8. OP-2107.006 Backfeed of Unit Auxiliary Transformer
9. NEI 99-01 SA1 to 0CAN091801

Page 220 of 257 Attachment 1 - Emergency Action Level Technical Bases

Category: S - System Malfunction Subcategory: 1 - Loss of Vital AC Power

Initiating Condition: Loss of all offsite power and all onsite AC power to vital buses for 15 minutes or longer

EAL: SS1.1 Site Area Emergency Loss of all offsite and all onsite AC power to vital 4.16 KV buses A3 [2A3] and A4 [2A4] for 15 min. (Note 1)

Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded.

Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown

Definition(s):

SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10 CFR 50.2).

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:

Although the AAC may be considered available, it will not prevent declaration of this EAL unless it is powering a vital bus within the 15 minute time period of the EAL.

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. Mitigative strategies using non-safety related power sources (FLEX generators, etc.) may be effective in supplying power to these buses. These power sources must be controlled in accordance with abnormal or emergency operating procedures, or beyond design basis to 0CAN091801

Page 221 of 257 Attachment 1 - Emergency Action Level Technical Bases

accident response guidelines (e.g., FLEX support guidelines) and must be capable (alone or in combination) of supplying power for long term decay heat removal systems.

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 IC s AG1, FG1 or SG1.

This EAL is the hot condition equivalent of the cold condition EAL CA2.1.

Reference(s):

1. 1SAR Figure 8-1 Station Single Line Diagram 2. OP-1202.007 Degraded Power
3. OP-1202.008 Blackout
4. OP-2104.037 Alternate AC Diesel Generator Operations 5. 2SAR Figure 8.3-1 Station Single Line Diagram 6. OP-2202.007 Loss of Off-Site Power
7. OP-2202.008 Station Blackout
8. OP-2107.006 Backfeed of Unit Auxiliary Transformer
9. NEI 99-01 SS1 to 0CAN091801

Page 222 of 257 Attachment 1 - Emergency Action Level Technical Bases

Category: S - System Malfunction Subcategory: 1 - Loss of Vital AC Power

Initiating Condition: Prolonged loss of all offsite and all onsite AC power to vital buses EAL: SG1.1 General Emergency Loss of all offsite and all onsite AC power to vital 4.16 KV buses A3 [2A3] and A4 [2A4] AND EITHER:

Restoration of at least one vital 4.16 KV bus in <

4 hours
4.62963e-5 days
0.00111 hours
6.613757e-6 weeks
1.522e-6 months

is not likely (Note 1) CETs > 1200°F Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded.

Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown

Definition(s):

SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10 CFR 50.2).

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: This IC addresses a prolonged loss of all power sources to AC emergency vital 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.

Mitigative strategies using non-safety related power sources (FLEX generators, etc.) may be effective in supplying power to these buses. These power sources must be controlled in accordance with abnormal or to 0CAN091801

Page 223 of 257 Attachment 1 - Emergency Action Level Technical Bases

emergency operating procedures, or beyond design basis accident response guidelines (e.g., FLEX support guidelines) and must be capable (alone or in combination) of supplying power for long term decay heat removal systems. 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 vital 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 a n increased greater likelihood of challenges to multiple fission product barriers.

4 hours
4.62963e-5 days
0.00111 hours
6.613757e-6 weeks
1.522e-6 months

is the site-specific SBO coping analysis time (ref. 4, 5).

The estimate for restoring at least one emergency vital 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.

Reference(s):

1. OP-1202.005 Inadequate Core Cooling 2. OP-2202.009 Functional Recovery
3. OP-2202.011 Lower Mode Functional Recovery
4. Unit 1 Calculation 85-E-0072-02 Time from Loss of All AC Power to Loss of Subcooling 5. Unit 2 Calculation 85-E-0072-01 Time from Loss of All AC Power to Loss of Subcooling 6. 1SAR Figure 8-1 Station Single Line Diagram
7. OP-1202.007 Degraded Power
8. OP-1202.008 Blackout
9. OP-2104.037 Alternate AC Diesel Generator Operations 10. 2SAR Figure 8.3-1 Station Single Line Diagram 11. OP-2202.007 Loss of Off-Site Power
12. OP-2202.008 Station Blackout
13. OP-2107.006 Backfeed of Unit Auxiliary Transformer
14. NEI 99-01 SG1 to 0CAN091801

Page 224 of 257 Attachment 1 - Emergency Action Level Technical Bases

Category: S - System Malfunction Subcategory: 1 - Loss of Vital AC Power

Initiating Condition: Loss of all vital AC and vital DC power sources for 15 minutes or longer EAL: SG1.2 General Emergency Loss of all offsite and all onsite AC power to vital 4.16 KV buses A3 [2A3] and A4 [2A4] for 15 min. (Note 1)

AND Indicated voltage is < 105 VDC on D01 [2D01] and D02 [2D02] vital 125 VDC buses for 15 min. (Note 1)

Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded.

Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown

Definition(s):

SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10 CFR 50.2).

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: Unit 1 batteries D06 and D07 and Unit 2 batteries 2D11 and 2D12 contain 58 cells each with a minimum cell voltage of 1.81 V or 105 VDC (ref. 9, 10).

This IC addresses a concurrent and prolonged loss of both vital AC and Vital DC power. A loss of all vital AC power compromises the performance of all SAFETY SYSTEMS requiring electric to 0CAN091801

Page 225 of 257 Attachment 1 - Emergency Action Level Technical Bases

power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink.

Mitigative strategies using non-safety related power sources (FLEX generators, etc.) may be effective in supplying power to these buses. These power sources must be controlled in accordance with abnormal or emergency operating procedures, or beyond design basis accident response guidelines (e.g., FLEX support guidelines) and must be capable (alone or in combination) of supplying power for long term decay heat removal systems. A loss of vital DC power compromises the ability to monitor and control SAFETY SYSTEMS. A sustained loss of both vital 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. Reference(s):

1. 1SAR Figure 8-1 Station Single Line Diagram 2. OP-1202.007 Degraded Power 3. OP-1202.008 Blackout 4. OP-2104.037 Alternate AC Diesel Generator Operations
5. 2SAR Figure 8.3-1 Station Single Line Diagram
6. OP-2202.007 Loss of Off-Site Power
7. OP-2202.008 Station Blackout 8. OP-2107.006 Backfeed of Unit Auxiliary Transformer 9. 1SAR 8.3.2.1.1 Batteries
10. 2SAR 8.3.2.1.1 Batteries
11. OP-1203.036 Loss of 125V DC
12. OP-2203.037 Loss of 125V DC
13. 2SAR Figure 8.3-6 Low Voltage Safety System Power Supplies 14. NEI 99-01 SG8

to 0CAN091801

Page 226 of 257 Attachment 1 - Emergency Action Level Technical Bases

Category: S - System Malfunction Subcategory: 2 - Loss of Vital DC Power

Initiating Condition: Loss of all vital DC power for 15 minutes or longer EAL: SS2.1 Site Area Emergency Indicated voltage is < 105 VDC on D01 [2D01] and D02 [2D02] vital 125 VDC buses for 15 min. (Note 1)

Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded.

Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown

Definition(s):

SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10 CFR 50.2).

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: Unit 1 batteries D06 and D07 and Unit 2 batteries 2D11 and 2D12 contain 58 cells each with a minimum cell voltage of 1.81 V or 105 VDC (ref.

2 , 3). 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 IC s AG1, FG1 or SG1SG8. to 0CAN091801

Page 227 of 257 Attachment 1 - Emergency Action Level Technical Bases

This EAL is the hot condition equivalent of the cold condition EAL CU4.1.

Reference(s):

1. 1SAR Figure 8-1 Station Single Line Diagram 2. 1SAR 8.3.2.1.1 Batteries 3. 2SAR 8.3.2.1.1 Batteries 4. OP-1203.036 Loss of 125V DC
5. OP-2203.037 Loss of 125V DC
6. 2SAR Figure 8.3-6 Low Voltage Safety System Power Supplies
7. NEI 99-01 SS8 to 0CAN091801

Page 228 of 257 Attachment 1 - Emergency Action Level Technical Bases

Category: S - System Malfunction Subcategory: 3 - Loss of Control Room Indications

Initiating Condition: UNPLANNED loss of Control Room indications for 15 minutes or longer EAL: SU3.1 Unusual Event An UNPLANNED event results in the inability to monitor one or more Table 1[2]S-2 parameters from within the Control Room for 15 min. (Note 1)

Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded.

Table 1[2]S-2 Safety System Parameters Reactor power RCS level RCS pressure CET temperature Level in at least one S/G EFW flow to at least one S/G Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):

SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10 CFR 50.2).

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. to 0CAN091801

Page 229 of 257 Attachment 1 - Emergency Action Level Technical Bases

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

Basis:

This IC addresses the difficulty associated with monitoring normal plant conditions without the ability to obtain SAFETY SYSTEM parameters from within the Control Room. This condition is a precursor to a more significant event and represents a potential degradation in the level of safety of the plant.

As used in this EAL, an "inability to monitor" means that values for one or more of the listed parameters cannot be determined from within the Control Room. This situation would require a loss of all of the Control Room sources for the given parameter(s). For example, the reactor power level cannot be determined from any analog, digital and recorder source within the Control Room.

An event involving a loss of plant indications, annunciators and/or display systems is evaluated in accordance with 10 CFR 50.72 (and associated guidance in NUREG-1022) to determine if an NRC event report is required. The event would be reported if it significantly impaired the capability to perform emergency assessments. In particular, emergency assessments necessary to implement abnormal operating procedures, emergency operating procedures, and emergency plan implementing procedures addressing emergency classification, accident assessment, or protective action decision-making.

This EAL is focused on a selected subset of plant parameters associated with the key safety functions of reactivity control, core cooling

[PWR] / RPV level [

BWR] 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 [BWR] 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 EAL SA3.1SA2. Reference(s):

1. 1SAR 7.5 Safety-Related Display Instrumentation 2. 2SAR 7.5 Safety-Related Display Instrumentation 3. NEI 99-01 SU2 to 0CAN091801

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Category: S - System Malfunction Subcategory: 3 - Loss of Control Room Indications

Initiating Condition: UNPLANNED loss of Control Room indications for 15 minutes or longer with a significant transient in progress

EAL: SA3.1 Alert An UNPLANNED event results in the inability to monitor one or more Table 1[2]S-2 parameters from within the Control Room for 15 min. (Note 1) AND Any significant transient is in progress, Table 1[2]S-3 Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded.

Table 1[2]S-2 Safety System Parameters Reactor power RCS level RCS pressure CET temperature Level in at least one S/G EFW flow to at least one S/G Table 1[2]S-3 Significant Transients Reactor trip Runback > 25% thermal power Electrical load rejection > 25% electrical load Safety injection actuation Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown

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Page 231 of 257 Attachment 1 - Emergency Action Level Technical Bases

Definition(s):

SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10 CFR 50.2):

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.

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

Basis: This IC addresses the difficulty associated with monitoring rapidly changing plant conditions during a transient without the ability to obtain SAFETY SYSTEM parameters from within the Control Room. During this condition, the margin to a potential fission product barrier challenge is reduced. It thus represents a potential substantial degradation in the level of safety of the plant.

As used in this EAL, an "inability to monitor" means that values for one or more of the listed parameters cannot be determined from within the Control Room. This situation would require a loss of all of the Control Room sources for the given parameter(s). For example, the reactor power level cannot be determined from any analog, digital and recorder source within the Control Room.

An event involving a loss of plant indications, annunciators and/or display systems is evaluated in accordance with 10 CFR 50.72 (and associated guidance in NUREG-1022) to determine if an NRC event report is required. The event would be reported if it significantly impaired the capability to perform emergency assessments. In particular, emergency assessments necessary to implement abnormal operating procedures, emergency operating procedures, and emergency plan implementing procedures addressing emergency classification, accident assessment, or protective action decision-making.

This EAL is focused on a selected subset of plant parameters associated with the key safety functions of reactivity control, core cooling

[PWR] / RPV level [

BWR] 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 [

BWR] cannot be determined from the to 0CAN091801

Page 232 of 257 Attachment 1 - Emergency Action Level Technical Bases

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 s FS1 or IC AS1. Reference(s):

1. 1SAR 7.1.3 Engineered Safeguards Actuation System 2. 2SAR 7.3 Engineered Safety Features Systems 3. NEI 99-01 SA2

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Category: S - System Malfunction Subcategory: 4 - RCS Activity

Initiating Condition: RCS activity greater than Technical Specification allowable limits EAL: SU4.1 Unusual Event Failed Fuel Iodine radiation monitor RI-1237S [2RITS-4806B] > 9.0 E5 cpm Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown

Definition(s):

None 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 IC s FA1 or the Recognition Category A ICs.

Unit 1 RE-1237S, Failed Fuel Monitor, is in the letdown system to monitor the letdown line for evidence of fuel damage.

Unit 2 specific activity monitor 2RITS-4806B monitors the Letdown fluid for the presence of Iodine-131.

A monitor reading corresponding to the instantaneous dose equivalent I-131 value of 60 uCi/gm is determined by multiplying by 30 the monitor reading listed in the table in OP-1203.019[OP-2203.020] that represents a projected 2.0 uCi/gm I-131 RCS activity in order to correlate to a Tech Spec instantaneous limit of 60 uCi/gm dose equivalent I-131 for the EAL (ref. 2, 5). This yields values of 3.1E6 cpm for Unit 1 and 3.9E6 cpm for Unit 2.

The top of scale of the monitor is 1E6.

The EAL value is set at 9.0 E5 cpm for both units which is 90% of the top of the scale.

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Reference(s):

1. 1SAR Table 11-7 2. OP-1203-019 High Activity in Reactor Coolant
3. Unit 1 Technical Specifications LCO 3.4.12 RCS Specific Activity 4. 2SAR 9.3.5 Failed Fuel Detection System
5. OP-2203.020 High Activity in RCS 6. OP- 2203.012L ANNUNCIATOR 2K12 CORRECTIVE ACTION, A-1 7. Unit 2 Technical Specifications LCO 3.4.

8 Reactor Coolant System Specific Activity 8. NEI 99-01 SU3

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Category: S - System Malfunction Subcategory: 4 - RCS Activity

Initiating Condition: RCS activity greater than Technical Specification allowable limits EAL: SU4.2 Unusual Event RCS sample activity > 1.0 µCi/gm dose equivalent I-131 for >

48 hours
5.555556e-4 days
0.0133 hours
7.936508e-5 weeks
1.8264e-5 months

(Note 1) OR RCS sample activity > 60 µCi/gm dose equivalent I-131 OR RCS sample activity > 2200[3100] µCi/gm dose equivalent Xe-133 for >

48 hours
5.555556e-4 days
0.0133 hours
7.936508e-5 weeks
1.8264e-5 months

(Note 1)

Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded.

Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):

None 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 IC s FA1 or the Recognition Category A ICs.

Reference(s):

1. Unit 1 Technical Specifications LCO 3.4.12 RCS Specific Activity 2. Unit 2 Technical Specifications LCO 3.4.

8 Reactor Coolant System Specific Activity 3. NEI 99-01 SU3 to 0CAN091801

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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. (Note 1) OR RCS identified leakage > 25 gpm for 15 min. (Note 1)

OR Reactor coolant leakage to a location outside containment > 25 gpm for 15 min. (Note 1)

Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded.

Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):

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

Basis: Failure to isolate the leak (from the Control Room or locally) within 15 minutes, or if known that the leak cannot be isolated within 15 minutes, from the start of the leak requires immediate classification.

Steam generator tube leakage is identified RCS leakage.

This IC addresses RCS leakage which may be a precursor to a more significant event. In this case, RCS leakage has been detected and operators, following applicable procedures, have been unable to promptly isolate the leak. This condition is considered to be a potential degradation of the level of safety of the plant.

The first and second EAL conditions EAL #1 and EAL #2 are focused on a loss of mass from the RCS due to "unidentified leakage", "pressure boundary leakage" or "identified leakage" (as these leakage types are defined in the plant Technical Specifications). The third condition EAL

  1. 3 addresses a n RCS mass loss caused by an UNISOLABLE leak through an interfacing to 0CAN091801

Page 237 of 257 Attachment 1 - Emergency Action Level Technical Bases

system. These conditions EALs thus apply to leakage into the containment, a secondary-side system (e.g., steam generator tube leakage in a PWR) or a location outside of containment.

The leak rate values for each condition EAL were selected because they are usually observable with normal Control Room indications. Lesser values typically require time-consuming calculations to determine (e.g., a mass balance calculation). The first condition EAL #1 uses a lower value that reflects the greater significance of unidentified or pressure boundary leakage.

The release of mass from the RCS due to the as-designed/expected operation of a relief valve does not warrant an emergency classification.

For PWRs, aAn emergency classification would be required if a mass loss is caused by a relief valve that is not functioning as designed/expected (e.g., a relief valve sticks open and the line flow cannot be isolated).

For BWRs, 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 or F.

Reference(s):

1. Unit 1 and Unit 2 Technical Specifications Section 1.1 Definitions 2. NEI 99-01 SU4

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Category: S - System Malfunction Subcategory: 6 - RPS Failure

Initiating Condition: Automatic or manual trip fails to shut down the reactor EAL: SU6.1 Unusual Event An automatic trip did not shut down the reactor as indicated by reactor power > 5% after any RPS setpoint is exceeded AND A subsequent automatic trip or manual trip action taken at the reactor control console (C03 [2C03/2C14]) (manual reactor trip pushbuttons or DROPS[DSS]) is successful in shutting down the reactor as indicated by reactor power 5% (Note 8)

Note 8: A manual action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron injection strategies.

Mode Applicability:

1 - Power Operation

Definition(s):

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

Basis:

This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor (trip [PWR] / scram [

BWR]) that results in a reactor shutdown, and either a subsequent operator manual action taken at the reactor control consoles or an automatic (trip [PWR] / scram [

BWR]) 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.

In the event that the operator identifies a reactor trip is IMMINENT and initiates a successful manual reactor trip before the automatic trip setpoint is reached, no declaration is required. The successful manual trip of the reactor before it reaches its automatic trip setpoint or reactor trip signals caused by instrumentation channel failures do not lead to a potential fission product barrier loss.

Following the failure o nf an automatic reactor (trip[PWR] / scram [

BWR]), operators will promptly initiate manual actions at the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor (trip[PWR] / scram [

BWR])). If these manual actions are successful in shutting to 0CAN091801

Page 239 of 257 Attachment 1 - Emergency Action Level Technical Bases

down the reactor, core heat generation will quickly fall to a level within the capabilities of the plant's decay heat removal systems.

If an initial manual reactor (trip [PWR] / scram [BWR]) 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[PWR] / scram [BWR]) using a different switch). Depending upon several factors, the initial or subsequent effort to manually (trip [PWR] / scram

[BWR]) the reactor, or a concurrent plant condition, may lead to the generation of an automatic reactor (trip [PWR] / scram [BWR]) signal. If a subsequent manual or automatic (trip [PWR] / scram [BWR]) is successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plant's decay heat removal systems.

A manual action at the reactor control consoles is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor

(trip[PWR] / scram [

BWR])). 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. [

BWR] The plant response to the failure of an automatic or manual reactor (trip [PWR] / 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 SA5SA6. 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 reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.

Should a reactor (trip [PWR] / scram [BWR]) signal be generated as a result of plant work (e.g., RPS setpoint testing), the following classification guidance should be applied.

If the signal generated as a result of plant work causes a plant transient that results in a condition that should have included an automatic reactor trip

[PWR] / scram [BWR]) and the RPS fails to automatically shutdown the reactor, then this IC and associated the EALs are applicable, and should be evaluated.

If the signal generated as a result of plant work does not cause a plant transient and the (trip [PWR] / scram [BWR])

failure is determined through other means (e.g., assessment of test results), then this IC and associated the EALs are not applicable and no classification is warranted.

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Reference(s):

1. Unit 1 Technical Specifications Table 3.3.1-1 Reactor Protection System Instrumentation 2. Unit 2 Technical Specifications Table 3.3-1 Reactor Protective Instrumentation
3. Unit 1 and Unit 2 Technical Specifications Table 1.1-1 Modes
4. OP-1202.001 Reactor Trip
5. OP-2202.001 Standard Post Trip Actions 6. NEI 99-01 SU5 to 0CAN091801

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Category: S - System Malfunction Subcategory: 6 - RPS Failure

Initiating Condition: Automatic or manual trip fails to shut down the reactor EAL: SU6.2 Unusual Event A manual trip did not shut down the reactor as indicated by reactor power > 5% after any manual trip action was initiated AND A subsequent automatic trip or manual trip action taken at the reactor control console (C03 [2C03/2C14]) (manual reactor trip pushbuttons or DROPS[DSS]) is 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 Operation

Definition(s):

None Basis: This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor (trip [PWR] / scram [BWR]) that results in a reactor shutdown, and either a subsequent operator manual action taken at the reactor control consoles or an automatic (trip [PWR] / scram [BWR])

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.

This EAL addresses a failure of a manually initiated trip in the absence of having exceeded an automatic RPS trip setpoint and a subsequent automatic or manual trip is successful in shutting down the reactor.

Following the failure on an automatic reactor (trip [PWR] / scram [BWR]), operators will promptly initiate manual actions at the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor (trip [PWR] / scram [BWR])). 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.

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If an initial manual reactor (trip [PWR] / scram [BWR]) 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 [PWR] / scram [BWR])

using a different switch). Depending upon several factors, the initial or subsequent effort to manually (trip [PWR] / scram

[BWR]) the reactor, or a concurrent plant condition, may lead to the generation of an automatic reactor (trip [PWR] / scram [BWR]) signal. If a subsequent manual or automatic (trip [PWR] / scram [BWR]) is successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plant's decay heat removal systems.

A manual action at the reactor control consoles is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor (trip [PWR] / scram [BWR])). 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. [BWR] The plant response to the failure of an automatic or manual reactor (trip [PWR] / 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 SA5SA6. 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 reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.

Should a reactor (trip [PWR] / scram [BWR]) signal be generated as a result of plant work (e.g., RPS setpoint testing), the following classification guidance should be applied.

If the signal generated as a result of plant work causes a plant transient that results in a condition that should have included an automatic reactor (trip [PWR] / scram [BWR]) and the RPS fails to automatically shutdown the reactor, then this IC and associated the EALs are applicable, and should be evaluated.

If the signal generated as a result of plant work does not cause a plant transient and the (trip [PWR] / scram [BWR])

failure is determined through other means (e.g., assessment of test results), then this IC and associated the EALs are not applicable and no classification is warranted.

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Reference(s):

1. Unit 1 Technical Specifications Table 3.3.1-1 Reactor Protection System Instrumentation 2. Unit 2 Technical Specifications Table 3.3-1 Reactor Protective Instrumentation
3. Unit 1 and Unit 2 Technical Specifications Table 1.1-1 Modes
4. OP-1202.001 Reactor Trip
5. OP-2202.001 Standard Post Trip Actions 6. NEI 99-01 SU5

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Category: S - System Malfunction Subcategory: 6 - RPS Failure

Initiating Condition: Automatic or manual trip fails to shut down the reactor and subsequent manual actions taken at the reactor control consoles are

not successful in shutting down the reactor EAL: SA6.1 Alert An automatic or manual trip fails to shut down the reactor as indicated by reactor power > 5% AND Manual trip actions taken at the reactor contro l console (C03 [2C03/2C14]) (manual reactor trip pushbuttons or DROPS[DSS]) are not successful in shutting down the reactor as indicated by reactor power > 5% (Note 8)

Note 8: A manual action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron injection strategies.

Mode Applicability:

1 - Power Operation

Definition(s):

None Basis:

This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor (trip [PWR] / scram [

BWR]) 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 (trip [PWR] / scram [BWR])). This action does not include manually driving in control rods or implementation of boron injection strategies. If this action(s) is unsuccessful, operators would immediately pursue additional manual actions at locations away from the reactor control

console s (e.g., locally opening breakers). Actions taken at back panels or other locations within to 0CAN091801

Page 245 of 257 Attachment 1 - Emergency Action Level Technical Bases

the Control Room, or any location outside the Control Room, are not considered to be "at the reactor control console s." Taking the Reactor Mode Switch to SHUTDOWN is considered to be a manual scram action. [BWR] The plant response to the failure of an automatic or manual reactor (trip [PWR] / 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 the failure to shut down the reactor is prolonged enough to cause a challenge to the core cooling

[PWR] / RPV water level [BWR]

or RCS RCS heat removal safety functions, the emergency classification level will escalate to a Site Area Emergency via IC SS 65. Depending upon plant responses and symptoms, escalation is also possible via IC FS1. Absent the plant conditions needed to meet either IC SS 65 or FS1, an Alert declaration is appropriate for this event.

It is recognized that plant responses or symptoms may also require an Alert declaration in accordance with the Recognition Category F ICs; however, this IC and EAL are included to ensure a timely emergency declaration.

A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.

Reference(s):

1. Unit 1 Technical Specifications Table 3.3.1-1 Reactor Protection System Instrumentation 2. Unit 2 Technical Specifications Table 3.3-1 Reactor Protective Instrumentation
3. Unit 1 and Unit 2 Technical Specifications Table 1.1-1 Modes
4. OP-1202.001 Reactor Trip 5. OP-2202.001 Standard Post Trip Actions 6. NEI 99-01 SA5

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Category: S - System Malfunction Subcategory: 6 - RPS Failure

Initiating Condition: Inability to shut down the reactor causing a challenge to core cooling or RCS heat removal

EAL: SS6.1 Site Area Emergency An automatic or manual trip fails to shut down the reactor as indicated by reactor power > 5% AND All actions to shut down the reactor are not successful as indicated by reactor power > 5% AND EITHER

CETs >1200°F RCS heat removal cannot be established using steam generators and an on-shift SRO has determined that the procedure conditions are met to commence initiation of HPI[Once Through] cooling.

Mode Applicability:

1 - Power Operation Definition(s):

None

Basis: This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor

(trip [PWR] / scram [

BWR]) that results in a reactor shutdown, all subsequent operator actions to manually shutdown the reactor are unsuccessful, and continued power generation is challenging the capability to adequately remove heat from the core and/or the RCS. This condition will lead to fuel damage if additional mitigation actions are unsuccessful and thus warrants the declaration of a Site Area Emergency.

In some instances, the emergency classification resulting from this IC/EAL may be higher than that resulting from an assessment of the plant responses and symptoms against the Recognition Category F ICs/EALs. This is appropriate in that the Recognition Category F ICs/EALs do not address the additional threat posed by a failure to shut down the reactor. The inclusion of this IC and EAL ensures the timely declaration of a Site Area Emergency in response to prolonged failure to shutdown the reactor.

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Page 247 of 257 Attachment 1 - Emergency Action Level Technical Bases

A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.

Escalation of the emergency classification level would be via IC AG1 or FG1.

Reference(s):

1. Unit 1 Technical Specifications Table 3.3.1-1 Reactor Protection System Instrumentation 2. Unit 2 Technical Specifications Table 3.3-1 Reactor Protective Instrumentation
3. Unit 1 and Unit 2 Technical Specifications Table 1.1-1 Modes
4. OP-1202.001 Reactor Trip
5. OP-2202.001 Standard Post Trip Actions 6. OP-1202.004 Overheating 7. OP-2202.006 Loss of Feedwater
8. OP-1202.013 Figure 1, Saturation and Adequate SCM
9. Calculation 90-E-0116-07 Unit 1 EOP Setpoint Document, Setpoint B.19
10. OP-2202.009 Functional Recovery
11. Calculation 90-E-0116-01 Unit 2 EOP Setpoint Document 12. NEI 99-01 SS5

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Category: S - System Malfunction Subcategory: 7 - Loss of Communications

Initiating Condition: Loss of all onsite or offsite communications capabilities EAL: SU7.1 Unusual Event Loss of all Table 1[2]S-4 onsite communication methods OR Loss of all Table 1[2]S-4 State and local agency communication methods OR Loss of all Table 1[2]S-4 NRC communication methods

Table 1[2]S-4 Communication Methods System Onsite State / Local NRC Station radio system X ANO plant phone system X Gaitronics X Telephone Systems: Commercial Microwave Satellite VOIP X X Emergency Notification System (ENS) X Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown

Definition(s):

None to 0CAN091801

Page 249 of 257 Attachment 1 - Emergency Action Level Technical Bases

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 State and local agencies 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 #1The 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 State and local agencies of an emergency declaration. The OROs State and local agencies referred to here are the Arkansas Department of Health, Arkansas Department of Emergency Management, Pope, Yell, Johnson, and Logan County agencies.(see Developer Notes) EAL #3The third EAL addresses a total loss of the communications methods used to notify the NRC of an emergency declaration.

This EAL is the hot condition equivalent of the cold condition EAL CU5.1.

Reference(s):

1. OP-1903.062 Communications System Operating Procedure 2. NEI 99-01 SU6 to 0CAN091801

Page 250 of 257 Attachment 1 - Emergency Action Level Technical Bases

Category: S - System Malfunction Subcategory: 8 - Containment Failure

Initiating Condition: Failure to isolate containment or loss of containment pressure control EAL: SU8.1 Unusual Event Any penetration is not closed within 15 min. of an ESAS [CIAS] actuation signal OR Containment pressure > 44.7 psia [23.3 psia] with < one full train of containment heat removal systems (Note 9) operating per design for 15 min. (Note 1)

Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Note 9: One full train of containment heat removal systems consists of one train of RB [Containment] Spray and one train of RB [Containment] Cooling System.

Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown

Definition(s):

None Basis:

A penetration is closed for this EAL if either side of the penetration has a closed valve or a check valve is intact (for penetrations that only have one automatic valve and a check valve).

This IC EAL addresses a failure of one or more containment penetrations to automatically isolate (close) when required by an actuation signal. It also addresses an event that results in high containment pressure with a concurrent failure of containment pressure control systems. Absent challenges to another fission product barrier, either condition represents potential degradation of the level of safety of the plant.

For EAL #1the first condition, the containment isolation signal must be generated as the result on an off-normal/accident condition (e.g., a safety injection or high containment pressure); a failure resulting from testing or maintenance does not warrant classification. The determination of containment and penetration status - isolated or not isolated - should be made in accordance with the appropriate criteria contained in the plant AOPs and EOPs. The 15-minute to 0CAN091801

Page 251 of 257 Attachment 1 - Emergency Action Level Technical Bases

criterion is included to allow operators time to manually isolate the required penetrations, if possible.

EAL #2The second condition addresses a condition where containment pressure is greater than the setpoint at which containment energy (heat) removal systems are designed to automatically actuate, and less than one full train of equipment is capable of operating per design. The 15-minute criterion is included to allow operators time to manually start equipment that may not have automatically started, if possible. The inability to start the required equipment indicates that containment heat removal/depressurization systems (e.g., containment sprays or ice condenser fans) are either lost or performing in a degraded manner.

This event would escalate to a Site Area Emergency in accordance with IC FS1 if there were a concurrent loss or potential loss of either the Fuel Clad or RCS fission product barriers.

Reference(s):

1. OP-1202.010 ESAS 2. 1SAR 6.2 Reactor Building Spray System 3. 1SAR 6.3 Reactor Building Cooling System 4. OP-2202.003 Loss of Coolant Accident
5. OP-2202.010 Standard Attachments, Attachment 22
6. 2SAR 6.2.2 Containment Heat Removal Systems
7. 2SAR 7.3.1.1.11.2 Containment Spray System 8. NEI 99-01 SU7

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Page 252 of 257 Attachment 1 - Emergency Action Level Technical Bases

Category: S - System Malfunction Subcategory: 9 - Hazardous Event Affecting Safety Systems

Initiating Condition: Hazardous event affecting SAFETY SYSTEMS needed for the current operating mode

EAL: SA9.1 Alert The occurrence of any Table 1[2]S-5 hazardous event AND Event damage has caused indications of degraded performance on one train of a SAFETY SYSTEM needed for the current operating mode AND EITHER:

Event damage has caused indications of degraded performance to the second train of the SAFETY SYSTEM needed for the current operating mode Event damage has resulted in VISIBLE DAMAGE to the second train of the SAFETY SYSTEM needed for the current operating mode (Notes 10, 11)

Note 10: If the affected SAFETY SYSTEM train was already inoperable or out of service before the hazardous event occurred, then emergency classification is not warranted. Note 11: If the hazardous event only resulted in VISIBLE DAMAGE, with no indications of degraded performance to at least one train of a SAFETY SYSTEM, then this emergency classification is not warranted.

Table 1[2]S-5 Hazardous Events Seismic event (earthquake) Internal or external FLOODING event High winds or tornado strike FIRE EXPLOSION Other events with similar hazard characteristics as determined by the Shift Manager to 0CAN091801

Page 253 of 257 Attachment 1 - Emergency Action Level Technical Bases

Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):

EXPLOSION - A rapid, violent and catastrophic failure of a piece of equipment due to combustion, chemical reaction or overpressurization. A release of steam (from high energy lines or components) or an electrical component failure (caused by short circuits, grounding, arcing, etc.) should not automatically be considered an explosion. Such events require a post-event inspection to determine if the attributes of an explosion are present.

FIRE - Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred but is not required if large quantities of smoke and heat are observed.

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

SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10 CFR 50.2).

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

(1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.

VISIBLE DAMAGE - Damage to a SAFETY SYSTEM train that is readily observable without measurements, testing, or analysis. The visual impact of the damage is sufficient to cause concern regarding the operability or reliability of the affected SAFETY SYSTEM train.

Basis: This IC addresses a hazardous event that causes damage to SAFETY SYSTEMS needed for the current operating mode. In order to provide the appropriate context for consideration of an ALERT classification, the hazardous event must have caused indications of degraded SAFETY SYSTEM performance in one train, and there must be either indications of performance issues with the second SAFETY SYSTEM train or VISIBLE DAMAGE to the second train such that the potential exists for this second SAFETY SYSTEM train to have performance issues. In other words, in order for this EAL to be classified, the hazardous event must occur, at least one SAFETY SYSTEM train must have indications of degraded performance, and the second SAFETY SYSTEM train must have indications of degraded performance or VISIBLE DAMAGE such that the potential exists for performance issues. to 0CAN091801

Page 254 of 257 Attachment 1 - Emergency Action Level Technical Bases

Note that this second SAFETY SYSTEM train is from the same SAFETY SYSTEM that has indications of degraded performance; commercial nuclear power plants are designed to be able to support single system issues without compromising public health and safety from radiological events. Indications of degraded performance addresses damage to a SAFETY SYSTEM train that is in service/operation since indications for it will be readily available. The indications of degraded performance should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train.

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

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.1 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 1.b.2 addresses damage to a SAFETY SYSTEM component that is not in service/operation or readily apparent through indications alone, or to a structure containing SAFETY SYSTEM components. Operators will make this determination based on the totality of available event and damage report information. This is intended to be a brief assessment not requiring lengthy analysis or quantification of the damage.

Escalation of the emergency classification level would be via IC FS1 or AS1.

This EAL is the hot condition equivalent of the cold condition EAL CA6.1.

Reference(s):

1. EP FAQ 2016-002 2. NEI 99-01 SA9 to 0CAN091801

Page 255 of 257 Attachment 2 - Safe Operation & Shutdown Areas Tables 1[2]A-3 & 1[2]H-2 Bases

Background

NEI 99-01 Revision 6 ICs AA3 and HA5 prescribe declaration of an Alert based on impeded access to rooms or areas (due to either area radiation levels or hazardous gas concentrations) where equipment necessary for normal plant operations, cooldown or shutdown is located.

These areas are intended to be plant operating mode dependent. Specifically the Developers Notes for AA3 and HA5 states:

The "site-specific list of plant rooms or areas with entry-related mode applicability identified" should specify those rooms or areas that contain equipment which require a manual/local action as specified in operating procedures used for normal plant operation, cooldown and shutdown. Do not include rooms or areas in which actions of a contingent or emergency nature would be performed (e.g., an action to address an off-normal or emergency condition such as emergency repairs, corrective measures or emergency operations). In addition, the list should specify the plant mode(s) during which entry would be required for each room or

area. The list should not include rooms or areas for which entry is required solely to perform actions of an administrative or record keeping nature (e.g., normal rounds or routine inspections).

Further, as specified in IC HA5:

The list need not include the Control Room if adequate engineered safety/design features are in place to preclude a Control Room evacuation due to the release of a hazardous gas. Such features may include, but are not limited to, capability to draw air from multiple air intakes at different and separate locations, inner and outer atmospheric boundaries, or the capability to acquire and maintain positive pressure within the Control Room envelope.

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Page 256 of 257 Attachment 2 - Safe Operation & Shutdown Areas Tables 1[2]A-3 & 1[2]H-2 Bases

ANO Table 1[2]A-3 and 1[2]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 1 AREA MODES PURPOSE REFERENCE A-4 Switchgear Room 3, 4 Core flood tank valves, decay heat removal (DHR)

OP-1102.010 OP-1104.004 Upper North Electrical Penetration Room 3, 4 DHR alignment OP-1104.004 Lower South Electrical Equipment Room 3, 4 DHR alignment OP-1104.004 Unit 2 AREA MODES PURPOSE REFERENCE Aux Building 317' Emergency Core Cooling Rooms 3, 4 Shutdown Cooling (SDC) venting and alignment OP-2104.004 Aux Building 317' Tendon Gallery Access 3, 4 SDC alignment OP-2104.004 Aux Building 335' Charging Pumps / Motor Control Center (MCC) 2B-52 3, 4 Charging low pressure operation, T-Hot injection valves, and SDC alignment OP-2102.010 OP-2104.004 Auxiliary Building 354' MCC 2B-62 Area 3, 4 SDC alignment and T-Hot injection valves at MCC 2B-62 OP-2102.010 OP-2104.004 Emergency Diesel Generator Corridor 3, 4 Close Safety Injection Tank (SIT) valves and SDC / Low Temperature Overpressure (LTOP) valve alignment at MCC 2B-51 OP-2102.010 Lower South Piping Penetration Room 3, 4 SDC alignment OP-2104.004 Aux Building 386' Containment Hatch 3, 4 Close SIT valves at MCC 2B-61 OP-2102.010 Mode 3 is included above for DHR- and SDC-related activities because the procedures begin alignment in Mode 3; however, these actions co uld be delayed until Mode 4, if necessary. In order to ensure adequate guidance to emergency response personnel, the above areas are added to the EAL in order to provide prompt operator guidance for EAL declaration.

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Page 257 of 257 Attachment 2 - Safe Operation & Shutdown Areas Tables 1[2]A-3 & 1[2]H-2 Bases

Both ANO-1 and ANO-2 Control Room ventilation systems have adequate engineered safety/design features in place to preclude a Control Room evacuation due to the release of a hazardous gas. Therefore the Control Room is not included in this assessment or in Tables 1[2]H-2.

Table 1[2]A-3 & 1[2]H-2 Results Table 1[2]A-3 & 1[2]H-2 Safe Operation & Shutdown Rooms/Areas Unit 1 Room/Area Mode Applicability A-4 Switchgear Room 3, 4 Upper North Electrical Penetration Room 3, 4 Lower South Electrical Equipment Room 3, 4 Unit 2 Room/Area Mode Applicability Aux Building 317' Emergency Core Cooling Rooms 3, 4 Aux Building 317' Tendon Gallery Access 3, 4 Aux Building 335' Charging Pumps / MCC 2B-52 3, 4 Auxiliary Building 354' MCC 2B-62 Area 3, 4 Emergency Diesel Generator Corridor 3, 4 Lower South Piping Penetration Room 3, 4 Aux Building 386' Containment Hatch 3, 4

Enclosure 3 to 0CAN091801 Proposed EAL Technical Basis Document (Clean) to 0CAN091801

Page 1 of 240 Table of Contents Section Page

1.0 INTRODUCTION

.................................................................................................................2

2.0 DISCUSSION

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

......2 2.1 Background ................................................................................................................

2 2.2 Fission Product Barriers .............................................................................................3 2.3 Fission Product Barrier Classification Criteria ............................................................3 2.4 EAL Organization .......................................................................................................4

2.5 Technical

Basis Information .......................................................................................5

2.6 Operations

Mode Applicability

....................................................................................7

3.0 GUIDANCE

ON MAKING EMERGENCY CLASSIFICATIONS ..........................................8 3.1 General Considerations .............................................................................................8

3.2 Classification

Methodology ......................................................................................10

4.0 REFERENCES

..................................................................................................................13 4.1 Developmental .........................................................................................................13 4.2 Implementing............................................................................................................13

5.0 DEFINITIONS, ACRONYMS & ABBREVIATIONS ...........................................................13 5.1 Definitions (ref. 4.1.1 except as noted) ....................................................................13 5.2 Abbreviations/Acronyms ..........................................................................................18 6.0 ANO-TO-NEI 99-01 Rev. 6 EAL CROSS-REFERENCE ..................................................21

7.0 ATTACHMENTS

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

24 7.1 Attachment 1, Emergency Action Level Technical Bases ........................................24 Category A - Abnormal Rad Levels / Rad Effluents ................................................25 Category C - Cold Shutdown / Refueling System Malfunction ................................65 Category E - Independent Spent Fuel Storage Installation (ISFSI) .......................104 Category F - Fission Product Barrier Degradation ................................................107 Table 1[2]F-1, Fission Product Barrier Threshold Matrix & Bases ...114 Category H - Hazards and Other Conditions Affecting Plant Safety .....................158 Category S - System Malfunction ..........................................................................196 7.2 Attachment 2, Safe Operation & Shutdown Areas Tables 1[2]A-3 & 1[2]H-2 Bases .....................................................239 to 0CAN091801

Page 2 of 240

1.0 INTRODUCTION

This document provides an explanation and rationale for each Emergency Action Level (EAL) included in the EAL Upgrade Project for Arkansas Nuclear One (ANO). It should be used to facilitate review of the ANO EALs and provide historical documentation for future reference. Decision-makers responsible for implementation of 1903.010, Emergency Action Level Classification, may use this document as a technical reference in support of EAL interpretation.

This information may assist the Emergency Director in making classifications, particularly those involving judgment or multiple events. The basis information may also be useful in training and for explaining event classifications to off-site officials.

The expectation is that emergency classifications are to be made as soon as conditions are present and recognizable for the classification, but within 15 minutes or less in all cases when conditions are present and have been recognized. Use of this document for assistance is not intended to delay the emergency classification.

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

10 CFR 50.54(q).

2.0 DISCUSSION

2.1 Background

EALs are the plant-specific indications, conditions or instrument readings that are utilized to classify emergency conditions defined in the ANO Emergency Plan.

In 1992, the NRC endorsed NUMARC/NESP-007 "Methodology for Development of Emergency Action Levels" as an alternative to NUREG-0654 EAL guidance.

NEI 99-01 (NUMARC/NESP-007) Revisions 4 and 5 were subsequently issued for industry implementation. Enhancements over earlier revisions included:

Consolidating the system malfunction initiating conditions and example emergency action levels which address conditions that may be postulated to occur during plant shutdown conditions. Initiating conditions and example emergency action levels that fully address conditions that may be postulated to occur at permanently Defueled Stations and Independent Spent Fuel Storage Installations (ISFSIs). Simplifying the fission product barrier EAL threshold for a Site Area Emergency.

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 t he Development of Emergency Action Levels for Non-Passive Reactors," November 2012 (ref. 4.1.1), ANO conducted an EAL implementation upgrade project that produced the EALs discussed herein.

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Page 3 of 240

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 a greater probability of barrier loss and reduced certainty of maintaining the barrier.

The primary fission product barriers are:

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

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

C. Containment Barrier (CNB): The Containment Barrier includes the Reactor Building and connections up to and including the outermost containment isolation valves. This barrier also includes the main steam, feedwater, and blowdown line extensions outside the Reactor Building up to and including the outermost secondary side isolation valve.

Containment Barrier thresholds are used as criteria for escalation of the Emergency Classification Level (ECL) from an 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

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Page 4 of 240 2.4 EAL Organization

The ANO EAL scheme includes the following features:

Division of the EAL set into three broad groups:

o EALs applicable under any plant operating modes - This group would be reviewed by the EAL-user any time emergency classification is considered.

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

o EALs applicable only under cold operating modes - This group would only be reviewed by the EAL-user when the plant is in Cold Shutdown, Refueling or Defueled mode. The purpose of the groups is to avoid review of hot condition EALs when the plant is in a cold condition and avoid review of cold condition EALs when the plant is in a hot condition. This approach significantly minimizes the total number of EALs that must be reviewed by the EAL-user for a given plant condition, reduces EAL-user reading burden and, thereby, speeds identification of the EAL that applies to the emergency.

Within each group, assignment of EALs to categories and subcategories:

Category and subcategory titles are selected to represent conditions that are operationally significant to the EAL-user. The ANO EAL categories are aligned to and represent the NEI 99-01, "Recognition Categories." Subcategories are used in the ANO scheme as necessary to further divide the EALs of a category into logical sets of possible emergency classification thresholds. The ANO EAL categories and subcategories are listed below.

The primary tool for determining the emergency classification level is the EAL Classification Matrix. The user of the EAL Classification Matrix may (but is not required to) consult the EAL technical bases in order to obtain additional information concerning the EALs under classification consideration. The user should consult Section 3.0 and Attachment 1 of this document for such information.

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Page 5 of 240 EAL Groups, Categories and Subcategories EAL Group/Category EAL Subcategory Any Operating Mode:

A - Abnormal Rad Levels / Rad Effluent 1 - Radiological Effluent 2 - Irradiated Fuel Event 3 - Area Radiation Levels H - Hazards and Other Conditions Affecting Plant Safety 1 - Security 2 - Seismic Event 3 - Natural or Technological Hazard 4 - Fire 5 - Hazardous Gas 6 - Control Room Evacuation 7 - Emergency Director Judgment E - Independent Spent Fuel Storage Installation (ISFSI) 1 - Confinement Boundary Hot Conditions:

S - System Malfunction 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 - Containment Failure 9 - Hazardous Event Affecting Safety Systems F - Fission Product Barrier Degradation None Cold Conditions:

C - Cold Shutdown / Refueling System Malfunction 1 - RCS Level 2 - Loss of Essential AC Power 3 - RCS Temperature 4 - Loss of Vital DC Power 5 - Loss of Communications 6 - Hazardous Event Affecting Safety Systems

2.5 Technical

Bases Information

EAL technical bases are provided in Attachment 1 for each EAL according to EAL group (Any, Hot, Cold), EAL category (A, C, E, F, H and S) 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:

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Page 6 of 240 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 (A, C, E, F, H or S) 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. If an ANO Unit 2 EAL threshold value differs from Unit 1, the Unit 2 threshold is enclosed in brackets. For example, in the EAL threshold "RVLMS Levels 1 through 8 indicate DRY [RVLMS Levels 1 through 5 indicate DRY]", "RVLMS Levels 1 through 5 indicate DRY" apply only to Unit 2.

Mode Applicability One or more of the following plant operating conditions comprise the mode to which each EAL is applicable: 1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown, 5 - Cold Shutdown, 6 - Refueling, DEF - Defueled, or All. (See Section 2.6 for operating mode definitions).

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Page 7 of 240 Definitions:

If the EAL wording contains a defined term, the definition of the term is included in this section. These definitions can also be found in Section 5.1.

Basis: An EAL basis section that provides ANO-relevant information concerning the EAL as well as a description of the rationale for the EAL as provided in NEI 99-01 Rev. 6.

Reference(s):

Source documentation from which the EAL is derived.

2.6 Operating

Mode Applicability

Unit 1 (ref. 4.1.6):

1 Power Operation K eff 0.99, reactor power > 5%

2 Startup K eff 0.99, reactor power 5% 3 Hot Standby K eff < 0.99, reactor coolant temperature 280°F 4 Hot Shutdown K eff < 0.99, reactor coolant temperature 280°F > Tavg > 200°F and all reactor vessel head closure bolts fully tensioned 5 Cold Shutdown K eff < 0.99, reactor coolant temperature 200°F and all reactor vessel head closure bolts fully tensioned 6 Refueling One or more reactor vessel head closure bolts less than fully tensioned DEF Defueled All fuel assemblies have been removed from Containment and placed in the spent fuel pool.

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Page 8 of 240 Unit 2 (ref. 4.1.6):

1 Power Operation K eff 0.99, reactor power > 5%, average coolant temperature 300°F 2 Startup K eff 0.99, reactor power 5%, average coolant temperature 300°F 3 Hot Standby K eff < 0.99, average coolant temperature 300°F 4 Hot Shutdown K eff < 0.99, average coolant temperature 300°F > Tavg > 200°F 5 Cold Shutdown K eff < 0.99, average coolant temperature 200°F 6 Refueling K eff 0.95, average coolant temperature 140°F, reactor vessel head unbolted or removed, and fuel in the vessel.

DEF Defueled All fuel assemblies have been removed from Containment and placed in the spent fuel pool.

The plant operating mode that exists at the time t hat 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.

3.0 GUIDANCE

ON MAKING EMERGENCY CLASSIFICATIONS

3.1 General

Considerations

When making an emergency classification, the Emergency Director 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.

EAL matrices should be read from left to right, from General Emergency to Unusual Event, and top to bottom. Declaration decisions should be independently verified before declaration is made except when gaining this verification would exceed the 15 minute declaration requirement. Place keeping should be used on all EAL matrices.

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Page 9 of 240

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 provi ded guidance on implementing this requirement in NSIR/DPR-ISG-01, "Interim Staff Guidance, Emergency Planning for Nuclear Power Plants" (ref. 4.1.8).

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 the indicator's operability, the condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment.

3.1.3 Imminent

Conditions

For ICs and EALs that have a stipulated time duration (e.g., 15 minutes, 30 minutes, etc.), the Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the condition has exceeded, or will likely exceed, the applicable time. If an ongoing radiological release is detected and the release start time is unknown, it should be assumed that the release duration specified in the IC/EAL has been exceeded, absent data to the contrary.

3.1.4 Planned

vs. Unplanned Events

A planned work activity that results in an expected event or condition which meets or exceeds an EAL does not warrant an emergency declaration provided that: 1) the activity proceeds as planned, and 2) the plant remains within the limits imposed by the operating license. Such activities include planned work to test, manipulate, repair, maintain or modify a system or component. In these cases, the controls associated with the planning, preparation and execution of the work will ensure that compliance is maintained with all aspects of the operating license provided that the activity proceeds and concludes as expected. Events or conditions of this type may be subject to the reporting requirements of 10 CFR 50.72 (ref. 4.1.4).

3.1.5 Classification

Based on Analysis

The assessment of some EALs is based on the results of analyses that are necessary to ascertain whether a specific EAL threshold has been exceeded (e.g., dose assessments, chemistry sampling, RCS leak rate calculation, etc.). For these EALs, the EAL wording or the to 0CAN091801

Page 10 of 240 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 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 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 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. 4.1.8).

3.2.1 Classification

of Multiple Events and Conditions

When multiple emergency events or conditions are present, the user will identify all met or exceeded EALs. The highest applicable ECL identified during this review is declared. For example: If an Alert EAL and a Site Area Emergency EAL are met, whether at one unit or at two units, a Site Area Emergency should be declared.

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

If two Alert EALs are met, whether at one unit or at two units, an Alert should be declared.

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 to 0CAN091801

Page 11 of 240 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 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, 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 Termination

An ECL may be terminated when the event or condition that meets the classified IC and EAL no longer exists, and other site-specific termination requirements are met.

3.2.5 Classification

of Short-Lived Events Event-based ICs and EALs define a variety of specific occurrences that have potential or actual safety significance. By their nature, some of these events may be short-lived and, thus, over before the emergency classification assessment can be completed. If an event occurs that meets or exceeds an EAL, the associated ECL must be declared regardless of its continued presence at the time of declaration. Examples of such events include an earthquake or a failure of the reactor protection system to automatically trip the reactor followed by a successful manual trip.

3.2.6 Classification

of Transient Conditions

Many of the ICs and/or EALs employ time-based criteria. These criteria will require that the IC/EAL conditions be present for a defined period of time before an emergency declaration is warranted. In cases where no time-based criterion is specified, it is recognized that some transient conditions may cause an EAL to be met for a brief period of time (e.g., a few seconds to a few minutes). The following guidance should be applied to the classification of these conditions.

EAL momentarily met during expected plant response - In instances in which 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.

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Page 12 of 240 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. 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 completing the review and steps necessary to make the emergency declaration. This provision is included to ensure that any public protective actions resulting from the emergency classification are truly warranted by the plant conditions.

3.2.7 After-the-Fact Discovery of an Emergency Event or Condition

In some cases, an EAL may be met but the emergency classification was not made at the time of the event or condition. This situation can occur when personnel discover that an event or condition existed which met an EAL, but no emergency was declared, and the event or condition no longer exists at the time of discovery. This may be due to the event or condition not being recognized at the time or an error that was made in the emergency classification process.

In these cases, no emergency declaration is warranted; however, the guidance contained in NUREG-1022 (ref. 4.1.3) is applicable. Specifically, the event should be reported to the NRC in accordance with 10 CFR § 50.72 (ref. 4.1.4) within one hour of the discovery of the undeclared event or condition. The licensee should also notify appropriate State and local agencies in accordance with the agreed upon arrangements.

3.2.8 Retraction

of an Emergency Declaration

Guidance on the retraction of an emergency declaration reported to the NRC is discussed in NUREG-1022 (ref. 4.1.3).

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Page 13 of 240

4.0 REFERENCES

4.1 Developmental

4.1.1 NEI 99-01 Revision 6, Methodology for the Development of Emergency Action Levels for Non-Passive Reactors, ADAMS Accession Number ML12326A805 4.1.2 RIS 2007-02 Clarification of NRC Guidance for Emergency Notifications During Quickly Changing Events, February 2, 2007. 4.1.3 NUREG-1022 Event Reporting Guidelines: 10CFR50.72 and 50.73 4.1.4 10 § CFR 50.72 Immediate Notification Requirements for Operating Nuclear Power Reactors 4.1.5 10 § CFR 50.73 License Event Report System 4.1.6 Unit 1[2] Technical Specifications Table 1.1-1[1.1], Modes[Operational Modes]

4.1.7 Arkansas

Nuclear One Offsite Dose Calculation Manual (ODCM) 4.1.8 NSIR/DPR-ISG-01 Interim Staff Guidance, Emergency Planning for Nuclear Power Plants 4.1.9 Arkansas Nuclear One Emergency Plan 4.1.10 1015.008 Unit 2 SDC Control

4.2 Implementing

4.2.1 1903.010 Emergency Action Level Classification 4.2.2 NEI 99-01 Rev. 6 to ANO EAL Comparison Matrix 4.2.3 ANO EAL Matrix 5.0 DEFINITIONS, ACRONYMS & ABBREVIATIONS

5.1 Definitions

(ref. 4.1.1 except as noted)

Selected terms used in Initiating Condition, Emergency Action Level statements and EAL bases are set in all capital letters (e.g., ALL CAPS). These words are defined terms that have specific meanings as used in this document. The definitions of these terms are provided below.

Alert Events are in progress, or have occurred, which involve an actual or potential substantial degradation of the level of safety of the plant or a security event that involves probable life threatening risk to site personnel or damage to site equipment because of HOSTILE ACTION.

Any releases are expected to be limited to small fractions of the EPA Protective Action Guideline exposure levels.

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Page 14 of 240 Confinement Boundary The barrier(s) between spent fuel and the environment once the spent fuel is processed for dry storage. As related to the ANO ISFSI, the Confinement Boundary is comprised of either the Multi-assembly Sealed Basket (MSB) (SNC System) or Multi-Purpose Canister (MPC) (Holtec System).

Containment Closure

The action to secure primary containment and its associated structures, systems, and components as a functional barrier to fission product release under existing plant conditions (ref. 4.1.10).

As applied to ANO, Containment Closure must be capable of being set within 30 minutes. Containment Closure is set when the penetrations are isolated by manual or automatic isolation valve, blind flange, or equivalent.

Emergency Action Level (EAL)

A pre-determined, site-specific, observable threshold for an INITIATING CONDITION that, when met or exceeded, places the plant in a given emergency classification level.

Emergency Classification Level (ECL)

One of a set of names or titles established by the US Nuclear Regulatory Commission (NRC) for grouping off-normal events or conditions according to (1) potential or actual effects or consequences, and (2) resulting onsite and offsite response actions. The emergency classification levels, in ascending order of severity, are:

Unusual Event (UE)

Alert Site Area Emergency (SAE)

General Emergency (GE)

Explosion

A rapid, violent and catastrophic failure of a piece of equipment due to combustion, chemical reaction or overpressurization. A release of steam (from high energy lines or components) or

an electrical component failure (caused by short circuits, grounding, arcing, etc.) should not automatically be considered an explosion. Such events require a post-event inspection to determine if the attributes of an explosion are present.

Faulted The term applied to a steam generator that has a steam leak on the secondary side of sufficient size to cause an uncontrolled drop in steam generator pressure or the steam generator to become completely depressurized.

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Page 15 of 240 Fire Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred but is not required if large quantities of smoke and heat are observed.

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

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

General Emergency Events are in 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 PAG exposure levels offsite for more than the immediate site area.

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

Hostile Action An act toward ANO 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 ANO. Non-terro rism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the SECURITY OWNER CONTROLLED AREA).

Hostile Force One or more individuals who are engaged in a determined assault, overtly or by stealth and deception, equipped with suitable weapons capable of killing, maiming, or causing destruction.

Imminent The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions.

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Page 16 of 240 Impede(d)

Personnel access to a room or area is hindered to an extent that extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., requiring use of protective equipment, such as SCBAs, that is not routinely employed).

Independent Spent Fuel Storage Installation (ISFSI)

A complex that is designed and constructed for the interim storage of spent nuclear fuel and other radioactive materials associated with spent fuel storage.

Initiating Condition (IC)

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

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

Protected Area An area clearly demarcated by a fence or building wall with an entrance portal that is regulated by Security Personnel to control access (ref. 4.1.9).

RCS Intact The RCS should be considered intact when the RCS pressure boundary is in its normal condition for the cold shutdown mode of operation (e.g., no freeze seals or nozzle dams).

Refueling Pathway All the cavities, tubes, canals and pools through which irradiated fuel may be moved, but not including the reactor vessel.

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

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 10 CFR 50.2):

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Page 17 of 240 Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:

(1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.

Security Condition Any security event as listed in the approved security contingency plan that constitutes a threat/compromise to site security, threat/risk to site personnel, or a potential degradation to the level of safety of the plant. A Security Condition does not involve a HOSTILE ACTION.

Security Owner Controlled Area (SOCA)

The SOCA is the area demarcated as a Vehicle Barrier System (VBS) consisting of passive elements including a series of large concrete blocks on the inside of a delay fence with early warning capabilities. The SOCA is the area between the SOCA Fence and the PROTECTED AREA Boundary (ref. 4.1.9).

Site Area Emergency Events are in progress or have occurred which involve actual or likely major failures of plant functions needed for protection of the public or HOSTILE ACTION that results in intentional damage or malicious acts; (1) toward site personnel or equipment that could lead to the likely failure of or; (2) that prevent effective access to equipment needed for the protection of the

public. Any releases are not expected to result in exposure levels which exceed EPA PAG exposure levels beyond the SITE BOUNDARY.

Site Boundary That boundary defined by a 1046 meter (0.65 mile) radius around the plant (ref. 4.1.7).

Unisolable

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

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

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Page 18 of 240 Unusual Event Events are in progress or have occurred which indicate a potential degradation in the level of safety of the plant or indicate a security threat to facility protection has been initiated. No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of SAFETY SYSTEMS occurs.

Valid An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment.

Visible Damage Damage to a SAFETY SYSTEM train that is readily observable without measurements, testing, or analysis. The visual impact of the damage is sufficient to cause concern regarding the operability or reliability of the affected SAFETY SYSTEM train.

5.2 Abbreviations/Acronyms

°F .................................................................................................................... Degrees Fahrenheit

° ...............................................................................................................................

.......... Degrees AC ..................................................................................................................... Altern ating Current ANO ............................................................................................................ Arkansas Nucle ar One AOP .............................................................................................. Abnormal Operating Procedure ATWS .................................................................................... Anticipated Transient Without Scram BMS ................................................................................................... Boron Management System BWST ................................................................................................ Borated Water Storage Tank CDE ................................................................................................... Committed Dose Equivale nt CET .......................................................................................................... Core Exit Thermo couple CFR ................................................................................................... Code of Federal Regulations CIAS .................................................................................. Containment Isolation Actuation Signal CMT, CNTMT, CTMT .................................................................................................. Containment CNB ................................................................................................................ Containmen t Barrier DBA ............................................................................................................. Design Basis Accident DBE ......................................................................................................... Design Basis Earthqua ke DC ............................................................................................................................. Direct Current DEF ...........................................................................................................................

........ Defueled to 0CAN091801

Page 19 of 240 D/G ....................................................................................................................... Diesel Generator DHR .............................................................................................................. Decay Heat Removal DROPS ............................................................ Diverse Reactor Overpressure Protection System DSC ............................................................................................................. Dry Shielded Canister DSS ............................................................................................................ Diverse Scram Syst em EAL .......................................................................................................... Emergency Action Level ECCS ......................................................................................... Emergency Core Cooling System ECL ............................................................................................... Emergency Classification Le vel DEF ...........................................................................................................................

........ Defueled ENS ............................................................................................... Emergency Notification System EOF ................................................................................................ Emergency Operations Facility EOP ........................................................................................... Emergency Operating Procedure EPA ............................................................................................ Environmental Protection Agenc y ERG ............................................................................................ Emergency Response Guideline EPIP ............................................................................. Emergency Plan Implementing Procedure ESAS ........................................................................... Engineered Safeguards Actuation System ESF ...................................................................................................... Engineered Safety Fe ature ESFAS .................................................................. Engineered Safety Features Actuation System FAA ............................................................................................... Federal Aviation Administration FBI ................................................................................................ Federal Bureau of Investigation FCB ...................................................................................................................... Fuel Clad Barrier FEMA ............................................................................ Federal Emergency Management Agency GE ................................................................................................................... General Emergency HPI ............................................................................................................. High Pressure Injection IC ...................................................................................................................... Initiating Condition IPEEE ............................. Individual Plant Examination of External Events (Generic Letter 88-20) ISFSI ......................................................................... Independent Spent Fuel Storage Installation

K eff ...................................................................................... Effective Neutron Multiplication Factor LCO ................................................................................................

Limiting Condition of Operation LER .............................................................................................................

Licensee Event Report LOCA ...................................................................................................... Loss of Coolant Accident LRW .................................................................................................................... Liquid Rad Waste LTOP ............................................................................................ Low Temperature Overpressure LWR ................................................................................................................ Light Wate r Reactor to 0CAN091801

Page 20 of 240 MCC .............................................................................................................. Motor Contro l Center MPC ................................................ Maximum Permissible Concentration/Multi-Purpose Canister mR, mRem, mrem, mREM ............................................................ milli-Roentgen Equivalent Man MSL ...................................................................................................................... Main Ste am Line MTS ................................................................................................................ Margin to Saturation MW .................................................................................................................................. Megawatt NDTT ...................................................................................... Nil Ductility Transition Temperatur e NEI ............................................................................................................

Nuclear Energy Institute NEIC ................................................................................ National Earthquake Information Center NESP ................................................................................ National Environmental Studies Project NORAD ................................................................ North American Aerospace Defense Command NOT .............................................................................................. Normal Operating Temperature (NO)UE ............................................................................................. Notification of Unusual Event NPP ................................................................................................................ Nuclear Power Plant NRC ............................................................................................ Nuclear Regulatory Commission NSSS ............................................................................................. Nuclear Steam Supply System OBE ................................................................................................... Operating Basis Earthqu ake ODCM ........................................................................................ Off-site Dose Calculation Manual ORO ............................................................................................... Offsite Response Organization PA ...........................................................................................................................

Protected Area PAG ..................................................................................................... Protective Action Gui deline PRA/PSA .................................. Probabilistic Risk Assessment / Probabilistic Safety Assessment P-T .............................................................................................................. Pressure-Temperature PTS ..................................................................................................... Pressurized Thermal Shock PWR ..................................................................................................... Pressurized Water Reactor PSIG ............................................................................................ Pounds per Square Inch Gauge R .............................................................................................................................

........ Roentgen RB .........................................................................................................................

Reactor Building RCC ......................................................................................................... Reactor Control C onsole RCB ............................................................................................. Reactor Coolant System Barrie r RCP ............................................................................................................ Reactor Coolan t Pump RCS ......................................................................................................... Reactor Coolant System Rem, rem, REM ..................................................................................... Roentgen Equivalent Man Rep CET ....................................................................... Representative Core Exit Thermocouples to 0CAN091801

Page 21 of 240 RETS ...................................................................... Radiological Effluent Technical Specifications RPS ...................................................................................................... Reactor Protection Syste m RV ...........................................................................................................................

Reactor Vessel RVLMS ........................................................................... Reactor Vessel Level Monitoring System RWT ............................................................................................................. Refueling Wat er Tank SAR ............................................................................................................. Safety Analysis Report SBO ...................................................................................................................... Stat ion Blackout SCBA ................................................................................... Self-Contained Breathing Apparatus SDC ................................................................................................................... Shutdown Cooling SOCA ........................................................................................... Security Owner Controlled Area SG ....................................................................................................................... Steam Generator SI ............................................................................................................................ Sa fety Injection SPDS ........................................................................................ Safety Parameter Display System SPING ..................................................................................... Super Particulate Iodine Noble Gas SRO ......................................................................................................... Senior Reactor Op erator TEDE ............................................................................................ Total Effective Dose Equivale nt TOAF ................................................................................................................. Top of A ctive Fuel TSC ........................................................................................................ Technical Support Center USGS .......................................................................................... United States Geological Survey VBS ............................................................................................................ Vehicle Barrie r System 6.0 ANO-TO-NEI 99-01 REV. 6 EAL CROSS-REFERENCE This cross-reference is provided to facilitate association and location of an ANO EAL within the NEI 99-01 IC/EAL identification scheme. Further information regarding the development of the ANO EALs based on the NEI guidance can be found in the EAL Comparison Matrix.

ANO NEI 99-01 Rev. 6 EAL IC Example EAL AU1.1 AU1 1, 2 AU1.2 AU1 3 AU2.1 AU2 1 AA1.1 AA1 1 AA1.2 AA1 2 to 0CAN091801

Page 22 of 240 ANO NEI 99-01 Rev. 6 EAL IC Example EAL AA1.3 AA1 3 AA1.4 AA1 4 AA2.1 AA2 1 AA2.2 AA2 2 AA2.3 AA2 3 AA3.1 AA3 1 AA3.2 AA3 2 AS1.1 AS1 1 AS1.2 AS1 2 AS1.3 AS1 3 AS2.1 AS2 1 AG1.1 AG1 1 AG1.2 AG1 2 AG1.3 AG1 3 AG2.1 AG2 1 CU1.1 CU1 1 CU1.2 CU1 2 CU2.1 CU2 1 CU3.1 CU3 1 CU3.2 CU3 2 CU4.1 CU4 1 CU5.1 CU5 1, 2, 3 CA1.1 CA1 1 CA1.2 CA1 2 CA2.1 CA2 1 CA3.1 CA3 1, 2 CA6.1 CA6 1 CS1.1 CS1 1 CS1.2 CS1 2 CS1.3 CS1 3 CG1.1 CG1 1 to 0CAN091801

Page 23 of 240 ANO NEI 99-01 Rev. 6 EAL IC Example EAL CG1.2 CG1 2 EU1.1 EU1 1 FA1.1 FA1 1 FS1.1 FS1 1 FG1.1 FG1 1 HU1.1 HU1 1, 2, 3 HU2.1 HU2 1 HU3.1 HU3 1 HU3.2 HU3 2 HU3.3 HU3 3 HU3.4 HU3 4 HU4.1 HU4 1 HU4.2 HU4 2 HU4.3 HU4 3 HU4.4 HU4 4 HU7.1 HU7 1 HA1.1 HA1 1, 2 HA5.1 HA5 1 HA6.1 HA6 1 HA7.1 HA7 1 HS1.1 HS1 1 HS6.1 HS6 1 HS7.1 HS7 1 HG7.1 HG7 1 SU1.1 SU1 1 SU3.1 SU2 1 SU4.1 SU3 1 SU4.2 SU3 2 SU5.1 SU4 1, 2, 3 SU6.1 SU5 1 SU6.2 SU5 2 to 0CAN091801

Page 24 of 240 ANO NEI 99-01 Rev. 6 EAL IC Example EAL SU7.1 SU6 1, 2, 3 SU8.1 SU7 1, 2 SA1.1 SA1 1 SA3.1 SA2 1 SA6.1 SA5 1 SA9.1 SA9 1 SS1.1 SS1 1 SS2.1 SS8 1 SS6.1 SS5 1 SG1.1 SG1 1 SG1.2 SG8 1

7.0 ATTACHMENTS

7.1 Attachment

1, Emergency Action Level Technical Bases

7.2 Attachment

2, Safe Operation & Shutdown Areas Tables 1[2]A-3 & 1[2]H-2 Bases

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Page 25 of 240 Attachment 1 - Emergency Action Level Technical Bases

Category A - Abnormal Rad Levels / Rad Effluent EAL Group: ANY (EALs in this category are applicable to any plant condition, hot or cold.)

Many EALs are based on actual or potential degradation of fission product barriers because of the elevated potential for offsite radioactivity release. Degradation of fission product barriers though is not always apparent via non-radiological symptoms. Therefore, direct indication of elevated radiological effluents or area radiation levels are appropriate symptoms for emergency classification.

At lower levels, abnormal radioactivity releases may be indicative of a failure of containment systems or precursors to more significant releases. At higher release rates, offsite radiological conditions may result which require offsite protective actions. Elevated area radiation levels in plant may also be indicative of the failure of containment systems or preclude access to plant vital equipment necessary to ensure plant safety.

Events of this category pertain to the following subcategories:

1. Radiological Effluent Direct indication of effluent radiation monitoring systems provides a rapid assessment mechanism to determine releases in excess of classifiable limits. Projected offsite doses, actual offsite field measurements or measured release rates via sampling indicate doses or dose rates above classifiable limits.
2. Irradiated Fuel Event Conditions indicative of a loss of adequate shielding or damage to irradiated fuel may preclude access to vital plant areas or result in radiological releases that warrant emergency classification.
3. Area Radiation Levels Sustained general area radiation levels which may preclude access to areas requiring continuous occupancy also warrant emergency classification.

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Page 26 of 240 Attachment 1 - Emergency Action Level Technical Bases

Category: A - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent

Initiating Condition: Release of gaseous or liquid radioactivity greater than 2 times the ODCM limits for 60 minutes or longer

EAL: AU1.1 Unusual Event Reading on any Table 1[2]A-1 effluent radiation monitor > column "UE" for 60 min. (Notes 1, 2, 3)

Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is 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 due to actions to isolate the release path, then the effluent monitor reading is no longer VALID for classification purposes. Table 1A-1 Unit 1 Effluent Monitor Classification Thresholds (2 min. avg reading)

Release Point Monitor GE SAE Alert UE Gaseous Containment Purge RX-9820 (SPING 1) 4.15E+01 Ci/cc 4.15E+00 µCi/cc4.15E-01 µCi/cc 1.21E-03 µCi/cc Radwaste Area RX-9825 (SPING 2) 2.67E+01 Ci/cc 2.67E+00 µCi/cc2.67E-01 µCi/cc 4.94E-04 µCi/cc Fuel Handling Area RX-9830 (SPING 3) 6.20E+02 Ci/cc 6.20E+01 µCi/cc6.20E+00 µCi/cc 5.44E-04 µCi/cc Emergency Penetration Room RX-9835 (SPING 4) 6.55E+02 Ci/cc 6.55E+01 µCi/cc6.55E+00 µCi/cc 1.21E-02 µCi/cc Liquid Liquid Radwaste RE-4642 ----


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Page 27 of 240 Attachment 1 - Emergency Action Level Technical Bases

Table 2A-1 Unit 2 Effluent Monitor Classification Thresholds (2 min. avg reading)

Release Point Monitor GE SAE Alert UE Gaseous Containment Purge 2RX-9820 (SPING 5) 1.88E+01 Ci/cc 1.88E+00 µCi/cc1.88E-01 µCi/cc 5.48E-04 µCi/cc Radwaste Area 2RX-9825 (SPING 6) 2.35E+01 Ci/cc 2.35E+00 µCi/cc2.35E-01 µCi/cc 4.35E-04 µCi/cc Fuel Handling Area 2RX-9830 (SPING 7) 6.86E+02 Ci/cc 6.86E+01 µCi/cc6.86E+00 µCi/cc 6.04E-04 µCi/cc Emergency Penetration Room 2RX-9835 (SPING 8) 5.88E+02 Ci/cc 5.88E+01 µCi/cc5.88E+00 µCi/cc 1.09E-02 µCi/cc Liquid BMS Liquid Discharge 2RE-2330 ----


---- 2.45E+04 cpm Regenerative Waste Discharge 2RE-4423 ----


---- 2.45E+05 cpm Mode Applicability:

All Definition(s):

VALID - An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment.

Basis: This IC addresses a potential reduction 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.

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Page 28 of 240 Attachment 1 - Emergency Action Level Technical Bases

Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.

Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer VALID for classification purposes.

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

This EAL addresses normally occurring continuous radioactivity releases from monitored gaseous or liquid effluent pathways as well as radioactivity releases that cause effluent radiation monitor readings to exceed 2 times the limit established by a radioactivity discharge permit. Such releases are typically associated with planned batch releases from non-continuous release pathways (e.g., radwaste, waste gas).

Escalation of the emergency classification level would be via IC AA1.

Reference(s):

1. OP-1604.051 Eberline Radiation Monitor System 2. Offsite Dose Calculation Manual
3. EP-CALC-ANO-1701 Radiological Effluent EAL Values
4. NEI 99-01 AU1

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Page 29 of 240 Attachment 1 - Emergency Action Level Technical Bases

Category: A - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent

Initiating Condition: Release of gaseous or liquid radioactivity greater than 2 times the ODCM limits for 60 minutes or longer

EAL: AU1.2 Unusual Event Sample analysis for a gaseous or liquid release indicates a concentration or release rate 2 x ODCM limits for 60 min. (Notes 1, 2)

Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded.

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

Mode Applicability:

All

Definition(s):

None Basis: This IC addresses a potential reduction 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.

to 0CAN091801

Page 30 of 240 Attachment 1 - Emergency Action Level Technical Bases

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

This EAL addresses uncontrolled gaseous or liquid releases that are detected by sample analyses or environmental surveys, particularly on unmonitored pathways (e.g., spills of radioactive liquids into storm drains, heat exchanger leakage in river water systems, etc.).

Escalation of the emergency classification level would be via IC AA1.

Reference(s):

1. Offsite Dose Calculation Manual 2. NEI 99-01 AU1

to 0CAN091801

Page 31 of 240 Attachment 1 - Emergency Action Level Technical Bases

Category: A - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent

Initiating Condition: Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid CDE

EAL: AA1.1 Alert Reading on any Table 1[2]A-1 effluent radiation monitor > column "ALERT" for 15 min. (Notes 1, 2, 3, 4)

Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is 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 due to actions to isolate the release path, then the effluent monitor reading is no longer VALID for classification purposes. Note 4 The pre-calculated effluent monitor values presented in EALs AA1.1, AS1.1 and AG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.

Table 1A-1 Unit 1 Effluent Monitor Classification Thresholds (2 min. avg reading)

Release Point Monitor GE SAE Alert UE Gaseous Containment Purge RX-9820 (SPING 1) 4.15E+01 Ci/cc 4.15E+00 µCi/cc4.15E-01 µCi/cc 1.21E-03 µCi/cc Radwaste Area RX-9825 (SPING 2) 2.67E+01 Ci/cc 2.67E+00 µCi/cc2.67E-01 µCi/cc 4.94E-04 µCi/cc Fuel Handling Area RX-9830 (SPING 3) 6.20E+02 Ci/cc 6.20E+01 µCi/cc6.20E+00 µCi/cc 5.44E-04 µCi/cc Emergency Penetration Room RX-9835 (SPING 4) 6.55E+02 Ci/cc 6.55E+01 µCi/cc6.55E+00 µCi/cc 1.21E-02 µCi/cc Liquid Liquid Radwaste RE-4642 ----


---- 2.46E+05 cpm to 0CAN091801

Page 32 of 240 Attachment 1 - Emergency Action Level Technical Bases

Table 2A-1 Unit 2 Effluent Monitor Classification Thresholds (2 min. avg reading)

Release Point Monitor GE SAE Alert UE Gaseous Containment Purge 2RX-9820 (SPING 5) 1.88E+01 Ci/cc 1.88E+00 µCi/cc1.88E-01 µCi/cc 5.48E-04 µCi/cc Radwaste Area 2RX-9825 (SPING 6) 2.35E+01 Ci/cc 2.35E+00 µCi/cc2.35E-01 µCi/cc 4.35E-04 µCi/cc Fuel Handling Area 2RX-9830 (SPING 7) 6.86E+02 Ci/cc 6.86E+01 µCi/cc6.86E+00 µCi/cc 6.04E-04 µCi/cc Emergency Penetration Room 2RX-9835 (SPING 8) 5.88E+02 Ci/cc 5.88E+01 µCi/cc5.88E+00 µCi/cc 1.09E-02 µCi/cc Liquid BMS Liquid Discharge 2RE-2330 ----


---- 2.45E+04 cpm Regenerative Waste Discharge 2RE-4423 ----


---- 2.45E+05 cpm Mode Applicability:

All Definition(s):

VALID - An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment.

Basis: 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. to 0CAN091801

Page 33 of 240 Attachment 1 - Emergency Action Level Technical Bases

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

Escalation of the emergency classification level would be via IC AS1.

Reference(s):

1. OP-1604.051 Eberline Radiation Monitor System 2. EP-CALC-ANO-1701 Radiological Effluent EAL Values
3. NEI 99-01 AA1 to 0CAN091801

Page 34 of 240 Attachment 1 - Emergency Action Level Technical Bases

Category: A - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent

Initiating Condition: Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid CDE

EAL: AA1.2 Alert Dose assessment using actual meteorology indicates doses > 10 mrem TEDE or 50 mrem thyroid CDE at or beyond the SITE BOUNDARY (Note 4)

Note 4: The pre-calculated effluent monitor values presented in EALs AA1.1, AS1.1 and AG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.

Mode Applicability:

All

Definition(s):

SITE BOUNDARY - That boundary defined by a 1046 meter (0.65 mile) radius around the plant.

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.

Escalation of the emergency classification level would be via IC AS1.

Reference(s):

1. OP-1904.002 Offsite Dose Projections 2. NEI 99-01 AA1 to 0CAN091801

Page 35 of 240 Attachment 1 - Emergency Action Level Technical Bases

Category: A - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent

Initiating Condition: Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid CDE

EAL: AA1.3 Alert Analysis of a liquid effluent sample indicates a concentration or release rate that would result in doses > 10 mrem TEDE or 50 mrem thyroid CDE at or beyond the SITE BOUNDARY for 60 min. of exposure (Notes 1, 2)

Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.

Mode Applicability:

All Definition(s):

SITE BOUNDARY - That boundary defined by a 1046 meter (0.65 mile) radius around the plant.

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.

This EAL is assessed per the ODCM (ref. 2).

Escalation of the emergency classification level would be via IC AS1. to 0CAN091801

Page 36 of 240 Attachment 1 - Emergency Action Level Technical Bases

Reference(s):

1. OP-1904.002 Offsite Dose Projections 2. Offsite Dose Calculation Manual
3. NEI 99-01 AA1 to 0CAN091801

Page 37 of 240 Attachment 1 - Emergency Action Level Technical Bases

Category: A - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent

Initiating Condition: Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid CDE

EAL: AA1.4 Alert Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY: Closed window dose rates > 10 mR/hr expected to continue for 60 min. Analyses of field survey samples indicate thyroid CDE > 50 mrem for 60 min. of inhalation (Notes 1, 2)

Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.

Mode Applicability:

All Definition(s):

SITE BOUNDARY - That boundary defined by a 1046 meter (0.65 mile) radius around the plant.

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.

to 0CAN091801

Page 38 of 240 Attachment 1 - Emergency Action Level Technical Bases

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

Escalation of the emergency classification level would be via IC AS1.

Reference(s):

1. OP-1905.002 Offsite Emergency Monitoring 2. NEI 99-01 AA1

to 0CAN091801

Page 39 of 240 Attachment 1 - Emergency Action Level Technical Bases

Category: A - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent

Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem thyroid CDE

EAL: AS1.1 Site Area Emergency Reading on any Table 1[2]A-1 effluent radiation monitor > column "SAE" for 15 min. (Notes 1, 2, 3, 4)

Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is 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 due to actions to isolate the release path, then the effluent monitor reading is no longer VALID for classification purposes. Note 4: The pre-calculated effluent monitor values presented in EALs AA1.1, AS1.1 and AG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.

Table 1A-1 Unit 1 Effluent Monitor Classification Thresholds (2 min. avg reading)

Release Point Monitor GE SAE Alert UE Gaseous Containment Purge RX-9820 (SPING 1) 4.15E+01 Ci/cc 4.15E+00 µCi/cc4.15E-01 µCi/cc 1.21E-03 µCi/cc Radwaste Area RX-9825 (SPING 2) 2.67E+01 Ci/cc 2.67E+00 µCi/cc2.67E-01 µCi/cc 4.94E-04 µCi/cc Fuel Handling Area RX-9830 (SPING 3) 6.20E+02 Ci/cc 6.20E+01 µCi/cc6.20E+00 µCi/cc 5.44E-04 µCi/cc Emergency Penetration Room RX-9835 (SPING 4) 6.55E+02 Ci/cc 6.55E+01 µCi/cc6.55E+00 µCi/cc 1.21E-02 µCi/cc Liquid Liquid Radwaste RE-4642 ----


---- 2.46E+05 cpm to 0CAN091801

Page 40 of 240 Attachment 1 - Emergency Action Level Technical Bases

Table 2A-1 Unit 2 Effluent Monitor Classification Thresholds (2 min. avg reading)

Release Point Monitor GE SAE Alert UE Gaseous Containment Purge 2RX-9820 (SPING 5) 1.88E+01 Ci/cc 1.88E+00 µCi/cc1.88E-01 µCi/cc 5.48E-04 µCi/cc Radwaste Area 2RX-9825 (SPING 6) 2.35E+01 Ci/cc 2.35E+00 µCi/cc2.35E-01 µCi/cc 4.35E-04 µCi/cc Fuel Handling Area 2RX-9830 (SPING 7) 6.86E+02 Ci/cc 6.86E+01 µCi/cc6.86E+00 µCi/cc 6.04E-04 µCi/cc Emergency Penetration Room 2RX-9835 (SPING 8) 5.88E+02 Ci/cc 5.88E+01 µCi/cc5.88E+00 µCi/cc 1.09E-02 µCi/cc Liquid BMS Liquid Discharge 2RE-2330 ----


---- 2.45E+04 cpm Regenerative Waste Discharge 2RE-4423 ----


---- 2.45E+05 cpm Mode Applicability:

All Definition(s):

VALID - An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment.

Basis: 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.

to 0CAN091801

Page 41 of 240 Attachment 1 - Emergency Action Level Technical Bases

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 AG1.

Reference(s):

1. OP-1604.051 Eberline Radiation Monitor System 2. EP-CALC-ANO-1701 Radiological Effluent EAL Values
3. NEI 99-01 AS1

to 0CAN091801

Page 42 of 240 Attachment 1 - Emergency Action Level Technical Bases

Category: A - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent

Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem thyroid CDE

EAL: AS1.2 Site Area Emergency Dose assessment using actual meteorology indicates doses > 100 mrem TEDE or 500 mrem thyroid CDE at or beyond the SITE BOUNDARY (Note 4)

Note 4 The pre-calculated effluent monitor values presented in EALs AA1.1, AS1.1 and AG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.

Mode Applicability:

All

Definition(s):

SITE BOUNDARY - That boundary defined by a 1046 meter (0.65 mile) radius around the plant).

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.

Escalation of the emergency classification level would be via IC AG1.

Reference(s):

1. OP-1904.002 Offsite Dose Projections 2. NEI 99-01 AS1 to 0CAN091801

Page 43 of 240 Attachment 1 - Emergency Action Level Technical Bases

Category: A - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent

Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem thyroid CDE

EAL: AS1.3 Site Area Emergency Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY: Closed window dose rates > 100 mR/hr expected to continue for 60 min. Analyses of field survey samples indicate thyroid CDE > 500 mrem for 60 min. of inhalation (Notes 1, 2)

Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.

Mode Applicability:

All Definition(s):

SITE BOUNDARY - That boundary defined by a 1046 meter (0.65 mile) radius around the plant.

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.

Escalation of the emergency classification level would be via IC AG1. to 0CAN091801

Page 44 of 240 Attachment 1 - Emergency Action Level Technical Bases

Reference(s):

1. OP-1905.002 Offsite Emergency Monitoring
2. NEI 99-01 AS1

to 0CAN091801

Page 45 of 240 Attachment 1 - Emergency Action Level Technical Bases

Category: A - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent

Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid CDE

EAL: AG1.1 General Emergency Reading on any Table 1[2]A-1 effluent radiation monitor > column "GE" for 15 min. (Notes 1, 2, 3, 4)

Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is 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 due to actions to isolate the release path, then the effluent monitor reading is no longer VALID for classification purposes. Note 4: The pre-calculated effluent monitor values presented in EALs AA1.1, AS1.1 and AG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.

Table 1A-1 Unit 1 Effluent Monitor Classification Thresholds (2 min. avg reading)

Release Point Monitor GE SAE Alert UE Gaseous Containment Purge RX-9820 (SPING 1) 4.15E+01 Ci/cc 4.15E+00 µCi/cc4.15E-01 µCi/cc 1.21E-03 µCi/cc Radwaste Area RX-9825 (SPING 2) 2.67E+01 Ci/cc 2.67E+00 µCi/cc2.67E-01 µCi/cc 4.94E-04 µCi/cc Fuel Handling Area RX-9830 (SPING 3) 6.20E+02 Ci/cc 6.20E+01 µCi/cc6.20E+00 µCi/cc 5.44E-04 µCi/cc Emergency Penetration Room RX-9835 (SPING 4) 6.55E+02 Ci/cc 6.55E+01 µCi/cc6.55E+00 µCi/cc 1.21E-02 µCi/cc Liquid Liquid Radwaste RE-4642 ----


---- 2.46E+05 cpm to 0CAN091801

Page 46 of 240 Attachment 1 - Emergency Action Level Technical Bases

Table 2A-1 Unit 2 Effluent Monitor Classification Thresholds (2 min. avg reading)

Release Point Monitor GE SAE Alert UE Gaseous Containment Purge 2RX-9820 (SPING 5) 1.88E+01 Ci/cc 1.88E+00 µCi/cc1.88E-01 µCi/cc 5.48E-04 µCi/cc Radwaste Area 2RX-9825 (SPING 6) 2.35E+01 Ci/cc 2.35E+00 µCi/cc2.35E-01 µCi/cc 4.35E-04 µCi/cc Fuel Handling Area 2RX-9830 (SPING 7) 6.86E+02 Ci/cc 6.86E+01 µCi/cc6.86E+00 µCi/cc 6.04E-04 µCi/cc Emergency Penetration Room 2RX-9835 (SPING 8) 5.88E+02 Ci/cc 5.88E+01 µCi/cc5.88E+00 µCi/cc 1.09E-02 µCi/cc Liquid BMS Liquid Discharge 2RE-2330 ----


---- 2.45E+04 cpm Regenerative Waste Discharge 2RE-4423 ----


---- 2.45E+05 cpm Mode Applicability:

All Definition(s):

VALID - An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment.

Basis: 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.

to 0CAN091801

Page 47 of 240 Attachment 1 - Emergency Action Level Technical Bases

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.

Reference(s):

1. OP-1604.051 Eberline Radiation Monitor System 2. EP-CALC-ANO-1701 Radiological Effluent EAL Values 3. NEI 99-01 AG1

to 0CAN091801

Page 48 of 240 Attachment 1 - Emergency Action Level Technical Bases

Category: A - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent

Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid CDE

EAL: AG1.2 General Emergency Dose assessment using actual meteorology indicates doses > 1,000 mrem TEDE or 5,000 mrem thyroid CDE at or beyond the SITE BOUNDARY (Note 4)

Note 4 The pre-calculated effluent monitor values presented in EALs AA1.1, AS1.1 and AG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.

Mode Applicability:

All

Definition(s):

SITE BOUNDARY - That boundary defined by a 1046 meter (0.65 mile) radius around the plant).

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.

Reference(s):

1. OP-1904.002 Offsite Dose Projections 2. NEI 99-01 AG1

to 0CAN091801

Page 49 of 240 Attachment 1 - Emergency Action Level Technical Bases

Category: A - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent

Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid CDE

EAL: AG1.3 General Emergency Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY: Closed window dose rates > 1,000 mR/hr expected to continue for 60 min. Analyses of field survey samples indicate thyroid CDE > 5,000 mrem for 60 min. of inhalation (Notes 1, 2)

Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.

Mode Applicability:

All Definition(s):

SITE BOUNDARY - That boundary defined by a 1046 meter (0.65 mile) radius around the plant.

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.

to 0CAN091801

Page 50 of 240 Attachment 1 - Emergency Action Level Technical Bases

Reference(s):

1. OP-1905.002 Offsite Emergency Monitoring 2. NEI 99-01 AG1

to 0CAN091801

Page 51 of 240 Attachment 1 - Emergency Action Level Technical Bases

Category: A - Abnormal Rad Levels / Rad Effluent Subcategory: 2 - Irradiated Fuel Event

Initiating Condition: UNPLANNED loss of water level above irradiated fuel EAL: AU2.1 Unusual Event UNPLANNED water level drop in the REFUELING PATHWAY as indicated by low water level alarm, visual observation, or BWST[RWT] level drop due to makeup demands AND UNPLANNED rise in corresponding area radiation levels as indicated by any of the following radiation monitors: Unit 1 o RE-8009 Spent Fuel Area o RE-8017 Fuel Handling Area Unit 2 o 2RE-8914 Spent Fuel Area o 2RE-8915 Spent Fuel Area o 2RE-8916 Spent Fuel Area o 2RE-8912 Containment Incore Instrumentation 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 - All the cavities, tubes, canals and pools through which irradiated fuel may be moved, but not including the reactor vessel.

to 0CAN091801

Page 52 of 240 Attachment 1 - Emergency Action Level Technical Bases

Basis:

This IC addresses a drop 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 drop 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 a rise 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 rise 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 AA2.

Reference(s):

1. OP-1203.050 Unit 1 Spent Fuel Pool Emergencies 2. OP-2203.002 Spent Fuel Pool Emergencies
3. 1SAR 11.2.5 Area Radiation Monitoring Systems Table 11-15 Area Radiation Monitors
4. 2SAR 12.1.4 Area Radiation Monitoring System 5. NEI 99-01 AU2

to 0CAN091801

Page 53 of 240 Attachment 1 - Emergency Action Level Technical Bases

Category: A - Abnormal Rad Levels / Rad Effluent Subcategory: 2 - Irradiated Fuel Event

Initiating Condition: Significant lowering of water level above, or damage to, irradiated fuel EAL: AA2.1 Alert IMMINENT uncovery of irradiated fuel in the REFUELING PATHWAY.

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 ANO ISFSI, the Confinement Boundary is comprised of either the Multi-assembly Sealed Basket (MSB) (SNC System) or Multi-Purpose Canister (MPC) (Holtec System).

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.

REFUELING PATHWAY - All the cavities, tubes, canals and pools through which irradiated fuel may be moved, but not including the reactor vessel.

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 REFUELING PATHWAY. 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 to irradiated fuel that is licensed for dry storage up to the point that the loaded storage cask is sealed. Once sealed, damage to a loaded cask causing loss of the CONFINEMENT BOUNDARY is classified in accordance with IC EU1.

This EAL escalates from AU2.1 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. to 0CAN091801

Page 54 of 240 Attachment 1 - Emergency Action Level Technical Bases

While an area radiation monitor could detect a rise 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 with Recognition Category C during the Cold Shutdown and Refueling modes.

Escalation of the emergency classification level would be via IC AS1.

Reference(s):

1. NEI 99-01 AA2

to 0CAN091801

Page 55 of 240 Attachment 1 - Emergency Action Level Technical Bases

Category: A - Abnormal Rad Levels / Rad Effluent Subcategory: 2 - Irradiated Fuel Event

Initiating Condition: Significant lowering of water level above, or damage to, irradiated fuel EAL: AA2.2 Alert Damage to irradiated fuel resulting in a release of radioactivity AND High alarm on any Table 1[2]A-2 radiation monitor.

Table 1A-2 Unit 1 Fuel Damage Radiation Monitors RE-8009 Spent Fuel Area RE-8017 Fuel Handling RE-8060 Containment High Range Radiation Monitor RE-8061 Containment High Range Radiation Monitor RX-9820 (SPING 1) Containment Purge RX-9825 (SPING 2) Radwaste Area RX-9830 (SPING 3) Fuel Handling Area Table 2A-2 Unit 2 Fuel Damage Radiation Monitors 2RE-8905 Containment Equipment Hatch Area 2RE-8909 Containment Personnel Access Area 2RE-8912 Containment Incore Inst. 2RE-8914 Spent Fuel Area 2RE-8915 Spent Fuel Area 2RE-8916 Spent Fuel Area 2RE-8925-1 Containment High Range Radiation Monitor 2RE-8925-2 Containment High Range Radiation Monitor 2RX-9820 (SPING 5) Containment Purge 2RX-9825 (SPING 6) Radwaste Area 2RX-9830 (SPING 7) Fuel Handling Area to 0CAN091801

Page 56 of 240 Attachment 1 - Emergency Action Level Technical Bases

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 ANO ISFSI, the Confinement Boundary is comprised of either the Multi-assembly Sealed Basket (MSB) (SNC System) or Multi-Purpose Canister (MPC) (Holtec System).

VALID - An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment.

Basis: This EAL addresses events that have caused actual damage to an irradiated fuel assembly.

These events present radiological safety challenges to plant personnel and are precursors to a release of radioactivity to the environment. As such, they represent an actual or potential substantial degradation of the level of safety of the plant.

This EAL applies to irradiated fuel that is licensed for dry storage up to the point that the loaded storage cask is sealed. Once sealed, damage to a loaded cask causing loss of the CONFINEMENT BOUNDARY is classified in accordance with IC EU1.

This EAL addresses a release of radioactive material caused by mechanical damage to irradiated fuel. Damaging events may include the dropping, bumping or binding of an assembly, or dropping a heavy load onto an assembly. A rise in readings on radiation monitors should be considered in conjunction with in-plant reports or observations of a potential fuel damaging event (e.g., a fuel handling accident).

Escalation of the emergency classification level would be via IC AS1.

Reference(s):

1. OP-1203.050 Unit 1 Spent Fuel Pool Emergencies 2. OP-1305.001 Radiation Monitoring System Check and Test 3. OP-2203.002 Spent Fuel Pool Emergencies 4. OP-1604.051 Eberline Radiation Monitoring System
5. OP-2304.133 Containment High Range Radiation Monitor Calibration
6. Offsite Dose Calculation Manual
7. NEI 99-01 AA2 to 0CAN091801

Page 57 of 240 Attachment 1 - Emergency Action Level Technical Bases

Category: A - Abnormal Rad Levels / Rad Effluent Subcategory: 2 - Irradiated Fuel Event

Initiating Condition: Significant lowering of water level above, or damage to, irradiated fuel EAL: AA2.3 Alert Lowering of spent fuel pool level to 387.0 ft.[389.5 ft.] (Alarm 2) on LIT-2020-3(4)

[2LIT-2020-1(2)]

Mode Applicability:

All Definition(s):

None Basis: This EAL addresses events that have caused a significant lowering of water level within the spent fuel pool. These events present radiological safety challenges to plant personnel and are precursors to a release of radioactivity to the environment. As such, they represent an actual or potential substantial degradation of the level of safety of the plant.

Spent fuel pool water level at this value is within the lower end of the level range necessary to prevent significant dose consequences from direct gamma radiation to personnel performing operations in the vicinity of the spent fuel pool. This condition reflects a significant loss of spent fuel pool water inventory and thus it is also a precursor to a loss of the ability to adequately cool the irradiated fuel assembles stored in the pool.

Escalation of the emergency classification level would be via IC AS1 or AS2.

Post-Fukushima order EA-12-051 required the installation of reliable SFP level indication capable of identifying normal level (Level 1), SFP level 10 ft. above the top of the fuel racks (Level 2) and SFP level at the top of the fuel racks (Level 3) (ref. 1, 2).

Reference(s):

1. MOHR-ANO-1, ANO-1 SFPI (Level) Configuration, Sheet 1, Revision 0 2. MOHR-ANO-2, ANO-2 SFPI (Level) Configuration, Sheet 1, Revision 0
3. NEI 99-01 AA2 to 0CAN091801

Page 58 of 240 Attachment 1 - Emergency Action Level Technical Bases

Category: A - Abnormal Rad Levels / Rad Effluent Subcategory: 2 - Irradiated Fuel Event

Initiating Condition: Spent fuel pool level at the top of the fuel racks EAL: AS2.1 Site Area Emergency Lowering of spent fuel pool level to 377.0 ft.[379.5 ft.] (Alarm 3) on LIT-2020-3(4)

[2LIT-2020-1(2)]

Mode Applicability:

All Definition(s):

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

Basis:

This EAL addresses a significant loss of spent fuel pool inventory control and makeup capability leading to IMMINENT fuel damage. This condition entails major failures of plant functions needed for protection of the public and thus warrant a Site Area Emergency declaration.

It is recognized that this IC would likely not be met until well after another Site Area Emergency IC was met; however, it is included to provide classification diversity.

Escalation of the emergency classification level would be via IC AG1 or AG2.

Post-Fukushima order EA-12-051 required the installation of reliable SFP level indication capable of identifying normal level (Level 1), SFP level 10 ft. above the top of the fuel racks (Level 2) and SFP level at the top of the fuel racks (Level 3) (ref. 1, 2).

Reference(s):

1. MOHR-ANO-1, ANO-1 SFPI (Level) Configuration, Sheet 1, Revision 0 2. MOHR-ANO-2, ANO-2 SFPI (Level) Configuration, Sheet 1, Revision 0 3. NEI 99-01 AS2

to 0CAN091801

Page 59 of 240 Attachment 1 - Emergency Action Level Technical Bases

Category: A - Abnormal Rad Levels / Rad Effluent Subcategory: 2 - Irradiated Fuel Event

Initiating Condition: Spent fuel pool level cannot be restored to at least the top of the fuel racks for 60 minutes or longer

EAL: AG2.1 General Emergency Spent fuel pool level cannot be restored to at least 377.0 ft.[379.5 ft.] (Alarm 3) on LIT-2020-3(4)[2LIT-2020-1(2)] for 60 min. (Note 1)

Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded.

Mode Applicability:

All

Definition(s):

None Basis: This EAL addresses a significant loss of spent fuel pool inventory control and makeup capability leading to a prolonged uncovery of spent fuel. This condition will lead to fuel damage and a radiological release to the environment.

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

Post-Fukushima order EA-12-051 required the installation of reliable SFP level indication capable of identifying normal level (Level 1), SFP level 10 ft. above the top of the fuel racks (Level 2) and SFP level at the top of the fuel racks (Level 3) (ref. 1, 2).

Reference(s):

1. MOHR-ANO-1, ANO-1 SFPI (Level) Configuration, Sheet 1, Revision 0 2. MOHR-ANO-2, ANO-2 SFPI (Level) Configuration, Sheet 1, Revision 0 3. NEI 99-01 AG2 to 0CAN091801

Page 60 of 240 Attachment 1 - Emergency Action Level Technical Bases

Category: A - Abnormal Rad Levels / Rad Effluent Subcategory: 3 - Area Radiation Levels

Initiating Condition: Radiation levels that IMPEDE access to equipment necessary for normal plant operations, cooldown or shutdown

EAL: AA3.1 Alert Dose rate > 15 mR/hr in EITHER of the following areas: Control Room Central Alarm Station (by survey)

Mode Applicability:

All Definition(s):

IMPEDE(D) - Personnel access to a room or area is hindered to an extent that extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., requiring use of protective equipment, such as SCBAs, that is not routinely employed).

Basis: Areas that meet this threshold include the Control Room (CR) and the Central Alarm Station (CAS). The Control Room envelope (Unit 1 and Unit 2) is monitored for excessive radiation by five detectors. These radiation detectors are RE-8001, 2RE-8001A, 2RE-8001B, 2RE-8750-1A, and 2RE-8750-1B (ref. 1). The CAS is included in this EAL because of its importance to permitting access to areas required to assure safe plant operations. There are no permanently installed area radiation monitors in CAS that may be used to assess this EAL threshold. Therefore, this threshold is evaluated using local radiation survey for this area.

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 should consider the cause of the rise in radiation levels and determine if another IC may be applicable.

Escalation of the emergency classification level would be via Recognition Category A, C or F

ICs.

to 0CAN091801

Page 61 of 240 Attachment 1 - Emergency Action Level Technical Bases

Reference(s):

1. STM 1-62 Radiation Monitoring 2. NEI 99-01 AA3

to 0CAN091801

Page 62 of 240 Attachment 1 - Emergency Action Level Technical Bases

Category: A - Abnormal Rad Levels / Rad Effluent Subcategory: 3 - Area Radiation Levels

Initiating Condition: Radiation levels that IMPEDE access to equipment necessary for normal plant operations, cooldown or shutdown

EAL: AA3.2 Alert An UNPLANNED event results in radiation levels that prohibit or IMPEDE access to any Table 1[2]A-3 room or area (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 1A-3 Unit 1 Safe Operation & Shutdown Rooms/Areas Room/Area Mode A-4 Switchgear Room 3, 4 Upper North Electrical Penetration Room 3, 4 Lower South Electrical Equipment Room 3, 4 Table 2A-3 Unit 2 Safe Operation & Shutdown Rooms/Areas Room/Area Mode Aux Building 317' Emergency Core Cooling Rooms 3, 4 Aux Building 317' Tendon Gallery Access 3, 4 Aux Building 335' Charging Pumps / MCC 2B-52 3, 4 Aux Building 354' MCC 2B-62 Area 3, 4 Emergency Diesel Generator Corridor 3, 4 Lower South Piping Penetration Room 3, 4 Aux Building 386' Containment Hatch 3, 4 to 0CAN091801

Page 63 of 240 Attachment 1 - Emergency Action Level Technical Bases

Mode Applicability:

3 - Hot Standby, 4 - Hot Shutdown Definition(s):

IMPEDE(D) - Personnel access to a room or area is hindered to an extent that extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., requiring use of protective equipment, such as SCBAs, that is not routinely employed).

UNPLANNED

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

Basis: 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 should consider the cause of the rise in radiation levels and determine if another IC may be applicable.

For AA3.2, 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 higher radiation levels. Access should be considered as IMPEDED if extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., installing temporary shielding, requiring use of non-routine protective equipment, requesting an extension in dose limits beyond normal administrative limits).

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

The plant is in an operating mode different than the mode specified for the affected room/area (i.e., entry is not required during the operating mode in effect at the time of the elevated radiation levels). For example, the plant is in Mode 1 when the radiation rise occurs, and the procedures used for normal operation, cooldown and shutdown do not require entry into the affected room until Mode 3. The higher 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 access control measures are of a conservative or precautionary nature, and would not actually prevent or IMPEDE a required action.

Escalation of the emergency classification level would be via Recognition Category A, C or F

ICs. to 0CAN091801

Page 64 of 240 Attachment 1 - Emergency Action Level Technical Bases

If the equipment in the listed room or area was already inoperable, or out-of-service, before the event occurred, then no emergency should be declared since the event will have no adverse impact beyond that already allowed by Technical Specifications at the time of the event.

The list of plant rooms or areas with entry-related mode applicability identified specify those rooms or areas that contain equipment which require a manual/local action as specified in operating procedures used for normal plant operation, cooldown and shutdown. Rooms or areas in which actions of a contingent or emergency nature would be performed (e.g., an action to address an off-normal or emergency condition such as emergency repairs, corrective measures or emergency operations) are not included. In addition, the list specifies the plant mode(s) during which entry would be required for each room or area (ref. 1).

EAL AA3.2 mode applicability has been limited to the mode limitations of Table 1[2]A-3 (Modes 3 and 4 only). Reference(s):

1. Attachment 2 Safe Operation & Shutdown Areas Tables 1[2]A-3 & 1[2]H-2 Bases 2. NEI 99-01 AA3

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Page 65 of 240 Attachment 1 - Emergency Action Level Technical Bases

Category C - Cold Shutdown / Refueling System Malfunction EAL Group: Cold Conditions (RCS temperature 200°F); EALs in this category are applicable only in one or more cold operating modes.

Category C EALs are directly associated with cold shutdown or refueling system safety functions. Given the variability of plant configurations (e.g., systems out-of-service for maintenance, containment open, reduced AC power redundancy, time since shutdown) during these periods, the consequences of any given init iating event can vary greatly. For example, a loss of decay heat removal capability that occurs at the end of an extended outage has less significance than a similar loss occurring during the first week after shutdown. Compounding these events is the likelihood that instrumentation necessary for assessment may also be inoperable. The cold shutdown and refueling system malfunction EALs are based on performance capability to the extent possible with consideration given to RCS integrity, CONTAINMENT CLOSURE, and fuel clad integrity for the applicable operating modes (5 - Cold Shutdown, 6 - Refueling, DEF - Defueled).

The events of this category pertain to the following subcategories:

1. RCS Level RCS water level is directly related to the status of adequate core cooling and, therefore, fuel clad integrity.
2. Loss of Vital AC Power Loss of vital 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 vital buses.
3. RCS Temperature Uncontrolled or inadvertent temperature or pressure rises are indicative of a potential loss of

safety functions.

4. Loss of Vital DC Power Loss of emergency plant electrical power can compromise plant SAFETY SYSTEM operability including decay heat removal and emergency core cooling systems which may be necessary to ensure fission product barrier integrity. This category includes loss of power to or degraded voltage on the 125V DC vital buses.
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. to 0CAN091801

Page 66 of 240 Attachment 1 - Emergency Action Level Technical Bases

Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - RCS Level

Initiating Condition: UNPLANNED loss of RCS inventory EAL: CU1.1 Unusual Event UNPLANNED loss of reactor coolant results in RCS water level less than a required lower limit for 15 min. (Note 1)

Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded.

Mode Applicability:

5 - Cold Shutdown, 6 - Refueling

Definition(s):

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

Basis: This EAL addresses the inability to restore and maintain water level to a required minimum level (or the lower limit of a level band), or a loss of the ability to monitor RCS level concurrent with indications of coolant leakage. Either of these conditions is considered to be a potential degradation of the level of safety of the plant.

Refueling evolutions that lower RCS water inventory are carefully planned and controlled. An UNPLANNED event that results in water level lowering below a procedurally required limit warrants the declaration of an Unusual Event due to the reduced water inventory that is available to keep the core covered.

This EAL recognizes that the minimum required RCS level can change several times during the course of a refueling outage as different plant configurations and system lineups are implemented. This EAL is met if the minimum level, specified for the current plant conditions, 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.

to 0CAN091801

Page 67 of 240 Attachment 1 - Emergency Action Level Technical Bases

The 15-minute threshold duration allows sufficient time for prompt operator actions to restore and maintain the expected water level. This criterion excludes transient conditions causing a brief lowering of water level.

Continued loss of RCS inventory may result in escalation to the Alert emergency classification

level via either IC CA1 or CA3.

Reference(s):

1. OP-1015.002 Decay Heat Removal and LTOP System 2. OP-1015.008 Unit 2 SDC Control
3. NEI 99-01 CU1 to 0CAN091801

Page 68 of 240 Attachment 1 - Emergency Action Level Technical Bases

Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - RCS Level

Initiating Condition: UNPLANNED loss of RCS inventory EAL: CU1.2 Unusual Event RCS level cannot be monitored AND EITHER:

UNPLANNED rise in any Table 1[2]C-1 sump/tank level due to loss of RCS inventory Visual observation of UNISOLABLE RCS leakage Table 1C-1 Unit 1 Sumps / Tanks Reactor Building Sump Reactor Drain Tank Aux. Building Equipment Drain Tank Aux. Building Sump Quench Tank Table 2C-1 Unit 2 Sumps / Tanks CNTMT Sump Reactor Drain Tank LRW Waste Tank (2T-20) Holdup Tank Aux. Building Sump Quench Tank Mode Applicability:

5 - Cold Shutdown, 6 - Refueling to 0CAN091801

Page 69 of 240 Attachment 1 - Emergency Action Level Technical Bases

Definition(s):

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

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

Basis: This IC addresses the inability to restore and maintain water level to a required minimum level (or the lower limit of a level band), or a loss of the ability to monitor RCS level concurrent with indications of coolant leakage. Either of these conditions is considered to be a potential degradation of the level of safety of the plant.

Refueling evolutions that lower RCS water inventory are carefully planned and controlled. An UNPLANNED event that results in water level lowering below a procedurally required limit warrants the declaration of an Unusual Event due to the reduced water inventory that is available to keep the core covered.

This EAL addresses a condition where all means to determine RCS level have been lost. In this condition, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels (Table C-1). Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the RCS.

Continued loss of RCS inventory may result in escalation to the Alert emergency classification

level via either IC CA1 or CA3.

Reference(s):

1. OP-1203.039 Excess RCS Leakage 2. OP-2203.016 Excess RCS Leakage
3. NEI 99-01 CU1 to 0CAN091801

Page 70 of 240 Attachment 1 - Emergency Action Level Technical Bases

Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - RCS Level

Initiating Condition: Significant Loss of RCS inventory EAL: CA1.1 Alert Loss of RCS inventory as indicated by EITHER: RVLMS Levels 1 through 8[1 through 5] indicate DRY Reactor vessel level < 370.2 ft. (LT-1195/LT-1196)[< 24 in. (L4791/L4792)] (minimum level for DHR operation @ 1000 gpm)[(minimum level for SDC operation)]

Mode Applicability:

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

None

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, a lowering of RPV water level below the specified level indicates that operator actions have not been successful in restoring and maintaining RCS water level. The heat-up rate of the coolant will rise as the available water inventory is reduced. A continuing drop in

water level will lead to core uncovery.

Although related, this EAL is concerned with the loss of RPV inventory and not the potential concurrent effects on systems needed for decay heat removal (e.g., loss of a Decay Heat Removal suction point). A rise in RCS temperature caused by a loss of decay heat removal capability is evaluated under IC CA3.

If water level continues to lower, then escalation to Site Area Emergency would be via IC CS1.

A loss of DHR/SDC will occur at approximately RLVMS Level 8 (Unit 1) or RVLMS Level 5 (Unit 2). However, RVLMS may not be available in the cold shutdown modes. Redundant means of level indication is provided in these modes and included in this EAL. The point at which a loss of DHR/SDC is likely to occur is 370.2 ft. (Unit 1) or 24 in. (Unit 2) as indicated in the respective Control Rooms. The value selected for ANO-1 is based on 1000 gpm DHR flow to 0CAN091801

Page 71 of 240 Attachment 1 - Emergency Action Level Technical Bases

which is the flow rate at which the low flow alarm is received. The ANO-2 value is the proceduralized minimum value. Below these levels, a loss of suction to decay heat removal systems will occur (ref. 1, 2, 3). The inability to restore and maintain level after reaching this value would be indicative of a failure of the RCS barrier.

Reference(s):

1. OP-1104.004 Decay Heat Removal Operating Procedure 2. OP-1105.008 Inadequate Core Cooling Monitor and Display
3. OP-2105.003 Reactor Vessel Level Monitoring System Operations
4. OP-2203.029 Loss of Shutdown Cooling
5. Calculation No. 90-E-0116-01 ANO-2 EOP Setpoint Basis Document, Setpoints R.3 and R.9 6. NEI 99-01 CA1

to 0CAN091801

Page 72 of 240 Attachment 1 - Emergency Action Level Technical Bases

Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - RCS Level

Initiating Condition: Significant Loss of RCS inventory EAL: CA1.2 Alert RCS level cannot be monitored for 15 min. (Note 1) AND EITHER:

UNPLANNED rise in any Table 1[2]C-1 sump/tank level due to a loss of RCS inventory Visual observation of UNISOLABLE RCS leakage Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded.

Table 1C-1 Unit 1 Sumps / Tanks Reactor Building Sump Reactor Drain Tank Aux. Building Equipment Drain Tank Aux. Building Sump Quench Tank Table 2C-1 Unit 2 Sumps / Tanks CNTMT Sump Reactor Drain Tank LRW Waste Tank (2T-20) Holdup Tank Aux. Building Sump Quench Tank to 0CAN091801

Page 73 of 240 Attachment 1 - Emergency Action Level Technical Bases

Mode Applicability:

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

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

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

Basis: This IC addresses conditions that are precursors to a loss of the ability to adequately cool irradiated fuel (i.e., a precursor to a challenge to the fuel clad barrier). This condition represents a potential substantial reduction in the level of plant safety.

For this EAL, the inability to monitor RCS level may be caused by instrumentation and/or power failures, or water level dropping below the range of available instrumentation. If water level cannot be monitored, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the RCS.

The 15-minute duration for the loss of level indication was chosen because it is half of the EAL duration specified in IC CS1.

If the RCS inventory level continues to lower, then escalation to Site Area Emergency would be via IC CS1.

Reference(s):

1. OP-1203.039 Excess RCS Leakage 2. OP-2203.016 Excess RCS Leakage 3. NEI 99-01 CA1 to 0CAN091801

Page 74 of 240 Attachment 1 - Emergency Action Level Technical Bases

Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - RCS Level

Initiating Condition: Loss of RCS inventory affecting core decay heat removal capability EAL: CS1.1 Site Area Emergency CONTAINMENT CLOSURE not established AND RVLMS Levels 1 through 9[1 through 6] indicate DRY Mode Applicability:

5 - Cold Shutdown, 6 - Refueling

Definition(s):

CONTAINMENT CLOSURE - The action to secure primary containment and its associated structures, systems, and components as a functional barrier to fission product release under existing plant conditions.

As applied to ANO, CONTAINMENT CLOSURE must be capable of being set within 30 minutes. CONTAINMENT CLOSURE is set when the penetrations are isolated by manual or automatic isolation valve, blind flange, or equivalent.

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

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

Following an extended loss of core decay heat removal and inventory makeup, decay heat will cause reactor coolant boiling and a further reduction in reactor vessel level. If RCS/reactor vessel 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.

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Page 75 of 240 Attachment 1 - Emergency Action Level Technical Bases

This EAL addresses concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal; SECY 91-283, Evaluation of Shutdown and Low Power Risk Issues; NUREG-1449, Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the United States

and NUMARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management.

Escalation of the emergency classification level would be via IC CG1 or AG1.

Reference(s):

1. OP-1105.008 Inadequate Core Cooling Monitor and Display 2. OP-2105.003 Reactor Vessel Level Monitoring System Operations
3. NEI 99-01 CS1 to 0CAN091801

Page 76 of 240 Attachment 1 - Emergency Action Level Technical Bases

Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - RCS Level

Initiating Condition: Loss of RCS inventory affecting core decay heat removal capability EAL: CS1.2 Site Area Emergency [RVLMS Levels 1 through 7 indicate DRY OR] RCS level cannot be monitored for 30 min. (Note 1) AND Core uncovery is indicated by any of the following: UNPLANNED rise in any Table 1[2]C-1 sump/tank level of sufficient magnitude to indicate core uncovery Containment high range radiation monitor RE-8060/8061[2RE-8925-1/8925-2] reading

> 10 R/hr Erratic Source Range Monitor indication Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded.

Table 1C-1 Unit 1 Sumps / Tanks Reactor Building Sump Reactor Drain Tank Aux. Building Equipment Drain Tank Aux. Building Sump Quench Tank to 0CAN091801

Page 77 of 240 Attachment 1 - Emergency Action Level Technical Bases

Table 2C-1 Unit 2 Sumps / Tanks CNTMT Sump Reactor Drain Tank LRW Waste Tank (2T-20) Holdup Tank Aux. Building Sump Quench Tank Mode Applicability:

5 - Cold Shutdown, 6 - Refueling

Definition(s):

CONTAINMENT CLOSURE - The action to secure primary containment and its associated structures, systems, and components as a functional barrier to fission product release under existing plant conditions.

As applied to ANO, CONTAINMENT CLOSURE must be capable of being set within 30 minutes. CONTAINMENT CLOSURE is set when the penetrations are isolated by manual or automatic isolation valve, blind flange, or equivalent.

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.

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

Basis: When in service, the Unit 2 RVLMS can measure RCS level below the top of active fuel.

Level 7 DRY on this system is an indication of core uncovery.

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

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Page 78 of 240 Attachment 1 - Emergency Action Level Technical Bases

Following an extended loss of core decay heat removal and inventory makeup, decay heat will cause reactor coolant boiling and a further reduction in reactor vessel level. If RCS/reactor vessel 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 levels of EALs CS1.1 and CS1.2 reflect the fact that with CONTAINMENT CLOSURE established, there is a lower probability of a fission product release to the environment.

The 30-minute criterion is tied to a readily recognizable event start time (i.e., the total loss of ability to monitor level), and allows sufficient time to monitor, assess and correlate reactor and plant conditions to determine if core uncovery has actually occurred (i.e., to account for various accident progression and instrumentation uncertainties). It also allows sufficient time for performance of actions to terminate leakage, recover inventory control/makeup equipment and/or restore level monitoring.

The inability to monitor RCS level may be caused by instrumentation and/or power failures, or water level dropping below the range of available instrumentation. If water level cannot be monitored, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the RCS.

This EAL addresses concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal

SECY 91-283, Evaluation of Shutdown and Low Power Risk Issues; NUREG-1449, Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the United States
and NUMARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management.

Containment High Range Radiation Monitors RE-8060/8061 [2RE-8925-1/8925-2] are the site-specific radiation monitors that would be indicative of possible core uncovery in the Refueling mode. The dose rate due to core shine when the top of the core becomes uncovered should result in dose rates > 10 R/hr.

Escalation of the emergency classification level would be via IC CG1 or AG1.

Reference(s):

1. OP-1203.039 Excess RCS Leakage 2. OP-2203.016 Excess RCS Leakage
3. OP-2105.003 Reactor Vessel Level Monitoring System Operations 4. 1SAR Table 7-11 5. 2SAR 12.1.4.2
6. NEI 99-01 CS1 to 0CAN091801

Page 79 of 240 Attachment 1 - Emergency Action Level Technical Bases

Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - RCS Level

Initiating Condition: Loss of RCS inventory affecting fuel clad integrity with containment challenged

EAL: CG1.1 General Emergency - UNIT 2 ONLY RVLMS Levels 1 through 7 indicate DRY AND Any Containment Challenge indication, Table 1[2]C-2 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 1[2]C-2 Containment Challenge Indications CONTAINMENT CLOSURE not established (Note 6) Containment hydrogen concentration > 4% UNPLANNED rise in containment pressure Mode Applicability:

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

CONTAINMENT CLOSURE - The action to secure primary containment and its associated structures, systems, and components as a functional barrier to fission product release under existing plant conditions.

As applied to ANO, CONTAINMENT CLOSURE must be capable of being set within 30 minutes. CONTAINMENT CLOSURE is set when the penetrations are isolated by manual or automatic isolation valve, blind flange, or equivalent.

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.

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. to 0CAN091801

Page 80 of 240 Attachment 1 - Emergency Action Level Technical Bases

Basis: When in service, the Unit 2 RVLMS can measure RCS level below the top of active fuel.

Level 7 DRY on this system is an indication of core uncovery.

This IC addresses the inability to restore and maintain reactor vessel level above the top of active fuel with containment challenged. This condition represents actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity. Releases can be reasonably expected to exceed EPA PAG exposure levels offsite for more than the immediate site area.

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

With CONTAINMENT CLOSURE not established, there is a high potential for a direct and unmonitored release of radioactivity to the environment. If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, then declaration of a General Emergency is not required.

The existence of an explosive mixture means, at a minimum, that the containment atmospheric hydrogen concentration is sufficient to support a hydrogen burn (i.e., at the lower deflagration limit). A hydrogen burn will raise containment pressure and could result in collateral equipment damage leading to a loss of containment integrity. It therefore represents a challenge to Containment integrity.

In the early stages of a core uncovery event, it is unlikely that hydrogen buildup due to 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.

This EAL addresses concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal

SECY 91-283, Evaluation of Shutdown and Low Power Risk Issues; NUREG-1449, Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the United States
and NUMARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management.

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Page 81 of 240 Attachment 1 - Emergency Action Level Technical Bases

Reference(s):

1. OP-1203.039 Excess RCS Leakage 2. OP-2203.016 Excess RCS Leakage
3. OP-2105.003 Reactor Vessel Level Monitoring System Operations
4. 1SAR Table 7-11
5. 2SAR 12.1.4.2 6. Unit 1 SAMG Figure III-1B 7. Unit 2 SAMG Phase 1 Instructions, Containment Flowchart
8. NEI 99-01 CG1

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Page 82 of 240 Attachment 1 - Emergency Action Level Technical Bases

Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - RCS Level

Initiating Condition: Loss of RCS inventory affecting fuel clad integrity with containment challenged

EAL: CG1.2 General Emergency RCS level cannot be monitored for 30 min. (Note 1) AND Core uncovery is indicated by any of the following: UNPLANNED rise in any Table 1[2]C-1 sump/tank level of sufficient magnitude to indicate core uncovery Containment High Range Radiation Monitor RE-8060/8061[2RE-8925-1/8925-2] reading > 10 R/hr Erratic Source Range Monitor indication AND Any Containment Challenge indication, Table 1[2]C-2 Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is 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 1C-1 Unit 1 Sumps / Tanks Reactor Building Sump Reactor Drain Tank Aux. Building Equipment Drain Tank Aux. Building Sump Quench Tank to 0CAN091801

Page 83 of 240 Attachment 1 - Emergency Action Level Technical Bases

Table 2C-1 Unit 2 Sumps / Tanks CNTMT Sump Reactor Drain Tank LRW Waste Tank (2T-20) Holdup Tank Aux. Building Sump Quench Tank Table 1[2]C-2 Containment Challenge Indications CONTAINMENT CLOSURE not established (Note 6) Containment hydrogen concentration > 4% UNPLANNED rise in containment pressure Mode Applicability:

5 - Cold Shutdown, 6 - Refueling

Definition(s):

CONTAINMENT CLOSURE - The action to secure primary containment and its associated structures, systems, and components as a functional barrier to fission product release under existing plant conditions.

As applied to ANO, CONTAINMENT CLOSURE must be capable of being set within 30 minutes. CONTAINMENT CLOSURE is set when the penetrations are isolated by manual or automatic isolation valve, blind flange, or equivalent.

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.

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

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Page 84 of 240 Attachment 1 - Emergency Action Level Technical Bases

Basis:

When in service, the Unit 2 RVLMS can measure RCS level below the top of active fuel. Level 7 DRY on this system is an indication of core uncovery.

This IC addresses the inability to restore and maintain reactor vessel level above the top of active fuel with containment challenged. This condition represents actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity. Releases can be reasonably expected to exceed EPA PAG exposure levels offsite for more than the immediate site area.

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

With CONTAINMENT CLOSURE not established, there is a high potential for a direct and unmonitored release of radioactivity to the environment. If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, then declaration of a General Emergency is not required.

The existence of an explosive mixture means, at a minimum, that the containment atmospheric hydrogen concentration is sufficient to support a hydrogen burn (i.e., at the lower deflagration limit). A hydrogen burn will raise containment pressure and could result in collateral equipment damage leading to a loss of containment integrity. It therefore represents a challenge to Containment integrity.

In the early stages of a core uncovery event, it is unlikely that hydrogen buildup due to core uncovery could result in an explosive gas mixture in containment. If all installed hydrogen gas monitors are out-of-service during an event leading to fuel cladding damage, it may not be possible to obtain a containment hydrogen gas concentration reading as ambient conditions within the containment will preclude personnel access. During periods when installed containment hydrogen gas monitors are out-of-service, operators may use the other listed indications to assess whether or not containment is challenged.

The 30-minute criterion is tied to a readily recognizable event start time (i.e., the total loss of ability to monitor level), and allows sufficient time to monitor, assess and correlate reactor and plant conditions to determine if core uncovery has actually occurred (i.e., to account for various accident progression and instrumentation uncertainties). It also allows sufficient time for performance of actions to terminate leakage, recover inventory control/makeup equipment and/or restore level monitoring.

The inability to monitor RCS level may be caused by instrumentation and/or power failures, or water level dropping below the range of available instrumentation. If water level cannot be monitored, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the RCS.

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Page 85 of 240 Attachment 1 - Emergency Action Level Technical Bases

Containment High Range Radiation Monitor RE-8060/8061 [2RE-8925-1/8925-2] are the site-specific radiation monitors that would be indicative of possible core uncovery in the Refueling mode. The dose rate due to core shine when the top of the core becomes uncovered should result in dose rates > 10 R/hr.

This EAL addresses concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal

SECY 91-283, Evaluation of Shutdown and Low Power Risk Issues; NUREG-1449, Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the United States
and NUMARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management. Reference(s)
1. OP-1203.039 Excess RCS Leakage 2. OP-2203.016 Excess RCS Leakage
3. OP-2105.003 Reactor Vessel Level Monitoring System Operations
4. 1SAR Table 7-11
5. 2SAR 12.1.4.2
6. Unit 1 SAMG Figure III-1B 7. Unit 2 SAMG Phase 1 Instructions, Containment Flowchart 8. NEI 99-01 CG1

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Page 86 of 240 Attachment 1 - Emergency Action Level Technical Bases

Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 2 - Loss of Vital AC Power

Initiating Condition: Loss of all but one AC power source to vital buses for 15 minutes or longer EAL: CU2.1 Unusual Event AC power capability, Table 1[2]C-3, to vital 4.16 KV buses A3[2A3] and A4[2A4] reduced to a single power source for 15 min. (Note 1) AND Any additional single power source failure will result in loss of all AC power to SAFETY SYSTEMS Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded.

Table 1C-3 Unit 1 AC Power Sources Offsite Startup Transformer No. 1 Startup Transformer No. 2 Unit Auxiliary Transformer (from 22 KV switchyard)

Onsite DG1 DG2 AAC Gen to 0CAN091801

Page 87 of 240 Attachment 1 - Emergency Action Level Technical Bases

Table 2C-3 Unit 2 AC Power Sources Offsite Startup Transformer No. 3 Startup Transformer No. 2 Unit Auxiliary Transformer (backfed from main transformer)

Onsite 2DG1 2DG2 AAC Gen Mode Applicability:

5 - Cold Shutdown, 6 - Refueling, DEF - 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 10 CFR 50.2):

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: 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 greater 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. to 0CAN091801

Page 88 of 240 Attachment 1 - Emergency Action Level Technical Bases

An "AC power source" is a source recognized in AOPs and EOPs, and capable of supplying required power to a vital bus. Some examples of this condition are presented below.

A loss of all offsite power with a concurrent failure of all but one vital power source. A loss of all offsite power and loss of all vital power sources with a single train of vital buses being back-fed from the unit main generator. A loss of vital power sources (e.g., onsite diesel generators) with a single train of vital 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.

This EAL is the cold condition equivalent of the hot condition EAL SA1.1.

Reference(s):

1. 1SAR Figure 8-1 Station Single Line Diagram 2. OP-1202.007 Degraded Power
3. OP-1202.008 Blackout
4. OP-2104.037 Alternate AC Diesel Generator Operations
5. 2SAR Figure 8.3-1 Station Single Line Diagram 6. OP-2202.007 Loss of Off-Site Power 7. OP-2202.008 Station Blackout
8. OP-2107.006 Backfeed of Unit Auxiliary Transformer
9. NEI 99-01 CU2 to 0CAN091801

Page 89 of 240 Attachment 1 - Emergency Action Level Technical Bases

Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 2 - Loss of Vital AC Power

Initiating Condition: Loss of all offsite and all onsite AC power to vital buses for 15 minutes or longer

EAL: CA2.1 Alert Loss of all offsite and all onsite AC power to vital 4.16 KV buses A3[2A3] and A4[2A4] for 15 min. (Note 1)

Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded.

Mode Applicability:

5 - Cold Shutdown, 6 - Refueling, DEF - 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 10 CFR 50.2):

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:

Although the AAC may be considered available, it will not prevent declaration of this EAL unless it is powering a vital bus within the 15-minute time period of the EAL.

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. Mitigative strategies using non-safety related power sources (FLEX generators, etc.) may be effective in supplying power to these buses. These power sources must be controlled in to 0CAN091801

Page 90 of 240 Attachment 1 - Emergency Action Level Technical Bases

accordance with abnormal or emergency operating procedures, or beyond design basis accident response guidelines (e.g., FLEX support guidelines) and must be capable (alone or in combination) of supplying power for long term decay heat removal systems.

When in the cold shutdown, refueling, or defueled mode, this condition is not classified as a Site Area Emergency because of the greater time available to restore a vital bus to service.

Additional time is available due to the reduced core decay heat load, and the lower temperatures and pressures in various plant systems. Thus, when in these modes, this condition represents an actual or potential substantial degradation of the level of safety of the plant.

Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.

Escalation of the emergency classification level would be via IC CS1 or AS1.

This EAL is the cold condition equivalent of the hot condition EAL SS1.1.

Reference(s):

1. 1SAR Figure 8-1 Station Single Line Diagram 2. OP-1202.007 Degraded Power
3. OP-1202.008 Blackout
4. OP-2104.037 Alternate AC Diesel Generator Operations 5. 2SAR Figure 8.3-1 Station Single Line Diagram 6. OP-2202.007 Loss of Off-Site Power
7. OP-2202.008 Station Blackout
8. OP-2107.006 Backfeed of Unit Auxiliary Transformer
9. NEI 99-01 CU2 to 0CAN091801

Page 91 of 240 Attachment 1 - Emergency Action Level Technical Bases

Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 3 - RCS Temperature

Initiating Condition: UNPLANNED rise in RCS temperature EAL: CU3.1 Unusual Event UNPLANNED rise in RCS temperature to > 200°F Mode Applicability:

5 - Cold Shutdown, 6 - Refueling

Definition(s):

CONTAINMENT CLOSURE - The action to secure primary containment and its associated structures, systems, and components as a functional barrier to fission product release under existing plant conditions.

As applied to ANO, CONTAINMENT CLOSURE must be capable of being set within 30 minutes. CONTAINMENT CLOSURE is set when the penetrations are isolated by manual or automatic isolation valve, blind flange, or equivalent.

UNPLANNED -

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

Basis: This IC addresses an UNPLANNED rise 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 should also refer to EAL CA3.1.

A momentary UNPLANNED excursion above the Technical Specification cold shutdown temperature limit when the heat removal function is available does not warrant a classification.

This EAL involves a loss of decay heat removal capability, or an addition of heat to the RCS in excess of that which can currently be removed, such that reactor coolant temperature cannot be maintained below the cold shutdown temperature limit specified in Technical Specifications.

During this condition, there is no immediate threat of fuel damage because the core decay heat load has been reduced since the cessation of power operation.

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Page 92 of 240 Attachment 1 - Emergency Action Level Technical Bases

During an outage, the level in the reactor vessel will normally be maintained at or above the reactor vessel flange. Refueling evolutions that lower water level below the reactor vessel flange are carefully planned and controlled. A loss of forced decay heat removal at lowered inventory may result in a rapid rise in reactor coolant temperature depending on the time after

shutdown.

Escalation to Alert would be via IC CA1 based on an inventory loss or IC CA3 based on exceeding plant configuration-specific time criteria.

Reference(s):

1. Unit 1 and Unit 2 Technical Specifications Table 1.1-1 2. NEI 99-01 CU3

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Page 93 of 240 Attachment 1 - Emergency Action Level Technical Bases

Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 3 - RCS Temperature

Initiating Condition: UNPLANNED rise in RCS temperature EAL: CU3.2 Unusual Event Loss of all RCS temperature and RCS level indication for 15 min. (Note 1)

Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded.

Mode Applicability:

5 - Cold Shutdown, 6 - Refueling

Definition(s):

CONTAINMENT CLOSURE - The action to secure primary containment and its associated structures, systems, and components as a functional barrier to fission product release under existing plant conditions.

As applied to ANO, CONTAINMENT CLOSURE must be capable of being set within 30 minutes. CONTAINMENT CLOSURE is set when the penetrations are isolated by manual or automatic isolation valve, blind flange, or equivalent.

UNPLANNED -

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

Basis: This EAL addresses the inability to determine RCS temperature and level, and represents a potential degradation of the level of safety of the plant. If the RCS is not intact and CONTAINMENT CLOSURE is not established during this event, the Emergency Director should also refer to EAL CA3.1.

This EAL reflects a condition where there has been a significant loss of instrumentation capability necessary to monitor RCS conditions and operators would be unable to monitor key parameters necessary to assure core decay heat removal. During this condition, there is no immediate threat of fuel damage because the core decay heat load has been reduced since the cessation of power operation.

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Page 94 of 240 Attachment 1 - Emergency Action Level Technical Bases

Fifteen minutes was selected as a threshold to exclude transient or momentary losses of indication.

Escalation to Alert would be via IC CA1 based on an inventory loss or IC CA3 based on exceeding plant configuration-specific time criteria.

Reference(s):

1. NEI 99-01 CU3

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Page 95 of 240 Attachment 1 - Emergency Action Level Technical Bases

Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 3 - RCS Temperature

Initiating Condition: Inability to maintain plant in cold shutdown EAL: CA3.1 Alert UNPLANNED rise in RCS temperature to > 200°F for > Table 1[2]C-4 duration (Note 1)

OR UNPLANNED RCS pressure rise > 10 psig (this EAL does not apply during water-solid plant conditions)

Note 1: The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded.

Table 1[2]C-4 RCS Heat-up Duration Thresholds RCS Status CONTAINMENT CLOSURE Status Heat-up Duration Intact (but not lowered inventory) N/A 60 min.*

Not intact OR lowered inventory established 20 min.*

not established 0 min.

  • If an RCS heat removal system is in operation within this time frame and RCS temperature is being reduced, the EAL is not applicable.

Mode Applicability:

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

CONTAINMENT CLOSURE - The action to secure primary containment and its associated structures, systems, and components as a functional barrier to fission product release under existing plant conditions.

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Page 96 of 240 Attachment 1 - Emergency Action Level Technical Bases

As applied to ANO, CONTAINMENT CLOSURE must be capable of being set within 30 minutes. CONTAINMENT CLOSURE is set when the penetrations are isolated by manual or automatic isolation valve, blind flange, or equivalent.

UNPLANNED -

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

Basis: In the absence of reliable RCS temperature indication, classification should be based on the RCS pressure rise criteria when the RCS is intact in Mode 5 or based on time to boil data when in Mode 6 or the RCS is not intact in Mode 5.

This EAL 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 a rise in RCS temperature when CONTAINMENT CLOSURE is established but the RCS is not intact, or RCS inventory is reduced (e.g., lowered inventory operation). The 20-minute criterion was included to allow time for operator action to address the temperature rise.

The RCS Heat-up Duration Thresholds table also addresses a rise 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 rise without a substantial degradation in plant safety.

Finally, in the case where there is a rise in RCS temperature, the RCS is not intact or is at lowered inventory and CONTAINMENT CLOSURE is not established, no heat-up duration is allowed (i.e., 0 minutes). This is because 1) the evaporated reactor coolant may be released directly into the containment atmosphere and subsequently to the environment, and 2) there is reduced reactor coolant inventory above the top of irradiated fuel.

The RCS pressure rise 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 AS1.

Reference(s):

1. Unit 1 and Unit 2 Technical Specifications Table 1.1-1 2. NEI 99-01 CA3 to 0CAN091801

Page 97 of 240 Attachment 1 - Emergency Action Level Technical Bases

Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 4 - Loss of Vital DC Power

Initiating Condition: Loss of Vital DC power for 15 minutes or longer EAL: CU4.1 Unusual Event Indicated voltage is < 105 VDC on vital 125 VDC buses for 15 min. (Note 1)

Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded.

Mode Applicability:

5 - Cold Shutdown, 6 - Refueling

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 10 CFR 50.2).

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: Unit 1 batteries D06 and D07 and Unit 2 batteries 2D11 and 2D12 contain 58 cells each with a minimum cell voltage of 1.81 V or 105 VDC.

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 raise 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.

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As used in this EAL, "required" means the vital DC buses necessary to support operation of the in-service, or operable, train or trains of SAFETY SYSTEM equipment. For example, if Train A is out-of-service (inoperable) for scheduled outage maintenance work and Train B is in-service (operable), then a loss of Vital DC power affecting Train B would require the declaration of an Unusual Event. A loss of Vital DC power to Train A would not warrant an emergency classification.

Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.

Depending upon the event, escalation of the emergency classification level would be via IC CA1 or CA3, or an IC in Recognition Category A.

This EAL is the cold condition equivalent of the hot condition EAL SS2.1.

Reference(s):

1. 1SAR 8.3.2.1.1 Batteries 2. 2SAR 8.3.2.1.1 Batteries 3. NEI 99-01 CU4

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Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 5 - Loss of Communications

Initiating Condition: Loss of all onsite or offsite communications capabilities EAL: CU5.1 Unusual Event Loss of all Table 1[2]C-5 onsite communication methods OR Loss of all Table 1[2]C-5 State and local agency communication methods OR Loss of all Table 1[2]C-5 NRC communication methods Table 1[2]C-5 Communication Methods System Onsite ORO NRC Station radio system X ANO plant phone system X Gaitronics X Telephone Systems: Commercial Microwave Satellite VOIP X X Emergency Notification System (ENS) X Mode Applicability:

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

None to 0CAN091801

Page 100 of 240 Attachment 1 - Emergency Action Level Technical Bases

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 State and local agencies and the NRC.

This IC should be assessed only when extraordinary means are being utilized to make communications possible (e.g., use of non-plant, privately owned equipment, relaying of on-site information via individuals or multiple radio transmission points, individuals being sent to offsite locations, etc.).

The first EAL condition addresses a total loss of the communications methods used in support of routine plant operations.

The second EAL condition addresses a total loss of the communications methods used to notify all State and local agencies of an emergency declaration. The State and local agencies referred to here are the Arkansas Department of Health, Arkansas Department of Emergency Management, Pope, Yell, Johnson, and Logan County offsite agencies.

The third EAL addresses a total loss of the communications methods used to notify the NRC of

an emergency declaration.

This EAL is the cold condition equivalent of the hot condition EAL SU7.1.

Reference(s):

1. OP-1903.062 Communications System Operating Procedure 2. NEI 99-01 CU5 to 0CAN091801

Page 101 of 240 Attachment 1 - Emergency Action Level Technical Bases

Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 6 - Hazardous Event Affecting Safety Systems

Initiating Condition: Hazardous event affecting SAFETY SYSTEMS needed for the current operating mode

EAL: CA6.1 Alert The occurrence of any Table 1[2]C-6 hazardous event AND Event damage has caused indications of degraded performance on one train of a SAFETY SYSTEM needed for the current operating mode AND EITHER:

Event damage has caused indications of degraded performance to the second train of the SAFETY SYSTEM needed for the current operating mode Event damage has resulted in VISIBLE DAMAGE to the second train of the SAFETY SYSTEM needed for the current operating mode (Notes 10, 11)

Note 10: If the affected SAFETY SYSTEM train was already inoperable or out of service before the hazardous event occurred, then emergency classification is not warranted.

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

Table 1[2]C-6 Hazardous Events Seismic event (earthquake) Internal or external FLOODING event High winds or tornado strike FIRE EXPLOSION Other events with similar hazard characteristics as determined by the Shift Manager to 0CAN091801

Page 102 of 240 Attachment 1 - Emergency Action Level Technical Bases

Mode Applicability:

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

EXPLOSION - A rapid, violent and catastrophic failure of a piece of equipment due to combustion, chemical reaction or overpressurization. A release of steam (from high energy lines or components) or an electrical component failure (caused by short circuits, grounding, arcing, etc.) should not automatically be considered an explosion. Such events require a post-event inspection to determine if the attributes of an explosion are present.

FIRE - Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred but is not required if large quantities of smoke and heat are observed.

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 10 CFR 50.2).

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

(1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.

VISIBLE DAMAGE - Damage to a SAFETY SYSTEM train that is readily observable without measurements, testing, or analysis. The visual impact of the damage is sufficient to cause concern regarding the operability or reliability of the affected SAFETY SYSTEM train.

Basis: This IC addresses a hazardous event that causes damage to SAFETY SYSTEMS needed for the current operating mode. In order to provide the appropriate context for consideration of an ALERT classification, the hazardous event must have caused indications of degraded SAFETY SYSTEM performance in one train, and there must be either indications of performance issues with the second SAFETY SYSTEM train or VISIBLE DAMAGE to the second train such that the potential exists for this second SAFETY SYSTEM train to have performance issues. In other words, in order for this EAL to be classified, the hazardous event must occur, at least one SAFETY SYSTEM train must have indications of degraded performance, and the second SAFETY SYSTEM train must have indications of degraded performance or VISIBLE DAMAGE such that the potential exists for performance issues. Note that this second SAFETY SYSTEM to 0CAN091801

Page 103 of 240 Attachment 1 - Emergency Action Level Technical Bases

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

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

VISIBLE DAMAGE addresses damage to a SAFETY SYSTEM train that is not in service/operation and that potentially could cause performance issues. 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. This VISIBLE DAMAGE should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train.

Escalation of the emergency classification level would be via IC CS1 or AS1.

This EAL is the cold condition equivalent of the hot condition EAL SA9.1.

Reference(s):

1. EP FAQ 2016-002 2. NEI 99-01 CA6

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Category E - Independent Spent Fuel Storage Installation (ISFSI)

EAL Group: ANY (EALs in this category are applicable to any plant condition, hot or cold.)

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

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

The ANO ISFSI is located wholly within the plant PROTECTED AREA. Therefore any security event related to the ISFSI is classified under Category H1 security event related EALs.

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Category: E - ISFSI Subcategory: Confinement Boundary

Initiating Condition: Damage to a loaded cask CONFINEMENT BOUNDARY EAL: EU1.1 Unusual Event Damage to a loaded cask CONFINEMENT BOUNDARY as indicated by an on-contact radiation reading on the surface of a loaded spent fuel cask (VSC-24 VCC or HI-STORM

overpack) > any Table 1[2]E-1 value Table 1[2]E-1 ISFSI Dose Rates VSC-24 VCC HI-STORM 200 mrem/hr on the sides 400 mrem/hr on the top 700 mrem/hr at the air inlet 200 mrem/hr at the air outlet 60 mrem/hr (gamma + neutron) on the top or outlet vent 600 mrem/hr (gamma + neutron) on the side of the overpack (excluding inlet and outlet ducts)

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 ANO ISFSI, the Confinement Boundary is comprised of either the Multi-assembly Sealed Basket (MSB) (SNC System) or Multi-Purpose Canister (MPC) (Holtec System).

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.

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 to 0CAN091801

Page 106 of 240 Attachment 1 - Emergency Action Level Technical Bases

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 specified EAL threshold values correspond to 2 times the cask technical specification values (ref. 1, 2). The technical specification (licensing bases document) multiple of "2 times", which is also used in Recognition Category A IC AU1, is used here to distinguish between non-emergency and emergency conditions. The emphasis for this classification is the degradation in the level of safety of the spent fuel cask and not the magnitude of the associated dose or dose rate. It is recognized that in the case of extreme damage to a loaded cask, the fact that the "on-contact" dose rate limit is

exceeded may be determined based on measurement of a dose rate at some distance from the cask.

Security-related events for ISFSIs are covered under ICs HU1 and HA1.

Reference(s):

1. Certificate of Compliance Appendix A Technical Specifications for the HI-STORM 100 Cask System Section 5.7.4 2. VSC-24 Storage Cask Final Safety Analysis Report Section 1.2.4 Maximum External Surface Dose Rate 3. NEI 99-01 E-HU1 to 0CAN091801

Page 107 of 240 Attachment 1 - Emergency Action Level Technical Bases

Category F - Fission Product Barrier Degradation EAL Group: Hot Conditions (RCS temperature > 200°F); EALs in this category are applicable only in one or more hot operating modes.

EALs in this category represent threats to the defense in depth design concept that precludes the release of highly radioactive fission products to the environment. This concept relies on multiple physical barriers any one of which, if maintained intact, precludes the release of significant amounts of radioactive fission products to the environment. The primary fission product barriers are:

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

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

C. Containment Barrier (CNB): The Containment Barrier includes the Reactor Building and connections up to and including the outermost containment isolation valves. This barrier also includes the main steam, feedwater, and blowdown line extensions outside the Reactor Building up to and including the outermost secondary side isolation valve. Containment Barrier thresholds are used as criteria for escalation of the Emergency Classification Level (ECL) from an Alert to a Site Area Emergency or a General Emergency.

The EALs in this category require evaluation of the loss and potential loss thresholds listed in the fission product barrier matrix of Table F-1. "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 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 third barrier to 0CAN091801

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The logic used for emergency classification based on fission product barrier monitoring should reflect the following considerations:

The Fuel Clad Barrier and the RCS Barrier are weighted more heavily than the Containment Barrier. Unusual Event ICs associated with RCS and Fuel Clad Barriers are addressed under System Malfunction ICs. For accident conditions involving a radiological release, evaluation of the fission product barrier thresholds will need to be performed in conjunction with dose assessments to ensure correct and timely escalation of the emergency classification. For example, an evaluation of the fission product barrier thresholds may result in a Site Area Emergency classification while a dose assessment may indicate that an EAL for General Emergency IC AG1 has been exceeded. The fission product barrier thresholds specified within a scheme reflect plant-specific ANO 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 containment, an interfacing system, or outside of the containment. The release of liquid or steam mass from the RCS due to the as-designed/expected operation of a relief valve is not considered to be RCS leakage. At the Site Area Emergency level, EAL users should maintain cognizance of how far present conditions are from meeting a threshold that would require a General Emergency declaration. For example, if the Fuel Clad and RCS fission product barriers were both lost, then there should be frequent assessments of containment radioactive inventory and integrity. Alternatively, if both the Fuel Clad and RCS fission product barriers were potentially lost, the Emergency Director would have more assurance that there was no immediate need to escalate to a General Emergency.

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Category: F - Fission Product Barrier Degradation Subcategory: N/A Initiating Condition:

Any loss or any potential loss of either Fuel Clad or RCS EAL: FA1.1 Aler t Any loss or any potential loss of either Fuel Clad or RCS barrier (Table 1[2]F-1)

Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown

Definition(s):

None Basis: Fuel Clad, RCS and Containment comprise the fission product barriers. Table 1[2]F-1 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.

Reference(s):

1. NEI 99-01 FA1 to 0CAN091801

Page 110 of 240 Attachment 1 - Emergency Action Level Technical Bases

Category: F - Fission Product Barrier Degradation Subcategory: N/A Initiating Condition: Loss or potential loss of any two barriers EAL: FS1.1 Site Area Emergency Loss or potential loss of any two barriers (Table 1[2]F-1)

Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown

Definition(s):

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

Basis: Fuel Clad, RCS and Containment comprise the fission product barriers. Table 1[2]F-1 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 would have greater assurance that escalation to a General Emergency is less IMMINENT.

Reference(s):

1. NEI 99-01 FS1

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Category: F - Fission Product Barrier Degradation Subcategory: N/A Initiating Condition: Loss of any two barriers and loss or potential loss of third barrier EAL: FG1.1 General Emergency Loss of any two barriers AND Loss or potential loss of the third barrier (Table 1[2]F-1)

Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown

Definition(s):

None Basis:

Fuel Clad, RCS and Containment comprise the fission product barriers. Table 1[2]F-1 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 Reference(s):

1. NEI 99-01 FG1 to 0CAN091801

Page 112 of 240 Attachment 1 - Emergency Action Level Technical Bases

Table 1[2]F-1 Fission Product Barrier Threshold Matrix & Bases

Table 1[2]F-1 lists the threshold conditions that define the Loss and Potential Loss of the three fission product barriers (Fuel Clad, Reactor Coolant System, and Containment). The table is structured so that each of the three barriers occupies adjacent columns. Each fission product barrier column is further divided into two columns; one for Loss thresholds and one for Potential Loss thresholds.

The first column of the table (to the left of the Fuel Clad Loss column) lists the categories (types) of fission product barrier thresholds. The fission product barrier categories are:

A. RCS or S/G Tube Leakage B. Inadequate Heat removal C. Containment Radiation / RCS Activity D. Containment Integrity or Bypass E. Emergency Director Judgment Each category occupies a row in Table 1[2]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 barrier column beginning with number one (ex., FCB1, FCB2-FCB9).

If a cell in Table 1[2]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 1[2]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 1[2]F-1, locates the likely category and then reads across the fission product barrier Loss and Potential Loss thresholds in that category to determine if a threshold has been exceeded. If a threshold has not been exceeded, the EAL-user proceeds to the next likely category and continues review of the thresholds in the new

category.

If the EAL-user determines that any threshold has been exceeded, by definition, the barrier is lost or potentially lost - even if multiple thresholds in the same barrier column are exceeded, only that one barrier is lost or potentially lost. The EAL-user must examine each of the three fission product barriers to determine if other barrier thresholds in the category are lost or potentially lost. For example, if containment radiation is sufficiently high, a Loss of the Fuel to 0CAN091801

Page 113 of 240 Attachment 1 - Emergency Action Level Technical Bases

Clad 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,-, E.

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Table 1[2]F-1 Fission Product Barrier Threshold Matrix Fuel Clad Barrier (FCB)

Reactor Coolant System Barrier (RCB)

Containment Barrier (CNB)

Category Loss Potential Loss Loss Potential Loss Loss Potential Loss A RCS or S/G Tube Leakage None FCB1 RVLMS Levels 1 through 9 [1 through 7] indicate DRY RCB1 An automatic or manual ESAS [ESFAS] actuation required by EITHER: UNISOLABLE RCS leakage S/G tube RUPTURE RCB2 UNISOLABLE RCS leakage or S/G tube leakage > 50[44] gpm excluding normal reductions in RCS inventory (e.g., letdown, RCP seal leakoff)

RCB3 Unit 1: PTS limits apply (RT14) AND RCS pressure and temperature are left of the NDTT/LTOP limit lines on EOP Figure 3 (Note 12) Unit 2: Uncontrolled RCS cooldown (> 50 °F step change or > 100 °F change in less than a one-hour period) AND RCS pressure and temperature are to the left of line B (200 degrees MTS),

Standard Attachment 1, P-T Limits (Note 12) CNB1 A S/G that is leaking > 50[44] gpm (excluding normal reductions in RCS inventory) or that is RUPTURED is also FAULTED outside of containment None B Inadequate Heat Removal FCB2 CETs > 1200°F FCB3 CETs > 700°F FCB4 RCS heat removal cannot be established using steam generators AND An on-shift SRO has determined that the procedure conditions are met to commence initiation of HPI[Once Through] cooling None RCB4 RCS heat removal cannot be established using steam generators AND An on-shift SRO has determined that the procedure conditions are met to commence initiation of HPI[Once Through] cooling None CNB2 CETs > 1200°F AND Restoration procedures not effective within 15 min. (Note 1)

C CTMT Radiation /

RCS Activity FCB5 Containment High Range Radiation Monitor RE-8060/8061

[2RE-8925-1/ 8925-2]

> 750 [700] R/hr FCB6 Coolant activity > 300 Ci/gm dose equivalent I-131 None RCB5 Containment High Range Radiation Monitor RE-8060/8061 [2RE-8925-1/ 8925-2] > 40[50] R/hr None None CNB3 Containment High Range Radiation Monitor RE-8060/8061 [2RE-8925-1/ 8925-2] > 10,000[12,000] R/hr to 0CAN091801

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Table 1[2]F-1 Fission Product Barrier Threshold Matrix Fuel Clad Barrier (FCB)

Reactor Coolant System Barrier (RCB)

Containment Barrier (CNB)

Category Loss Potential Loss Loss Potential Loss Loss Potential Loss D CTMT Integrity or Bypass None None None None CNB4 Containment isolation is required AND EITHER: Containment integrity has been lost based on Emergency Director judgment UNISOLABLE pathway from Containment to the environment exists CNB5 Indications of RCS leakage outside of Containment CNB6 Containment pressure > 73.7 psia CNB7 Containment hydrogen concentration

> 4% CNB8 Containment pressure > 44.7 psia [23.3 psia] with < one full train of containment heat removal systems (Note 9) operating per design for 15 min. (Note 1)

E Emergency Director Judgment FCB7 Any condition in the opinion of the Emergency Director that indicates loss of the Fuel Clad barrier FCB8 Any condition in the opinion of the Emergency Director that indicates potential loss of the Fuel Clad barrier RCB6 Any condition in the opinion of the Emergency Director that indicates loss of the RCS barrier RCB7 Any condition in the opinion of the Emergency Director that indicates potential loss of the RCS barrier CNB9 Any condition in the opinion of the Emergency Director that indicates loss of the Containment barrier CNB10 Any condition in the opinion of the Emergency Director that indicates potential loss of the Containment barrier to 0CAN091801

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Barrier: Fuel Clad Category: A - RCS or S/G Tube Leakage

Degradation Threat: Loss Threshold:

None

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Barrier: Fuel Clad Category: A - RCS or S/G Tube Leakage

Degradation Threat: Potential Loss Threshold:

FCB1 RVLMS Levels 1 through 9[1 through 7] indicate DRY Definition(s):

None Basis: This reading indicates a reduction in reactor vessel water level sufficient to allow the onset of heat-induced cladding damage.

There is no Fuel Clad Barrier Loss threshold associated with RCS or S/G Tube Leakage.

Reference(s):

1. ULD-1-SYS-24 Unit 1 Inadequate Core Cooling System 2. Calculation 84-EQ-0080-02 Loop Error Analysis for Reactor Vessel Level Monitoring System 3. ULD-2-SYS-24 Unit 2 Inadequate Core Cooling Monitoring System
4. Calculation 90-E-0116-01 Unit 2 EOP Setpoint Document, Setpoint R.3 5. NEI 99-01 RCS or SG Tube Leakage Potential Loss 1.A

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Barrier: Fuel Clad Category: B - Inadequate Heat Removal

Degradation Threat: Loss Threshold:

FCB2 CETs > 1200°F Definition(s):

None Basis: This reading indicates temperatures within the core are sufficient to cause significant superheating of reactor coolant.

Reference(s):

1. NEI 99-01 Inadequate Heat Removal Loss 2.A to 0CAN091801

Page 119 of 240 Attachment 1 - Emergency Action Level Technical Bases

Barrier: Fuel Clad Category: B - Inadequate Heat Removal

Degradation Threat: Potential Loss Threshold:

FCB3 CETs > 700°F Definition(s):

None Basis: This reading indicates a reduction in reactor vessel water level sufficient to allow the onset of heat-induced cladding damage.

Reference(s):

1. NEI 99-01 Inadequate Heat Removal Potential Loss 2.A to 0CAN091801

Page 120 of 240 Attachment 1 - Emergency Action Level Technical Bases

Barrier: Fuel Clad Category: B - Inadequate Heat Removal

Degradation Threat: Potential Loss Threshold:

FCB4 RCS heat removal cannot be established using steam generators AND An on-shift SRO has determined that the procedure conditions are met to commence initiation of HPI[Once Through] cooling Definition(s):

None

Basis: This condition indicates an extreme challenge to the ability to remove RCS heat using the steam generators (i.e., loss of an effective secondary-side heat sink). This condition represents a potential loss of the Fuel Clad Barrier. In accordance with EOPs, there may be unusual accident conditions during which operators intentionally reduce the heat removal capability of the steam generators; during these conditions, classification using this threshold is not

warranted.

In combination with Potential Loss RCB4, meeting this threshold results in a Site Area Emergency.

Reference(s):

1. OP-1202.004 Overheating 2. OP-1202.013 Figure 4, Core Exit Thermocouple for Inadequate Core Cooling 3. OP-2202.006 Loss of Feedwater 4. OP-2202.009 Functional Recovery, Safety Function Status Check 5
5. NEI 99-01 Inadequate Heat Removal Potential Loss 2.B

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Barrier: Fuel Clad Category: C - CTMT Radiation / RCS Activity

Degradation Threat: Loss Threshold:

FCB5 Containment High Range Radiation Monitor RE-8060/8061[2RE-8925-1/8925-2]

> 750[700] R/hr Definition(s):

None Basis:

The containment radiation monitor reading (768[682] R/hr rounded to 750[700] R/hr for readability) corresponds to an instantaneous release of all reactor coolant mass into the containment, assuming that reactor coolant activity equals 300 µCi/gm dose equivalent I-131.

Reactor coolant activity above this level is greater than that expected for iodine spikes and corresponds to 1.49[1.13]% 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 RCB5 since it indicates a loss of both the Fuel Clad Barrier and the RCS Barrier.

Note that a combination of the two monitor readings appropriately escalates the ECL to a Site Area Emergency.

There is no Potential Loss threshold associated with CTMT Radiation/RCS Activity.

Reference(s):

1. EP-CALC-ANO-1702 Containment High Range Radiation Monitor EAL Values 2. NEI 99-01 RCS Activity/Containment Radiation FC Loss 3.A

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Barrier: Fuel Clad Category: C - CTMT Radiation / RCS Activity

Degradation Threat: Loss Threshold:

FCB6 Coolant activity > 300 µCi/gm dose equivalent I-131 Definition(s):

None Basis: This threshold indicates that RCS radioactivity concentration is greater than 300 µCi/gm dose equivalent I-131. Reactor coolant activity above this level is greater than that expected for iodine spikes and corresponds to an approximate range of 2% to 5% fuel clad damage. Since this condition indicates that a significant amount of fuel clad damage has occurred, it represents a loss of the Fuel Clad Barrier.

It is recognized that sample collection and analysis of reactor coolant with highly elevated activity levels could require several hours to complete. Nonetheless, a sample-related threshold is included as a backup to other indications.

There is no Potential Loss threshold associated with CTMT Radiation/RCS Activity.

Reference(s):

1. NEI 99-01 RCS Activity/Containment Radiation Fuel Clad Loss 3.B

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Barrier: Fuel Clad Category: C - CTMT Radiation / RCS Activity

Degradation Threat: Potential Loss Threshold:

None

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Barrier: Fuel Clad Category: D - CTMT Integrity or Bypass

Degradation Threat: Loss Threshold:

None

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Barrier: Fuel Clad Category: D - CTMT Integrity or Bypass

Degradation Threat: Potential Loss Threshold:

None

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Barrier: Fuel Clad Category: E - Emergency Director Judgment

Degradation Threat: Loss Threshold:

FCB7 Any condition in the opinion of the Emergency Director that indicates loss of the Fuel Clad barrier Definition(s):

None Basis:

This threshold addresses any other factors that are to be used by the Emergency Director in determining whether the Fuel Clad barrier is lost.

Reference(s):

1. NEI 99-01 Emergency Director Judgment Fuel Clad Loss 6.A

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Barrier: Fuel Clad Category: E - Emergency Director Judgment

Degradation Threat: Potential Loss Threshold:

FCB8 Any condition in the opinion of the Emergency Director that indicates potential loss of the Fuel Clad barrier Definition(s):

None Basis:

This threshold addresses any other factors that are to be used by the Emergency Director in determining whether the Fuel Clad barrier is potentially lost. The Emergency Director should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored.

Reference(s):

1. NEI 99-01 Emergency Director Judgment Potential Fuel Clad Loss 6.A

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Barrier: Reactor Coolant System Category: A - RCS or S/G Tube Leakage

Degradation Threat: Loss Threshold:

RCB1 An automatic or manual ESAS[ESFAS] actuation required by EITHER: UNISOLABLE RCS leakage S/G tube RUPTURE Definition(s):

FAULTED - The term applied to a steam generator that has a steam leak on the secondary side of sufficient size to cause an uncontrolled drop in steam generator pressure or the steam generator to become completely depressurized.

RUPTURED - The condition of a steam generator in which primary-to-secondary leakage is of sufficient magnitude to require a safety injection (automatic or manual).

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

Basis: Failure to isolate the leak (from the Control Room or locally), within 15 minutes or if known that the leak cannot be isolated within 15 minutes, from the start of the leak requires immediate classification.

This threshold is based on an UNISOLABLE RCS leak of sufficient size to require an automatic or manual actuation of the Emergency Core Cooling System (ECCS). This condition clearly represents a loss of the RCS Barrier.

This threshold is applicable to unidentified and pressure boundary leakage, as well as identified leakage. It is also applicable to UNISOLABLE RCS leakage through an interfacing system.

The mass loss may be into any location - inside containment, to the secondary-side (i.e., steam generator tube leakage) or outside of containment.

A steam generator with primary-to-secondary leakage of sufficient magnitude to require a safety injection is considered to be RUPTURED. If a RUPTURED steam generator is also FAULTED outside of containment, the declaration escalates to a Site Area Emergency since the Containment Barrier Loss threshold CNB1 will also be met.

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Reference(s):

1. OP-1202.010 ESAS 2. OP-2202.003 Loss of Coolant Accident
3. NEI 99-01 RCS or SG Tube Leakage Reactor Coolant System Loss 1.A

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Barrier: Reactor Coolant System Category: A - RCS or S/G Tube Leakage

Degradation Threat: Potential Loss Threshold:

RCB2 UNISOLABLE RCS leakage or S/G tube leakage > 50[44] gpm excluding normal reductions in RCS inventory (e.g., letdown, RCP seal leakoff)

Definition(s):

FAULTED - The term applied to a steam generator that has a steam leak on the secondary side of sufficient size to cause an uncontrolled drop in steam generator pressure or the steam generator to become completely depressurized.

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

Basis: Failure to isolate the leak (from the Control Room or locally), within 15 minutes or if known that the leak cannot be isolated within 15 minutes, from the start of the leak requires immediate classification.

This threshold is based on an UNISOLABLE RCS leak that results in the inability to maintain pressurizer level within specified limits by operation of a normally used makeup [charging]

pump, but an ESAS [ESFAS] actuation has not occurred.

This threshold is applicable to unidentified and pressure boundary leakage, as well as identified leakage. It is also applicable to UNISOLABLE RCS leakage through an interfacing system.

The mass loss may be into any location - inside containment, to the secondary-side (i.e., steam generator tube leakage) or outside of containment.

If a leaking steam generator is also FAULTED outside of containment, the declaration escalates to a Site Area Emergency since the Containment Barrier Loss threshold CNB1 will also be met.

Reference(s):

1. 1SAR 9.1 Makeup and Purification System 2. 2SAR 9.3.4 Chemical and Volume Control System
3. NEI 99-01 RCS or SG Tube Leakage Reactor Coolant System Potential Loss 1.A

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Barrier: Reactor Coolant System Category: A - RCS or S/G Tube Leakage

Degradation Threat: Potential Loss Threshold:

RCB3 Unit 1: PTS limits apply (RT14) AND RCS pressure and temperature are left of the NDTT/LTOP limit lines, on EOP Figure 3 (Note 12)

Unit 2: Uncontrolled RCS cooldown (> 50 °F step change or > 100 °F change in less than a one-hour period)

AND RCS pressure and temperature are to the left of line B (200 degrees MTS), Standard , P-T Limits (Note 12)

Note 12: Once PTS limits are first invoked, if RCS temperature and pressure are not brought within the limits within 15 minutes, this threshold is met and an immediate declaration is warranted. This threshold is met immediately upon exceeding the limits after this initial 15 minute period until PTS limits no longer apply.

Definition(s):

None Basis: This condition indicates an extreme challenge to the integrity of the RCS pressure boundary due to pressurized thermal shock - a transient that causes rapid RCS cooldown while the RCS is in Mode 3 or higher (i.e., hot and pressurized).

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Reference(s):

1. OP-1202.012 Repetitive Task 14 Control RCS Pressure 2. OP-1202.013 EOP Figures, Figure 3 RCS Pressure vs Temperature Limits
3. OP-1202.011 HPI Cooldown
4. Calculation No: 90-E-0116-01 ANO- EOP Setpoint Basis Document OP Setpoint P.2, RCS Pressure-Temperature 5. OP-2202.010 Standard Attachments, Attachment 1, P-T Limits 6. NEI 99-01 RCS or SG Tube Leakage Reactor Coolant System Potential Loss 1.B

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Barrier: Reactor Coolant System Category: B - Inadequate Heat Removal

Degradation Threat: Loss Threshold:

None

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Barrier: Reactor Coolant System Category: B - Inadequate Heat Removal

Degradation Threat: Potential Loss Threshold:

RCB4 RCS heat removal cannot be established using steam generators AND An on-shift SRO has determined that the procedure conditions are met to commence initiation of HPI[Once Through] cooling Definition(s):

None

Basis: This condition indicates an extreme challenge to the ability to remove RCS heat using the steam generators (i.e., loss of an effective secondary-side heat sink). This condition represents a potential loss of the RCS Barrier. In accordance with EOPs, there may be unusual accident conditions during which operators intentionally reduce the heat removal capability of the steam generators; during these conditions, classification using this threshold is not warranted.

In combination with Potential Loss FCB4, meeting this threshold results in a Site Area Emergency.

This condition warrants a Site Area Emergency declaration because inadequate RCS heat removal may result in fuel heat-up sufficient to damage the cladding and raise RCS pressure to the point where mass will be lost from the system.

There is no RCS barrier Loss threshold associated with Inadequate Heat Removal.

Reference(s):

1. OP-1202.004 Overheating 2. OP-1202.013 Figure 4, Core Exit Thermocouple for Inadequate Core Cooling 3. OP-2202.006 Loss of Feedwater 4. OP-2202.009 Functional Recovery, Safety Function Status Check 5
5. NEI 99-01 Inadequate Heat Removal RCS Potential Loss 2.B to 0CAN091801

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Barrier: Reactor Coolant System Category: C - CTMT Radiation/ RCS Activity

Degradation Threat: Loss Threshold:

RCB5 Containment High Range Radiation Monitor RE-8060/8061[2RE-8925-1/8925-2] > 40[50] R/hr Definition(s):

None Basis: NRC Information Notice 97-045, Supplement 1, identifies the potential for erratic indications from the high range radiation monitors (HRRMs) as a result of thermally induced currents (TIC) which may cause the HRRM to read falsely high (for approximately 15 minutes) on a rapid temperature rise, and fail low intermittently on a rapid temperature fall. Because of this phenomenon, any trends or alarms on the HRRM's should be validated by comparison to the containment low range/area radiation monitors and Air Monitoring Systems trends before actions are taken.

The containment radiation monitor reading (42.8[50.4] R/hr rounded to 40[50] R/hr for readability) corresponds to an instantaneous release of all reactor coolant mass into the containment, assuming that reactor coolant activity equals Technical Specification allowable limits. This value is lower than that specified for Fuel Clad Barrier Loss threshold FCB5 since it indicates a loss of the RCS Barrier only.

There is no Potential Loss threshold associated with CTMT Radiation/RCS Activity.

Reference(s):

1. EP-CALC-ANO-1702 Containment High Range Radiation Monitor EAL Values 2. NEI 99-01 CMT Radiation / RCS Activity RCS Loss 3.A

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Barrier: Reactor Coolant System Category: B - CTMT Radiation/ RCS Activity

Degradation Threat: Potential Loss Threshold:

None

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Barrier: Reactor Coolant System Category: D - CTMT Integrity or Bypass

Degradation Threat: Loss Threshold:

None

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Page 138 of 240 Attachment 1 - Emergency Action Level Technical Bases

Barrier: Reactor Coolant System Category: D - CTMT Integrity or Bypass

Degradation Threat: Potential Loss Threshold:

None

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Barrier: Reactor Coolant System Category: E - Emergency Director Judgment

Degradation Threat: Loss Threshold:

RCB6 Any condition in the opinion of the Emergency Director that indicates loss of the RCS barrier Definition(s):

None Basis: This threshold addresses any other factors that may be used by the Emergency Director in determining whether the RCS Barrier is lost.

Reference(s):

1. NEI 99-01 Emergency Director Judgment RCS Loss 6.A to 0CAN091801

Page 140 of 240 Attachment 1 - Emergency Action Level Technical Bases

Barrier: Reactor Coolant System Category: E - Emergency Director Judgment

Degradation Threat: Potential Loss Threshold:

RCB7 Any condition in the opinion of the Emergency Director that indicates potential loss of the RCS barrier Definition(s):

None Basis:

This threshold addresses any other factors that may be used by the Emergency Director in determining whether the RCS Barrier is potentially lost. The Emergency Director should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored.

Reference(s):

1. NEI 99-01 Emergency Director Judgment RCS Potential Loss 6.A

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Barrier: Containment Category: A - RCS or S/G Tube Leakage

Degradation Threat: Loss Threshold:

CNB1 A S/G that is leaking > 50[44] gpm (excluding nor mal reductions in RCS inventory) or that is RUPTURED is also FAULTED outside of containment Definition(s):

FAULTED - The term applied to a steam generator that has a steam leak on the secondary side of sufficient size to cause an uncontrolled drop in steam generator pressure or the steam generator to become completely depressurized.

RUPTURED - The condition of a steam generator in which primary-to-secondary leakage is of sufficient magnitude to require a safety injection (automatic or manual).

Basis: This threshold addresses a leaking or RUPTURED Steam Generator (SG) that is also FAULTED outside of containment. The condition of the SG, whether leaking or RUPTURED, is determined in accordance with the thresholds for RCS Barrier Potential Loss RCB2 and Loss RCB1, respectively. This condition represents a bypass of the containment barrier.

FAULTED is a defined term within the NEI 99-01 methodology; this determination is not necessarily dependent upon entry into, or diagnostic steps within, an EOP. For example, if the pressure in a steam generator is dropping uncontrollably (part of the FAULTED definition) and the FAULTED steam generator isolation procedure is not entered because EOP user rules are dictating implementation of another procedure to address a higher priority condition, the steam generator is still considered FAULTED for emergency classification purposes.

The FAULTED criterion establishes an appropriate lower bound on the size of a steam release that may require an emergency classification. Steam releases of this size are readily observable with normal Control Room indications. The lower bound for this aspect of the containment barrier is analogous to the lower bound criteria specified in IC SU4 for the fuel clad barrier (i.e., RCS activity values) and IC SU5 for the RCS barrier (i.e., RCS leak rate values).

This threshold also applies to prolonged steam releases necessitated by operational considerations such as the forced steaming of a leaking or RUPTURED steam generator directly to atmosphere to cooldown the plant. These types of conditions will result in a significant and sustained release of radioactive steam to the environment (and are thus similar to a FAULTED condition). The inability to isolate the steam flow without an adverse effect on plant cooldown meets the intent of a loss of containment. to 0CAN091801

Page 142 of 240 Attachment 1 - Emergency Action Level Technical Bases

Steam releases associated with the expected operation of a SG power operated relief valve or safety relief valve do not meet the intent of this threshold. Such releases may occur intermittently for a short period of time following a reactor trip as operators process through emergency operating procedures to bring the plant to a stable condition and prepare to initiate a plant cooldown. Steam releases associated with the unexpected operation of a valve (e.g., a stuck-open safety valve) do meet this threshold.

Following a SG tube leak or rupture, there may be minor radiological releases through a

secondary-side system component (e.g., air ejectors, glad seal exhausters, valve packing, etc.). These types of releases do not constitute a loss or potential loss of containment but should be evaluated using the Recognition Category A ICs.

The ECLs resulting from primary-to-secondary leakage, with or without a steam release from the FAULTED SG, are summarized below.

Affected SG is FAULTED Outside of Containment?

P-to-S Leak Rate Yes No Less than or equal to 25 gpm No classification No classification Greater than 25 gpm Unusual Event per SU5.1Unusual Event per SU5.1Greater than 50[44] gpm (RCS Barrier Potential Loss

) Site Area Emergency per FS1.1 Alert per FA1.1 Requires an automatic or manual ESAS[ESFAS] actuation (RCS Barrier Loss

) Site Area Emergency per FS1.1 Alert per FA1.1 There is no Potential Loss threshold associated with RCS or S/G Tube Leakage.

Reference(s):

1. NEI 99-01 RCS or SG Tube Leakage Containment Loss 1.A

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Barrier: Containment Category: A - RCS or S/G Tube Leakage

Degradation Threat: Potential Loss Threshold:

None

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Barrier: Containment Category: B - Inadequate Heat Removal

Degradation Threat: Loss Threshold:

None

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Barrier: Containment Category: B - Inadequate Heat Removal

Degradation Threat: Potential Loss Threshold:

CNB2 CETs > 1200°F AND Restoration procedures not effective within 15 min. (Note 1)

Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded.

Definition(s):

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

Basis: The restoration procedure is considered "effective" if core exit thermocouple readings are dropping and/or if reactor vessel level is rising.

Whether or not the procedure(s) will be effective should be apparent within 15 minutes. The Emergency Director should escalate the emergency classification level as soon as it is determined that the procedure(s) will not be effective.

This condition represents an IMMINENT core melt sequence which, if not corrected, could lead to vessel failure and an increased potential for containment failure. For this condition to occur, there must already have been a loss of the RCS Barrier and the Fuel Clad Barrier. If implementation of a procedure(s) to restore adequate core cooling is not effective (successful) within 15 minutes, it is assumed that the event trajectory will likely lead to core melting and a subsequent challenge of the Containment Barrier.

Severe accident analyses (e.g., NUREG-1150) have concluded that function restoration procedures can arrest core degradation in a significant fraction of core damage scenarios, and that the likelihood of containment failure is very small in these events. Given this, it is appropriate to provide 15 minutes beyond the required entry point to determine if procedural actions can reverse the core melt sequence.

Reference(s):

1. NEI 99-01 Inadequate Heat Removal Containment Potential Loss 2.A to 0CAN091801

Page 146 of 240 Attachment 1 - Emergency Action Level Technical Bases

Barrier: Containment Category: C - CTMT Radiation/RCS Activity

Degradation Threat: Loss Threshold:

None

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Barrier: Containment Category: C - CTMT Radiation/RCS Activity

Degradation Threat: Potential Loss Threshold:

CNB3 Containment High Range Radiation Monitor RE-8060/8061[2RE-8925-1/8925-2]

> 10,000[12,000] R/hr Definition(s):

None Basis:

The containment radiation monitor reading (10,300[12,100] R/hr rounded to 10,000[12,000] R/hr for readability) corresponds to an instantaneous release of all reactor coolant mass into the containment, assuming that 20% of the fuel cladding has failed. This level of fuel clad failure is well above that used to determine the analogous Fuel Clad Barrier Loss and RCS Barrier Loss thresholds.

NUREG-1228, Source Estimations During Incident Response to Severe Nuclear Power Plant Accidents, indicates the fuel clad failure must be greater than approximately 20% in order for there to be a major release of radioactivity requiring offsite protective actions. For this condition to exist, there must already have been a loss of the RCS Barrier and the Fuel Clad Barrier. It is therefore prudent to treat this condition as a potential loss of containment which would then escalate the ECL to a General Emergency.

There is no Loss threshold associated with CTMT Radiation/RCS Activity.

Reference(s):

1. EP-CALC-ANO-1702 Containment High Range Radiation Monitor EAL Values 2. NEI 99-01 CTMT Radiation / RCS Activity Containment Potential Loss 3.A

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Barrier: Containment Category: D - CTMT Integrity or Bypass

Degradation Threat: Loss Threshold:

CNB4 Containment isolation is required AND EITHER:

Containment integrity has been lost based on Emergency Director judgment UNISOLABLE pathway from Containment to the environment exists Definition(s):

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

Basis: Failure to isolate the leak (from the Control Room or locally), within 15 minutes or if known that the leak cannot be isolated within 15 minutes, from the start of the leak requires immediate classification.

The status of the containment barrier during an event involving steam generator tube leakage is assessed using Loss Threshold CNB1.

These thresholds address a situation where containment isolation is required and one of two conditions exists as discussed below. Users are reminded that there may be accident and release conditions that simultaneously meet both bulleted thresholds.

First Threshold - Containment integrity has been lost, i.e., the actual containment atmospheric leak rate likely exceeds that associated with allowable leakage (or sometimes referred to as design leakage). Following the release of RCS mass into containment, containment pressure will fluctuate based on a variety of factors; a loss of containment integrity condition may (or may not) be accompanied by a noticeable drop in containment pressure. Recognizing the inherent difficulties in determining a containment leak rate during accident conditions, it is expected that the Emergency Director will assess this threshold using judgment, and with due consideration given to current plant conditions, and available operational and radiological data (e.g., containment pressure, readings on radiation monitors outside containment, operating status of containment pressure control equipment, etc.).

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Refer to the middle piping run of Figure 1. Two simplified examples are provided. One is leakage from a penetration and the other is leakage from an in-service system valve. Depending upon radiation monitor locations and sensitivities, the leakage could be detected by any of the four monitors depicted in the figure.

Another example would be a loss or potential loss of the RCS barrier, and the simultaneous occurrence of two FAULTED locations on a steam generator where one fault is located inside containment (e.g., on a steam or feedwater line) and the other outside of containment. In this case, the associated steam line provides a pathway for the containment atmosphere to escape to an area outside the containment.

Following the leakage of RCS mass into containment and a rise in containment pressure, there may be minor radiological releases associated with allowable (design) containment leakage through various penetrations or system components. These releases do not constitute a loss or potential loss of containment but should be evaluated using the Recognition Category A ICs.

Second Threshold - Conditions are such that there is an UNISOLABLE pathway for the migration of radioactive material from the containment atmosphere to the environment. As used here, the term "environment" includes the atmosphere of a room or area, outside the containment, that may, in turn, communicate with the outside-the-plant atmosphere (e.g.,

through discharge of a ventilation system or atmospheric leakage). Depending upon a variety of factors, this condition may or may not be accompanied by a noticeable drop in containment pressure.

Refer to the top piping run of Figure 1. In this simplified example, the inboard and outboard isolation valves remained open after a containment isolation was required (i.e., containment isolation was not successful). There is now an UNISOLABLE pathway from the containment to

the environment.

The existence of a filter is not considered in the threshold assessment. Filters do not remove fission product noble gases. In addition, a filter could become ineffective due to iodine and/or particulate loading beyond design limits (i.e., retention ability has been exceeded) or water saturation from steam/high humidity in the release stream.

Leakage between two interfacing liquid systems, by itself, does not meet this threshold.

Refer to the bottom piping run of Figure 1. In this simplified example, leakage in an RCP seal cooler is allowing radioactive material to enter the Auxiliary Building. The radioactivity would be detected by the Process Monitor. If there is no leakage from the closed water cooling system to the Auxiliary Building, then no threshold has been met. If the pump developed a leak that allowed steam/water to enter the Auxiliary Building, then second threshold would be met. Depending upon radiation monitor locations and sensitivities, this leakage could be detected by any of the four monitors depicted in the figure and cause the first threshold to be met as well.

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Following the leakage of RCS mass into containment and a rise in containment pressure, there may be minor radiological releases associated with allowable containment leakage through various penetrations or system components. Minor releases may also occur if a containment isolation valve(s) fails to close but the containment atmosphere escapes to an enclosed system. These releases do not constitute a loss or potential loss of containment but should be evaluated using the Recognition Category A ICs.

Reference(s):

1. NEI 99-01 CTMT Integrity or Bypass Containment Loss 4.A

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Barrier: Containment Category: D - CTMT Integrity or Bypass

Degradation Threat: Loss Threshold:

CNB5 Indications of RCS leakage outside of Containment Definition(s):

None Basis: To ensure proper escalation of the emergency classification, the RCS leakage outside of containment must be related to the mass loss that is causing the RCS Loss RCB1 and/or Potential Loss RCB2 threshold to be met.

The status of the containment barrier during an event involving steam generator tube leakage is assessed using Containment Loss Threshold CNB1.

Containment sump, temperature, pressure and/or radiation levels will rise if reactor coolant mass is leaking into the containment. If these parameters have not risen, then the reactor coolant mass may be leaking outside of containment (i.e., a containment bypass sequence).

Rises in sump, temperature, pressure, flow and/or radiation level readings outside of the containment may indicate that the RCS mass is being lost outside of containment.

Unexpected elevated readings and alarms on radiation monitors with detectors outside

containment should be corroborated with other available indications to confirm that the source is a loss of RCS mass outside of containment. If the fuel clad barrier has not been lost, radiation monitor readings outside of containment may not rise significantly; however, other unexpected changes in sump levels, area temperatures or pressures, flow rates, etc. should be sufficient to determine if RCS mass is being lost outside of the containment.

Refer to the middle piping run of Figure 1. In this simplified example, a leak has occurred at a reducer on a pipe carrying reactor coolant in the Auxiliary Building. Depending upon radiation monitor locations and sensitivities, the leakage could be detected by any of the four monitors depicted in the figure and cause threshold CNB4 to be met as well.

Reference(s):

1. NEI 99-01 CTMT Integrity or Bypass Containment Loss 4.B to 0CAN091801

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Figure 1: Containment Integrity or Bypass Examples

Open valve Open valve Open valve Open valve Open valve Open valvePenetrationDamperDamper Interface leakage RCP Seal Cooling Inside Reactor Building Auxiliary Building 1 s t Threshold - Airborne 1 s t Threshold - Airborne release from penetration 2 n d Threshold - Airborne release from pathway 2 n d Threshold - RCS leakage outside AB Airborne Monitor Area Monitor Process Monitor Effluent MonitorClosed Cooling Pump to 0CAN091801

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Barrier: Containment Category: D - CTMT Integrity or Bypass

Degradation Threat: Potential Loss

Threshold:

CNB6 Containment pressure > 73.7 psia Definition(s):

None Basis:

If containment pressure exceeds the design pressure, there exists a potential to lose the Containment Barrier. To reach this level, there must be an inadequate core cooling condition for an extended period of time; therefore, the RCS and Fuel Clad barriers would already be lost.

Thus, this threshold is a discriminator between a Site Area Emergency and General Emergency since there is now a potential to lose the third barrier.

Reference(s):

1. 1SAR 1.4.43 Criterion 50 - Containment Design Basis 2. 2SAR Table 6.2-7 Principle Containment Design Parameters
3. NEI 99-01 CTMT Integrity or Bypass Containment Potential Loss 4.A to 0CAN091801

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Barrier: Containment Category: D - CTMT Integrity or Bypass

Degradation Threat: Potential Loss Threshold:

CNB7 Containment hydrogen concentration > 4%

Definition(s):

None Basis: 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). The 4% hydrogen concentration is generally considered the lower limit for hydrogen deflagrations. A hydrogen burn will raise containment pressure and could result in collateral equipment damage leading to a loss of containment integrity. It therefore represents a potential loss of the Containment Barrier.

Reference(s):

1. Unit 1 SAMG Figure III-1B 2. Unit 2 SAMG Phase 1 Instructions, Containment Flowchart
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Barrier: Containment Category: D - CTMT Integrity or Bypass

Degradation Threat: Potential Loss Threshold:

CNB8 Containment pressure > 44.7 psia[23.3 psia] with < one full train of containment heat removal systems (Note 9) operating per design for 15 min. (Note 1)

Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Note 9: One full train of containment heat removal systems consists of one train of RB [Containment] Spray and one train of RB [Containment] Cooling System.

Definition(s):

None Basis:

This threshold describes a condition where containment pressure is greater than the setpoint at which containment energy (heat) removal systems are designed to automatically actuate, and less than one full train of equipment is capable of operating per design. The 15-minute criterion is included to allow operators time to manually start equipment that may not have automatically started, if possible. This threshold represents a potential loss of containment in that containment heat removal/depressurization systems (e.g., containment sprays but not including containment venting strategies) are either lost or performing in a degraded manner.

Reference(s):

1. 1SAR 6.2 Reactor Building Spray System 2. 1SAR 6.3 Reactor Building Cooling System 3. OP-2202.003 Loss of Coolant Accident
4. OP-2202.010 Standard Attachments, Attachment 22
5. 2SAR 6.2.2 Containment Heat Removal Systems
6. 2SAR 7.3.1.1.11.2 Containment Spray System 7. NEI 99-01 CTMT Integrity or Bypass Containment Potential Loss 4.C to 0CAN091801

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Barrier: Containment Category: E - Emergency Director Judgment

Degradation Threat: Loss Threshold:

CNB9 Any condition in the opinion of the Emergency Director that indicates loss of the Containment barrier Definition(s):

None Basis:

This threshold addresses any other factors that may be used by the Emergency Director in determining whether the Containment Barrier is lost.

Reference(s):

1. NEI 99-01 Emergency Director Judgment Containment Loss 6.A

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Barrier: Containment Category: E - Emergency Director Judgment

Degradation Threat: Potential Loss Threshold:

CNB10 Any condition in the opinion of the Emergency Director that indicates potential loss of the Containment barrier Definition(s):

None Basis:

This threshold addresses any other factors that may be used by the Emergency Director in determining whether the Containment Barrier is potentially lost. The Emergency Director should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored.

Reference(s):

1. NEI 99-01 Emergency Director Judgment Containment Potential Loss 6.A

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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 Technological 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 plant 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 Events that are indicative of loss of Control Room habitability. If the Control Room must be evacuated, additional support for monitoring and controlling plant functions is necessary through the emergency response facilities.
7. Emergency Director Judgment The EALs defined in other categories specify the predetermined symptoms or events that are indicative of emergency or potential emergency conditions and thus warrant classification.

While these EALs have been developed to address the full spectrum of possible emergency conditions which may warrant classification and subsequent implementation of the Emergency Plan, a provision for classification of emergencies based on operator/management experience and judgment is still necessary. The EALs of this category provide the Emergency Director the latitude to classify emergency conditions consistent with the established classification criteria based upon Emergency Director judgment. to 0CAN091801

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Category: H - Hazards Subcategory: 1 - Security

Initiating Condition: Confirmed SECURITY CONDITION or threat EAL: HU1.1 Unusual Event A SECURITY CONDITION that does not involve a HOSTILE ACTION as reported by ANO 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):

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 ANO 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 ANO. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the SECURITY OWNER CONTROLLED AREA).

PROJECTILE - An object directed toward a Nuclear Power Plant that could cause concern for its continued operability, reliability, or personnel safety.

PROTECTED AREA - An area clearly demarcated by a fence or building wall with an entrance portal that is regulated by Security Personnel to control access.

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 10 CFR 50.2).

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

(1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.

SECURITY CONDITION - Any security event as listed in the approved security contingency plan that constitutes a threat/compromise to site security, threat/risk to site personnel, or a potential degradation to the level of safety of the plant. A security condition does not involve a HOSTILE ACTION.

SECURITY OWNER CONTROLLED AREA - The SOCA is the area demarcated as a Vehicle Barrier System (VBS) consisting of passive elements including a series of large concrete blocks on the inside of a delay fence with early warning capabilities. The SOCA is the area between the SOCA Fence and the PROTECTED AREA Boundary.

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 or 10 CFR 50.72. Security events assessed as HOSTILE ACTIONS are classifiable under ICs HA1 and HS1.

Timely and accurate communications between Security Shift Supervision and the Control Room is essential for proper classification of a security-related event. Classification of these events will initiate appropriate threat-related notifications to plant personnel and Offsite Response Organizations.

Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan [and Independent Spent Fuel Storage Installation Security Program]. The first threshold references the Security Shift Supervision because these are the individuals trained to confirm that a security event is occu rring or has occurred. Training on security event confirmation and classification is controlled due to the nature of Safeguards and 10 CFR 2.39

information.

The second threshold addresses the receipt of a credible security threat. The credibility of the threat is assessed in accordance with OP-1203.048 Security Event .

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The third threshold addresses the threat from the impact of an aircraft on the plant. The NRC Headquarters Operations Officer (HOO) will communicate to the licensee if the threat involves an aircraft. The status and size of the plane may also be provided by NORAD through the NRC. Validation of the threat is performed in accordance with 11-S-82-1 Security Contingency Events (ref. 2).

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 Security Plan for ANO (ref. 1).

Escalation of the emergency classification level would be via IC HA1.

Reference(s):

1. ANO Security Plan 2. OP-1203.048 Security Event 3. NEI 99-01 HU1

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Category: H - Hazards Subcategory: 1 - Security

Initiating Condition: HOSTILE ACTION within the SECURITY OWNER CONTROLLED AREA or airborne attack threat within 30 minutes

EAL: HA1.1 Alert A HOSTILE ACTION is occurring or has occurred within the SECURITY OWNER CONTROLLED AREA as reported by ANO 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):

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 ANO 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 ANO. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the SECURITY OWNER CONTROLLED AREA).

HOSTILE FORCE - One or more individuals who are engaged in a determined assault, overtly or by stealth and deception, equipped with suitable weapons capable of killing, maiming, or causing destruction.

PROJECTILE - An object directed toward a Nuclear Power Plant that could cause concern for its continued operability, reliability, or personnel safety.

PROTECTED AREA - An area clearly demarcated by a fence or building wall with an entrance portal that is regulated by Security Personnel to control access.

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SECURITY OWNER CONTROLLED AREA - The SOCA is the area demarcated as a Vehicle Barrier System (VBS) consisting of passive elements including a series of large concrete blocks on the inside of a delay fence with early warning capabilities. The SOCA is the area between the SOCA Fence and the PROTECTED AREA Boundary.

Basis:

This IC addresses the occurrence of a HOSTILE ACTION within the SECURITY 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 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 [and Independent Spent Fuel Storage Installation Security Program]. As time and conditions allow, these events require a heightened state of readiness by the plant staff and implementation of onsite protective measures (e.g., evacuation, dispersal or sheltering). The 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 EAL does not apply to incidents that are accidental events, acts of civil disobedience, or otherwise are not a HOSTILE ACTION perpetrated by a HOSTILE FORCE. Examples include the crash of a small aircraft, shots from hunters, physical disputes between employees, etc. Reporting of these types of events is adequately addressed by other EALs, or the requirements of 10 CFR 73.71 or 10 CFR 50.72.

The first threshold is applicable for any HOSTILE ACTION occurring, or that has occurred, in the SECURITY OWNER CONTROLLED AREA.

The second threshold addresses the threat from the impact of an aircraft on the plant, and the anticipated arrival time is within 30 minutes. The intent of this EAL is to ensure that threat-related notifications are made in a timely 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 OP-1203.048 Security Event (ref. 2).

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.

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In some cases, it may not be readily apparent if an aircraft impact within the SECURITY 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 Security Plan for ANO (ref. 1).

Escalation of the emergency classification level would be via IC HS1.

Reference(s):

1. ANO Security Plan 2. OP-1203.048 Security Event 3. NEI 99-01 HA1

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Category: H - Hazards Subcategory: 1 - Security

Initiating Condition: HOSTILE ACTION within the PROTECTED AREA EAL: HS1.1 Site Area Emergency A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by ANO Security Shift Supervision Mode Applicability:

All Definition(s):

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 ANO 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 ANO. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the SECURITY OWNER CONTROLLED AREA).

HOSTILE FORCE - One or more individuals who are engaged in a determined assault, overtly or by stealth and deception, equipped with suitable weapons capable of killing, maiming, or causing destruction.

PROJECTILE - An object directed toward a Nuclear Power Plant that could cause concern for its continued operability, reliability, or personnel safety.

PROTECTED AREA - An area clearly demarcated by a fence or building wall with an entrance portal that is regulated by Security Personnel to control access.

SECURITY OWNER CONTROLLED AREA - The SOCA is the area demarcated as a Vehicle Barrier System (