Text

Response to U.S. Nuclear Regulatory Commission Request for Additional Information Regarding License Amendment Request for Adoption of Emergency Action Level Schemes Pursuant to NEI 99-01, Revision 6
	ML20352A231
Person / Time
Site:	Waterford
Issue date:	12/15/2020
From:	Gaston R; Entergy Operations
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	W3F1-2020-0062
	Download: ML20352A231 (253)
	v • d • e

Entergy Operations, Inc.

1340 Echelon Parkway Jackson, MS 39213 Tel 601-368-5138 Ron Gaston Director, Nuclear Licensing 10 CFR 50.90 W3F1-2020-0062 December 15, 2020 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

Subject:

Response to U. S. Nuclear Regulatory Commission Request for Additional Information Regarding License Amendment Request for Adoption of Emergency Action Level Schemes Pursuant to NEI 99-01, Revision 6 Waterford Steam Electric Station, Unit 3 NRC Docket No. 50-382 Renewed Facility Operating License No. NPF-38

References:

1) Entergy Operations, Inc. (Entergy) letter to U. S. Nuclear Regulatory Commission (NRC), "License Amendment Request Adoption of Emergency Action Level Schemes Pursuant to NEI 99-01, Revision 6,"

W3F1-2020-0036, (ADAMS Accession No. ML20153A457),

dated June 1, 2020

2) NRR E-mail capture, U. S. NRC to Entergy, "NRC Request for Additional Information - WF3 EAL Scheme Change -

L 2020 LLA-0122.docx," (ADAMS Accession No. ML20289A350),

dated October 5, 2020

3) NRC letter "U. S. Nuclear Regulatory Commission Review and Endorsement of NEI 99-01, Revision 6, dated November, 2012 (TAC No. D92368)," (ADAMS Accession No. ML12346A463),

dated March 28, 2013 By letter dated June 1, 2020 (Reference 1), Entergy Operations Inc., (Entergy) requested an amendment to the license Steam Electric Station, Unit 3 (Waterford 3) to revise the Emergency Plan for Waterford 3 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 correspondence dated October 5, 2020 (Reference 2), the NRC staff informed Entergy that they have reviewed the license amendment request and have determined that additional

W3F1-2020-0062 Page 2 of 2 information is required to complete the review. A clarification call between the NRC and Entergy was previously held on September 24, 2020.

The additional information requested by the NRC in Reference 2 is provided in the Enclosure to this letter. Based upon the responses to the requested additional information, the submittal has been modified as shown in the markup in Enclosure attachment 1 with the final copy included in Enclosure attachment 2.

In Reference 1, Entergy stated that once approved by the NRC, the amendment shall be implemented by August 31, 2021. Per this correspondence, Entergy is changing the planned implementation date to be completed within 180 days following the issuance of the amendment.

This letter contains no new regulatory commitments.

Should you have any questions or require additional information, please contact Paul Wood, Regulatory Assurance Manager, at 504-464-3786.

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

Executed on December 15, 2020.

Respectfully, Ron Gaston RWG/rrd

Enclosure:

Response to NRC Request for Additional Information : Waterford 3 SES EAL Basis Document Proposed Changes : Copy of Revised Waterford 3 SES EAL Basis Document cc: NRC Region IV Regional Administrator NRC Senior Resident Inspector - Waterford 3 NRC Project Manager Waterford 3 Louisiana Department of Environmental Quality, Office of Environmental Compliance

Enclosure W3F1-2020-0062 Response to NRC Request for Additional Information

W3F1-2020-0062 Enclosure Page 1 of 8 RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION Regulatory Requirements/Background The requirements of Section 50.47(b)(4) to Title 10 of the Code of Federal Regulations (10 CFR) state, in part:

A standard emergency classification and action level scheme, the bases of which include facility system and effluent parameters, is in use by the nuclear facility licensee...

In addition, the requirements of Section IV.B.1 to Appendix E of 10 CFR Part 50 state, in part:

The means to be used for determining the magnitude of, and for continually assessing the impact of, the release of radioactive materials shall be described, including emergency action levels that are to be used as criteria for determining the need for notification and participation of local and State agencies, the Commission, and other Federal agencies, and the emergency action levels that are to be used for determining when and what type of protective measures should be considered within and outside the site boundary to protect health and safety. The emergency action levels shall be based on in-plant conditions and instrumentation in addition to onsite and offsite monitoring The most recent industry EAL scheme development guidance is provided in the Nuclear Energy Institute (NEI) document NEI 99-01, Development of Emergency Action Levels for Non-Passive Reactors, Revision 6 (ADAMS Accession Number ML12326A805). By letter dated March 28, 2013 (ADAMS Accession No. ML12346A463), the NRC endorsed NEI 99-01, Revision 6, as acceptable generic (i.e., non-plant specific) EAL scheme development guidance.

The licensee proposed to revise its current EAL scheme to one based on NEI 99-01, Revision 6.

Requests for Additional Information Request 1 The licensees current AU1.1 includes the Dry Cooling Tower Sumps or Turbine Building Industrial Waste Sump Monitors. However, the proposed AU1.1 does not include the Dry Cooling Tower Sumps nor Turbine Building Industrial Waste Sump Monitors. The licensees basis in the LAR for the proposed AU1.1 states, [T]he values used on the Dry Cooling Tower and Turbine Building sump discharge are based on the pathway being aligned to the Storm Water or Discharge Canal vice the circulating water system and are not applicable if the pathway is aligned to the circulating water system. Additionally, the WF3 EAL Comparison Matrix in the LAR provides that the gland steam condenser and main condenser exhausts are not included because they are not normally radioactive. However, the guidance in NEI 99-01, Revision 6, for AU1 provides that listed monitors should include the effluent monitors described in their Offsite Dose Calculation Manual (ODCM) and not simply based on the current line-up or radioactivity level.

Therefore, the NRC staff requests the licensee to provide a justification, in greater detail, that supports removing the Condenser Exhaust, Dry Cooling Tower Sumps, and Turbine Building

W3F1-2020-0062 Enclosure Page 2 of 8 Industrial Sumps radiation monitors from AU1, including a description of (1) how these effluent flow paths will be monitored by a downstream radiation monitor, or (2) that these flow paths are not effluent flow paths as described in the WF3 ODCM or Updated Final Safety Analysis Report as effluent flow paths. If the licensee determines that this this request identifies a concern with the current LAR, then the staff requests the licensee to either propose a change to the LAR or another solution Entergy Response Turbine Building and Dry Cooling Tower Sumps The NEI 99-01 AU1 basis for monitored liquid effluent thresholds include normally occurring continuous radioactivity releases pathways and planned batch releases from non-continuous release pathways.

The Turbine Building and Dry Cooling Tower Sumps are not normally occurring sources for continuous or non-continuous radioactive release pathways. They are not designed to be radioactive systems.

Per FSAR 11.2.1, The contents of turbine building sumps and detergent wastes will be routinely discharged unprocessed due to their very small potential for radioactive contamination (the Turbine Building and Dry Cooling Tower Sumps are monitored as discharges for Tritium, but do not typically contain gamma emitting isotopes).

Additional system or barrier failures are necessary for the Turbine Building or Dry Cooling Tower Sumps to be contaminated with radioactivity.

In the event that those sumps were to be contaminated from a secondary source of activity, releases from these sumps would still be monitored by the installed radiation monitors.

Although it is possible to align Turbine Building and Dry Cooling Tower Sump discharge to the discharge canal, historic discharge permit records indicate that discharges to the discharge canal are not performed. If a radioactive release from the Turbine Building and Dry Cooling Tower Sumps were necessary due to failures in other systems or barriers, it would be indicated by EAL AU1.2.

Additionally, a paragraph was included with the initial submittal in error. At the top of page 32 in the original Clean submittal, a paragraph was included that stated:

The values used on the Dry Cooling Tower and Turbine Building sump discharge are based on the release pathway being aligned to the Storm Water System or Discharge Canal vice the circulating water system and are not applicable if the pathway is aligned to the circulating water system.

This paragraph should not have been included and has now been deleted.

Gland Steam Condenser and Main Condenser Exhausts The NEI 99-01 AU1 basis for monitored gaseous effluent thresholds include normally occurring continuous radioactivity releases pathways and planned batch releases from non-continuous release pathways.

W3F1-2020-0062 Enclosure Page 3 of 8 The Gland Steam Condenser and Main Condenser Exhausts are not normally occurring sources for continuous or non-continuous radioactive release pathways. Additional system or barrier failures would be required for the Gland Steam Condenser and Main Condenser Exhausts to be contaminated with radioactivity.

Per FSAR Section 2.1.1.3, the Turbine Gland Seal System exhaust and the Main Condenser Evacuation System exhaust are not normally radioactive.

Since these pathways are not sources for normally occurring continuous radioactivity releases or for planned batch releases from non-continuous release pathways, and are not normally radioactive, they do not meet the NEI 99-01 criteria for use as EAL thresholds.

Request 2 The proposed threshold values for AG1.1, AS1.1, and AA1.1, if approved, would change substantially from the currently approved values for WF3. Additionally, if approved, the threshold values for PRM-IRE-0110-1 to 3 and PRM-IRE-3032 1 to 3 would also change from Ci/sec to Ci/cc (cubic centimeter).

The NRC staff requests the licensee to provide a justification that clearly indicates why the proposed threshold values have changed from their current values and how the proposed values correspond to the appropriate total effective or committed dose equivalents.

Entergy Response AA1.1 Differences The current WF3 AA1.1 gaseous effluent monitor EAL thresholds are based on NEI 99-01 Rev 5, which establish the values as more than 200 times the Radiological Effluent Technical Specifications [ODCM Limits].

The proposed WF3 AA1.1 gaseous effluent monitor EAL thresholds are based on NEI 99-01 Rev 6, which establish the values as 1/100 the AG1.1 EAL threshold values, which are based on EPA-400 protective action guides (PAGs) of 1 Rem TEDE or 5 Rem CDE Thyroid at the site boundary.

The AA1.1 basis and development methodologies for the current WF3 EALs is based on NEI 99-01 Rev 5 and the proposed WF3 EALs are based on NEI 99-01 Rev 6.

AG1.1 and AS1.1 Differences The current and proposed WF3 AG1.1 and AS1.1 EALs share an underlying EPA-400 PAG limit basis.

The current WF3 AG1.1 and AS1.1 EALs were documented in calculations performed in 2005 using inputs from power uprate based source term, dose factors and I:NG ratio, and historical X/Q values.

The proposed WF3 AG1.1 and AS1.1 are developed using the WF3 site specific dose assessment model using URI/RASCAL, URI/RASCAL uses site adjusted NUREG-1940 source

W3F1-2020-0062 Enclosure Page 4 of 8 term and process reduction factors, accident X/Q dispersion vales from recent historic predominant meteorological inputs, and design basis vent flows.

Monitor Unit Differences 3ODQWPRQLWRUHIIOXHQWLQGLFDWLRQLVSURYLGHGLQXQLWVRIERWK&LVHFDQG&LFF:DWHUIRUGKDV

reviewed the usefulness of having both sets of units in Table A-1. Based on this review, WateUIRUGZLOOUHPRYHWKH($/WKUHVKROGYDOXHVLQXQLWVRI&LFFDQGZLOORQO\UHWDLQWKHXQLWVLQ

&LVHFUHPDLQLQJFRQVLVWHQWZLWK:DWHUIRUG¶VFXUUHQW($/VFKHPH

Request 3 The proposed CS1.1 threshold value includes the condition, RVLMS [Reactor Vessel Water Level Monitoring System] upper plenum 0%, which appears to be approximately the top of active fuel and not approximately the bottom inner diameter (ID) of the reactor coolant system (RCS) loop. However, this threshold value does not appear to be consistent with a standard EAL scheme in NEI 99-01, Revision 6. As such, the proposed change could result in the licensee delaying a declaration of a site area emergency classification.

The current required threshold, as provided in Rev 5, was based on 0% being approximately the bottom of the RCS loop (i.e., a RVLMS level of 0% was the first observable point below the bottom ID of the RCS loop), which is why the NRC approved Rev 5 (refer to the WF3 submittal letter dated September 16, 2010 (ADAMS Accession No. ML102630124). However, in this LAR, the licensee is now stating that the 0% value is approximately top of active fuel (i.e. the proposed WF3 EAL Comparison Matrix states that the RVLMS 0% sensor is the closest indication of level near top of active fuel), which means that the licensee would inappropriately delay a site area emergency. It is not clear that a RVLMS upper level plenum indication of 0%

is an appropriate threshold value for the proposed threshold value for CS1.1.

The NRC staff requests the licensee to explain the specific characteristics and any limitations on the available level indication unique to WF3 that support the proposed deviation (i.e., using top of active fuel) from the guidance provided by NEI 99-01, Revision 6, and the extent of the deviation from the levels provided by NEI 99-01, Revision 6. If the licensee determines that this this request identifies a concern with the current LAR, then the staff requests the licensee to either propose a change to the LAR or another solution.

Entergy Response The cited statement in the LAR Comparison Matrix was not intended to infer that the RVLMS 0% indication is being used as an indicator of RCS level at the top of active fuel. There are 5 detectors in the RVLMS system to indicate RCS level in the plenum area. RVLMS upper plenum 0% indicates RCS level is 12.6 in. above the fuel alignment plate (FAP). The next detector above that will indicate voided at 32.6 inches above the FAP. This detector corresponds to the bottom of the RCS loop penetration into the reactor vessel. The top of the active fuel is 22 inches below the Fuel Alignment Plate.

In NEI 99-01, the developer notes for CS1.b describe that the intended PWR site-specific level is 6" below the bottom ID of the RCS loop. This is the level at 6 below the bottom ID of the reactor vessel penetration and not the low point of the loop. Using detector 7 of the RVLMS System would be at the reactor vessel penetration vice 6 inches below the vessel penetration.

W3F1-2020-0062 Enclosure Page 5 of 8 Reactor upper plenum level of 0% is the appropriate indication of RCS level below the bottom of the RCS hot leg.

This wording and bases are in alignment with the generic intent of the CS1 EAL #1 bases of water level below the RCS hot-leg. The plants cold shutdown RCS level monitoring capability represents the same technical capability as that for the NEI 99-01 Revision 5 based Waterford 3 EAL scheme that was previously approved by NRC in a letter dated July 18, 2011 (ADAMS Accession No. ML111380558). The current proposed use of RVLMS upper plenum level of 0%

in CS1.1 is consistent with the use of this indication in the approved EAL scheme.

Request 4 The proposed EAL threshold value for CS1.2 and CG1.1 includes the condition, Representative CETs [Core Exit Thermocouples] indicate superheat. In its LAR, the licensee states and the NRC agrees, that superheated conditions in the core can only occur with core uncovery.

However, given uncertainties associated with establishing superheated conditions and subsequent monitoring of those conditions, the NRC staff finds that superheated conditions represent a core uncovery indication and not a specific reactor vessel level indication.

The NRC staff requests the licensee to explain why it treats CETs indicating superheat as an actual level indication and not an indication of core uncovery. If the licensee determines that this this request identifies a concern with the current LAR, then the staff requests the licensee to either propose a change to the LAR or another solution.

Entergy Response CET superheat is not indicative of a specific RCS water level, but is rather an indication that the core has been uncovered. Waterford 3 does not have the capability to monitor reactor vessel level at or below the top of active fuel and therefore uses superheat indication on CETs for EAL CS1.2 and CG 1.1. The Waterford 3 Reactor Vessel Level Monitoring System (RVLMS) does not provide monitoring below the level of 12.6 above the fuel alignment plate. Backup means of level monitoring used in mode 6 do not provide monitoring below 12 ft MSL. As level falls below the top of active fuel, CETs will begin to indicate superheat conditions. This difference of using CET superheat rather than an actual level indication corresponding to the NEI EAL is because of plant design. The Entergy treatment of this EAL provides a progression consistent with the NEI guidance for a Site Area Emergency in that the RCS level drop indicates a loss of the RCS barrier and the superheat condition indicates a potential loss (or loss) of the fuel clad barrier.

The plants cold shutdown RCS level monitoring capability represents the same technical capability as that for the NEI 99-01 Revision 5 based Waterford 3 EAL scheme that was previously approved by NRC in a letter dated July 18, 2011 (ADAMS Accession No. ML111380558). The current proposed use of the CET superheat indication in CS1.2 and CG1.1 is consistent with the use of this indication in the approved EAL scheme.

Request 5 The proposed threshold value for SG1.1 proposes to retain a 4-hour coping time to restore alternating current (AC) power based on the site blackout coping analysis performed in

W3F1-2020-0062 Enclosure Page 6 of 8 conformance with 10 CFR 50.63. The proposed SG1.1 provides that mitigative strategies using non-safety related power sources may be effective in supplying power to the safety buses.

Although 10 CFR 50.63 continues to apply, 10 CFR 50.155, Mitigation of beyond-design-basis events, also applies. Considering that FLEX strategies should provide the capability to provide core cooling for an extend loss of AC power (ELAP), using a coping time based on 10 CFR 50.63 may result in an unnecessary declaration of an unnecessary General Emergency classification.

a. The NRC staff requests the licensee to clarify whether WF3 currently has a procedure that could be used to extend the availability of direct current (DC) power, such as a FLEX Implementing Guideline for an Extended Loss of AC power.

If so, then the NRC staff requests the licensee to explain whether the WF3 procedure or guideline provides a strategy that could reasonably provide core cooling (and thus availability of DC power) for a coping time significantly longer than the 10 CFR 50.63 coping time or until an alternate AC power source, such as FLEX, is aligned.

b. If WF3 has the capability to provide mitigation for greater than the coping time based on the site blackout coping analysis performed in conformance with 10 CFR 50.63, then the NRC staff requests the licensee to explain why WF3 proposes to potentially declare a general emergency when FLEX mitigation strategies could provide a reasonable probability of success. If the licensee determines that this this request identifies a concern with the current LAR, then the staff requests the licensee to either propose a change to the LAR or another solution.

Entergy Response A proposed revised SG1.1 omits the Station Blackout (SBO) coping time threshold. As proposed, the General Emergency classification would be based on a loss of all onsite and offsite AC power to the emergency buses with indications of degraded core cooling or the inability to establish and maintain core heat removal. The Waterford-3 SBO analysis and derived coping time was determined in accordance with 10 CFR 50.63 and Regulatory Guide 1.155. This analysis does not take credit for plant capabilities in place to mitigate the effects of an extended loss of AC power (ELAP). These capabilities were developed and implemented to meet the requirements of NRC Orders EA-12-049 and EA-12-051, and pending regulations in 10 CFR 50.155 (per SECY-16-0142).

In accordance with OP-902-005 Station Blackout Recovery, operators will declare an ELAP within 60 min. of the loss of all AC power to the emergency buses and direct implementation of FLEX Support Guidelines, including the deployment of dedicated portable equipment and performance of DC load shedding. Even if no AC emergency bus is energized, these actions will maintain or restore core cooling, containment, and spent fuel pool cooling capabilities indefinitely. Therefore, the underlying basis for the generic EAL coping time statement, that power must be restored to an AC emergency bus within a fixed amount of time to avoid a severe challenge to one or more fission product barriers, is not valid for Waterford-3.

Additionally, the omission of the SBO coping time threshold does not remove the attribute of a likely General Emergency declaration prior to meeting the IC FG1 thresholds for ELAP events in which the RCS barrier has not been lost.

W3F1-2020-0062 Enclosure Page 7 of 8 Request 6 The LARs Technical Basis section for the proposed changes to SU6.1, SU6.2, SA6.1, and SS6.1 provides that after a successful reactor trip, neutron power should immediately decrease to approximately 6% due to a prompt drop. The licensee further provides in its LAR that, for the purpose of emergency classification, reactor power less than or equal to 5% is used to identify a successful reactor trip.

The NRC staff notes that although the reactor trip neutron power should immediately decrease due to a prompt drop, neutron power will be approximately 6% of the reactor neutron power prior to the trip and may be significantly below 5% due to the prompt drop. As such, a power level of 5% does not provide positive indication that reactor power is lowering and that the reactor is shutdown. Additionally, the guidance provided in NEI 99-01, Revision 6, discusses power in relation to EAL mode applicability and does not include a reactor power level as a threshold value. However, NEI 99-01, Revision 6, does provide that developers may include site specific emergency operating procedure criteria indicative of a reactor trip. As such, the entry criteria for a site specific anticipated transient without scram procedure would be appropriate criteria indicating whether a reactor was successfully shutdown or if the reactor was not successfully shutdown and further actions were required to assure reactor shutdown and maintain positive control of reactivity.

The NRC staff requests the licensee to justify why a reactor power value of 5% or less is the only criteria that is used to determine whether the licensee should enter the WF3 site specific procedure. If the licensee determines that this this request identifies a concern with the current LAR, then the staff requests the licensee to either propose a change to the LAR or another solution.

Entergy Response NEI 99-01 revision 6 describes that the potential degradation of the level of safety of the plant for a failure to trip is continued power generation post trip of a magnitude to cause a challenge to core cooling or the RCS heat removal safety functions. The Waterford-3 reactivity control acceptance criteria (EOP) does not include a reactor power value, simply that power is dropping with a negative startup rate. Waterford-3 thus proposes a power-based criteria consistent with other PWR designs and consistent with the Entergy fleet sites that have implemented the NEI 99-01 revision 6 EAL scheme.

The referenced Calculation EC-S98-001 states that after the reactor is tripped, neutron power should immediately drop to approximately 6% due to the prompt drop from a successful trip.

Reactor power will then asymptomatically approach a negative period for 15-20 minutes to about 10-6% power. Therefore, a prompt post-trip drop of reactor power to less than 5% is a clear indication of a successful reactor trip relative to the intended EAL bases.

Following in line with the developer notes (Developers may include site-specific EOP criteria indicative of a successful reactor shutdown in an EAL statement, the Basis or both), additional data has been added to the basis for these 4 EALs to include information from Waterfords EOP criteria and additional details from the Waterford EOP technical guide. This additional data will provide additional data to the Emergency Director to assist in evaluating these EALs.

W3F1-2020-0062 Enclosure Page 8 of 8 The use of 5% power as an indication of a successful reactor trip is consistent with the NEI 99-01 Revision 5 based Waterford 3 EAL scheme that was previously approved by NRC in a letter dated July 18, 2011 (ADAMS Accession No. ML111380558).

Attachments:

1. Waterford 3 SES EAL Basis Document Proposed Changes

2. Copy of Revised Waterford 3 SES EAL Basis Document

Enclosure, Attachment 1 W3F1-2020-0062 Waterford 3 SES EAL Basis Document Proposed Changes

Waterford 3 SES EAL Basis Document Revision A Pages changed by RAI responses Waterford 3 Steam Electric Station EAL Technical Basis Page 1 of 228

Waterford 3 SES EAL Basis Document Revision A 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 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 A-1 Effluent Monitor Classification Thresholds Release Point Monitor GE SAE Alert UE 4.01 E+08 4.01 E+07 4.01 E+06 2.27 E+05 PRM-IRE-0110-4 Ci/sec Ci/sec Ci/sec Ci/sec Plant Stack WRGM 6.54 E+01 6.54 E+00 6.54 E-01 5.81 E-03 PRM-IRE-0110-1 to 3 Gaseous Ci/cc Ci/cc Ci/cc Ci/cc 1.48 E+10 1.48 E+09 1.48 E+08 2.27 E+05 PRM-IRE-3032-4 Fuel Handling Bldg. Ci/sec Ci/sec Ci/sec Ci/sec Exhaust WRGM 7.85 E+03 7.85 E+02 7.85 E+01 1.60 E-02 PRM-IRE-3032-1 to 3 Ci/cc Ci/cc Ci/cc Ci/cc Circulating Water 7.27 E-04 PRM-IRE-1900 N/A N/A N/A Discharge Monitor Ci/ml Liquid Waste Liquid 2.40 E-03 Management PRM-IRE-0647 N/A N/A N/A Discharge Monitor Ci/ml Boron Management 2.40 E-03 PRM-IRE-0627 N/A N/A N/A Discharge Monitor Ci/ml Page 30 of 228

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases The values used on the Dry Cooling Tower and Turbine Building sump discharge are based on the release pathway being aligned to the Storm Water System or Discharge Canal vice the circulating water system and are not applicable if the pathway is aligned to the circulating water system.

Escalation of the emergency classification level would be via IC AA1.

Reference(s):

1. UNT-005-014 Offsite Dose Calculation Manual

2. EC86890, EP-CALC-WF3-1701, Radiological Effluent EAL Values

3. NEI 99-01 AU1 Page 32 of 228

Waterford 3 SES EAL Basis Document Revision A 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 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 A-1 Effluent Monitor Classification Thresholds Release Point Monitor GE SAE Alert UE 4.01 E+08 4.01 E+07 4.01 E+06 2.27 E+05 PRM-IRE-0110-4 Ci/sec Ci/sec Ci/sec Ci/sec Plant Stack WRGM 6.54 E+01 6.54 E+00 6.54 E-01 5.81 E-03 PRM-IRE-0110-1 to 3 Gaseous Ci/cc Ci/cc Ci/cc Ci/cc 1.48 E+10 1.48 E+09 1.48 E+08 2.27 E+05 PRM-IRE-3032-4 Fuel Handling Bldg. Ci/sec Ci/sec Ci/sec Ci/sec Exhaust WRGM 7.85 E+03 7.85 E+02 7.85 E+01 1.60 E-02 PRM-IRE-3032-1 to 3 Ci/cc Ci/cc Ci/cc Ci/cc Circulating Water 7.27 E-04 PRM-IRE-1900 N/A N/A N/A Discharge Monitor Ci/ml Liquid Waste Liquid 2.40 E-03 Management PRM-IRE-0647 N/A N/A N/A Discharge Monitor Ci/ml Boron Management 2.40 E-03 PRM-IRE-0627 N/A N/A N/A Discharge Monitor Ci/ml Page 35 of 228

Waterford 3 SES EAL Basis Document Revision A 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 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 A-1 Effluent Monitor Classification Thresholds Release Point Monitor GE SAE Alert UE 4.01 E+08 4.01 E+07 4.01 E+06 2.27 E+05 PRM-IRE-0110-4 Ci/sec Ci/sec Ci/sec Ci/sec Plant Stack WRGM 6.54 E+01 6.54 E+00 6.54 E-01 5.81 E-03 PRM-IRE-0110-1 to 3 Gaseous Ci/cc Ci/cc Ci/cc Ci/cc 1.48 E+10 1.48 E+09 1.48 E+08 2.27 E+05 PRM-IRE-3032-4 Fuel Handling Bldg. Ci/sec Ci/sec Ci/sec Ci/sec Exhaust WRGM 7.85 E+03 7.85 E+02 7.85 E+01 1.60 E-02 PRM-IRE-3032-1 to 3 Ci/cc Ci/cc Ci/cc Ci/cc Circulating Water 7.27 E-04 PRM-IRE-1900 N/A N/A N/A Discharge Monitor Ci/ml Liquid Waste Liquid 2.40 E-03 Management PRM-IRE-0647 N/A N/A N/A Discharge Monitor Ci/ml Boron Management 2.40 E-03 PRM-IRE-0627 N/A N/A N/A Discharge Monitor Ci/ml Mode Applicability:

All Page 42 of 228

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases Table A-1 Effluent Monitor Classification Thresholds Release Point Monitor GE SAE Alert UE 4.01 E+08 4.01 E+07 4.01 E+06 2.27 E+05 PRM-IRE-0110-4 Ci/sec Ci/sec Ci/sec Ci/sec Plant Stack WRGM 6.54 E+01 6.54 E+00 6.54 E-01 5.81 E-03 PRM-IRE-0110-1 to 3 Gaseous Ci/cc Ci/cc Ci/cc Ci/cc 1.48 E+10 1.48 E+09 1.48 E+08 2.27 E+05 PRM-IRE-3032-4 Fuel Handling Bldg. Ci/sec Ci/sec Ci/sec Ci/sec Exhaust WRGM 7.85 E+03 7.85 E+02 7.85 E+01 1.60 E-02 PRM-IRE-3032-1 to 3 Ci/cc Ci/cc Ci/cc Ci/cc Circulating Water 7.27 E-04 PRM-IRE-1900 N/A N/A N/A Discharge Monitor Ci/ml Liquid Waste 2.40 E-03 Liquid Management PRM-IRE-0647 N/A N/A N/A Discharge Monitor Ci/ml Boron Management 2.40 E-03 PRM-IRE-0627 N/A N/A N/A Discharge Monitor Ci/ml 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 indicators operability, the conditions existence, or the reports 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.

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.

Page 48 of 228

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases Category: S -System Malfunction Subcategory: 1 - Loss of Safety Bus AC Power Initiating Condition: Prolonged loss of all offsite and all onsite AC power to safety buses EAL:

SG1.1 General Emergency Loss of all offsite and all onsite AC power to 4160 VAC safety buses 3A and 3B AND EITHER:

Long-term RCS heat removal capability is not likely to be established and maintained per FIG-001, FLEX Implementing Guideline, Extended Loss of AC Power Restoration of at least one 4160 VAC safety bus in < 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is not likely (Note 1)

Representative CETs reading > 1,200ºF 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 safety buses that result in degraded core cooling. 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 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 Page 197 of 228

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases 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 long-term RCS heat removal capability is not likely to be established and maintained per FIG-001, FLEX Implementing Guideline, Extended Loss of AC Power. FIG-001 contains attachments that list execution timelines for use of Train A or Train B power. These timelines describe the execution times for Flex actions, such as Deep Load Shed and aligning Flex DG, to assist in evaluating the success of Flex 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 safety 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 greater likelihood of challenges to multiple fission product barriers.

The 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to restore AC power is based on the site blackout coping analysis performed in conformance with 10 CFR 50.63 (ref. 6).

The estimate for restoring at least one safety bus or establishing long term cooling with Flex equipment 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.

Temporary Emergency Diesels (TEDs) can be credited if already installed in accordance with site procedures (ref. 4, 5).

Reference(s):

1. UFSAR Section 8.1, Onsite Power System

2. UFSAR Section 8.2, Offsite Power System

3. OP-902-005 Station Blackout Recovery

4. Technical Specifications 3/4.8.1 A.C. Sources

5. ME-001-012, Temporary Power from Temporary Diesel for 3A2 and 3B2 4KV Buses (Modes 1-6)

6. F I G - 0 0 1 , FLEX Implementing Guideline, Extended Loss of AC Power UFSAR Appendix 8.1A Station Blackout (SBO) Evaluation

7. UFSAR Section 1.9A Inadequate Core Cooling Instrumentation

8. CEOG Generic Accident Management Guidelines - Phase 1, Initial Diagnosis

9. NEI 99-01 SG1 Page 198 of 228

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases 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 (manual reactor trip push buttons or DRT) 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:

Consistent with OP-902-000, Standard Post Trip Actions, and its associated technical guide TGOP-902-000, a successful reactor trip includes inserting CEAs. Actions are included for failure of individual CEAs to insert. Should an automatic reactor trip fail to occur, the operator is directed to use the manual pushbuttons on the main control board. Should this fail, the operator is directed to manually initiate Diverse Reactor Trip (DRT), also from the main control board. Should this also fail, the operator is directed to open the feeder breakers for the 32 bus, which supply power to the CEA motor generator set and then reclose the breakers.

After a successful reactor trip neutron power should immediately drop to approximately 6%

due to prompt drop. Therefore, for the purpose of emergency classification, reactor power less than or equal to 5% is used to identify a successful reactor trip (ref. 2).

This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor trip that results in a reactor shutdown, and either a subsequent operator manual action taken at the reactor control consoles or an automatic trip is successful in shutting down the reactor.

This event is a precursor to a more significant condition and thus represents a potential degradation of the level of safety of the plant.

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 Page 211 of 228

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases 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 (manual reactor trip push buttons or DRT) 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):

None Basis:

Consistent with OP-902-000, Standard Post Trip Actions, and its associated technical guide TGOP-902-000, a successful reactor trip includes inserting CEAs. Actions are included for failure of individual CEAs to insert. Should an automatic reactor trip fail to occur, the operator is directed to use the manual pushbuttons on the main control board. Should this fail, the operator is directed to manually initiate Diverse Reactor Trip (DRT), also from the main control board. Should this also fail, the operator is directed to open the feeder breakers for the 32 bus, which supply power to the CEA motor generator set and then reclose the breakers.

After a successful reactor trip neutron power should immediately drop to approximately 6%

due to prompt drop. Therefore, for the purpose of emergency classification, reactor power less than or equal to 5% is used to identify a successful reactor trip (ref. 2).

This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor trip that results in a reactor shutdown, and either a subsequent operator manual action taken at the reactor control consoles or an automatic trip is successful in shutting down the reactor.

This event is a precursor to a more significant condition and thus represents a potential degradation of the level of safety of the plant.

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.

Page 214 of 228

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases 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 control console (manual reactor trip push buttons and DRT) 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:

Consistent with OP-902-000, Standard Post Trip Actions, and its associated technical guide TGOP-902-000, a successful reactor trip includes inserting CEAs. Actions are included for failure of individual CEAs to insert. Should an automatic reactor trip fail to occur, the operator is directed to use the manual pushbuttons on the main control board. Should this fail, the operator is directed to manually initiate Diverse Reactor Trip (DRT), also from the main control board. Should this also fail, the operator is directed to open the feeder breakers for the 32 bus, which supply power to the CEA motor generator set and then reclose the breakers.

After a successful reactor trip neutron power should immediately drop to approximately 6%

due to prompt drop. Therefore, for the purpose of emergency classification, reactor power less than or equal to 5% is used to identify a successful reactor trip (ref. 2).

This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor trip that results in a reactor shutdown, and subsequent operator manual actions taken at the reactor control consoles to shutdown the reactor are also unsuccessful. This condition represents an actual or potential substantial degradation of the level of safety of the plant. An emergency declaration is required even if the reactor is subsequently shutdown by an action Page 217 of 228

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases 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:

Representative CET readings > 1,200ºF Any OP-902-008 Functional Recovery RCS/Core Heat Removal safety function criterion is not met 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 Definition(s):

None Basis:

Consistent with OP-902-000, Standard Post Trip Actions, and its associated technical guide TGOP-902-000, a successful reactor trip includes inserting CEAs. Actions are included for failure of individual CEAs to insert. Should an automatic reactor trip fail to occur, the operator is directed to use the manual pushbuttons on the main control board. Should this fail, the operator is directed to manually initiate Diverse Reactor Trip (DRT), also from the main control board. Should this also fail, the operator is directed to open the feeder breakers for the 32 bus, which supply power to the CEA motor generator set and then reclose the breakers.

After a successful reactor trip neutron power should immediately drop to approximately 6%

due to prompt drop. Therefore, for the purpose of emergency classification, reactor power less than or equal to 5% is used to identify a successful reactor trip (ref. 2).

This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor trip that results in a reactor shutdown, all subsequent operator actions to manually shutdown the reactor are unsuccessful, and continued power generation is challenging the capability to Page 219 of 228

Enclosure, Attachment 2 W3F1-2020-0062 Copy of Waterford 3 SES EAL Basis Document

Waterford 3 SES EAL Basis Document Revision A Waterford 3 Steam Electric Station EAL Technical Basis Page 1 of 225

Waterford 3 SES EAL Basis Document Revision A Table of Contents

1.0 INTRODUCTION

....................................................................................................... 3 2.0 DISCUSSION............................................................................................................. 3 2.1 Background........................................................................................................ 3 2.2 Fission Product Barriers..................................................................................... 4 2.3 Fission Product Barrier Classification Criteria .................................................... 4 2.4 EAL Organization............................................................................................... 5 2.5 Technical Bases Information.............................................................................. 7 2.6 Operating Mode Applicability ............................................................................. 8 3.0 GUIDANCE ON MAKING EMERGENCY CLASSIFICATIONS .................................. 9 3.1 General Considerations ..................................................................................... 9 3.2 Classification Methodology .............................................................................. 11

4.0 REFERENCES

........................................................................................................ 14 4.1 Developmental ................................................................................................. 14 4.2 Implementing ................................................................................................... 14 5.0 DEFINITIONS, ACRONYMS & ABBREVIATIONS .................................................. 15 5.1 Definitions (ref. 4.1.1 except as noted) ............................................................ 15 5.2 Abbreviations/Acronyms .................................................................................. 20 6.0 WF3-TO-NEI 99-01 Rev. 6 EAL CROSS-REFERENCE .......................................... 24 7.0 ATTACHMENTS ...................................................................................................... 28 7.1 Attachment 1, Emergency Action Level Technical Bases ................................ 28 Category A - Abnormal Rad Levels/ Rad Effluent ........................................... 29 Category C - Cold Shutdown / Refueling System Malfunction ........................ 65 Category E - Independent Spent Fuel Storage Installation (ISFSI) ............... 100 Category F - Fission Product Barrier Degradation ........................................ 103 Table F-1 Fission Product Barrier Threshold Matrix and Bases ....... 108 Category H - Hazards and Other Conditions Affecting Plant Safety ............. 153 Category S - System Malfunction.................................................................. 188 7.2 Attachment 2, Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases .. 226 Page 2 of 225

Waterford 3 SES EAL Basis Document Revision A

1.0 INTRODUCTION

This document provides an explanation and rationale for each Emergency Action Level (EAL) included in the EAL Upgrade Project for Waterford 3 Steam Electric Station (WF3). It should be used to facilitate review of the WF3 EALs and provide historical documentation for future reference. Decision-makers responsible for implementation of EP-001-001, Recognition and Classification of Emergency Conditions, 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 of conditions present. Use of this document for assistance is not intended to delay the emergency classification.

Because the information in a basis document can affect emergency classification decision-making (e.g., the Emergency 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 WF3 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 the Development of Emergency Action Levels for Non-Passive Reactors, November 2012 (ref. 4.1.1), WF3 conducted an EAL implementation upgrade project that produced the EALs discussed herein.

Page 3 of 225

Waterford 3 SES EAL Basis Document Revision A 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 containment building and connections up to and including the outermost containment isolation valves. This barrier also includes the main steam, feedwater, and blowdown line extensions outside the containment building up to and including the outermost secondary side isolation valve. Containment Barrier thresholds are used as criteria for escalation of the 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 Page 4 of 225

Waterford 3 SES EAL Basis Document Revision A 2.4 EAL Organization The WF3 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 WF3 EAL categories are aligned to and represent the NEI 99-01 Recognition Categories. Subcategories are used in the WF3 scheme as necessary to further divide the EALs of a category into logical sets of possible emergency classification thresholds. The WF3 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.

Page 5 of 225

Waterford 3 SES EAL Basis Document Revision A 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 1 - Security Affecting Plant Safety 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 1 - Confinement Boundary Installation (ISFSI)

Hot Conditions:

S - System Malfunction 1 - Loss of Safety Bus 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 1 - RCS Level Malfunction 2 - Loss of Safety Bus AC Power 3 - RCS Temperature 4 - Loss of Vital DC Power 5 - Loss of Communications 6 - Hazardous Event Affecting Safety Systems Page 6 of 225

Waterford 3 SES EAL Basis Document Revision A 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:

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.

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 Any. (See Section 2.6 for operating mode definitions)

Page 7 of 225

Waterford 3 SES EAL Basis Document Revision A Definitions:

If the EAL or Basis 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 WF3-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 1 Power Operation Keff 0.99, rated reactor thermal power > 5%

2 Startup Keff 0.99, rated reactor thermal power 5%

3 Hot Standby Keff < 0.99, average reactor coolant temperature 350ºF 4 Hot Shutdown Keff < 0.99, average reactor coolant temperature 350ºF > Tavg > 200ºF 5 Cold Shutdown Keff < 0.99, average reactor coolant temperature 200ºF 6 Refueling Keff < 0.95, average reactor coolant temperature 140ºF Fuel in the reactor vessel with the vessel head closure bolts less than fully tensioned or with the head removed DEF Defueled All reactor fuel removed from reactor vessel (full core off load during refueling or extended outage).

The plant operating mode that exists at the time that the event occurs (prior to any protective system or operator action being initiated in response to the condition) should be compared to the mode applicability of the EALs. If a lower or higher plant operating mode is reached before the emergency classification is made, the declaration shall be based on the mode that existed at the time the event occurred.

Page 8 of 225

Waterford 3 SES EAL Basis Document Revision A 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.

3.1.1 Classification Timeliness NRC regulations require the licensee to establish and maintain the capability to assess, classify, and declare an emergency condition within 15 minutes after the availability of indications to plant operators that an emergency action level has been exceeded and to promptly declare the emergency condition as soon as possible following identification of the appropriate emergency classification level. The NRC staff has provided guidance on implementing this requirement in NSIR/DPR-ISG-01, "Interim Staff Guidance, Emergency Planning for Nuclear Power Plants" (ref. 4.1.8).

For ICs and EALs that have a stipulated time duration (e.g., 15 minutes, 30 minutes, etc.), the Emergency Director is not allowed an additional 15 minutes to declare after the specified time limit is exceeded.

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 indicators operability, the conditions existence, or the reports accuracy. For example, validation could be accomplished through an instrument channel check, response on related or redundant indicators, or direct observation by plant personnel.

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

Page 9 of 225

Waterford 3 SES EAL Basis Document Revision A 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 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.

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Waterford 3 SES EAL Basis Document Revision A 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 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 an Alert should be declared.

If a declaration has been made and conditions for another EAL of the equal significance occurs, another initial declaration should not be made.

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.

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Waterford 3 SES EAL Basis Document Revision A 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.

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.

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Waterford 3 SES EAL Basis Document Revision A 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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Waterford 3 SES EAL Basis Document Revision A

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: 10 CFR 50.72 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 Technical Specifications Table 1.2, Operational Modes 4.1.7 WF3 Offsite Dose Calculation Manual 4.1.8 NSIR/DPR-ISG-01 Interim Staff Guidance, Emergency Planning for Nuclear Power Plants 4.1.9 Waterford 3 SES Emergency Plan 4.1.10 UFSAR Section 9.1.5 Spent Fuel Dry Cask Storage 4.1.11 Technical Specifications section 3/4.9.4 Containment Building Penetrations 4.1.12 WF3 Safeguards Contingency Plan 4.2 Implementing 4.2.1 EP-001-001, Recognition and Classification of Emergency Conditions 4.2.2 NEI 99-01 Rev. 6 to WF3 EAL Comparison Matrix 4.2.3 WF3 EAL Matrix Page 14 of 225

Waterford 3 SES EAL Basis Document Revision A 5.0 DEFINITIONS, ACRONYMS & ABBREVIATIONS 5.1 Definitions (ref. 4.1.1 except as noted)

Selected terms used in the Initiating Condition statement, the Emergency Action Level statement, 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.

Confinement Boundary The barrier(s) between spent fuel and the environment once the spent fuel is processed for dry storage. As related to the WF3 ISFSI, Confinement Boundary is considered the Multi-Purpose Canister (MPC) for the HI-STORM 100 dry storage system (ref. 4.1.10).

Containment Closure The procedurally defined actions taken to secure containment and its associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions. As applied to WF3, all Containment Penetrations are closed as described in the requirements of Technical Specifications section 3/4.9.4 Containment Building Penetrations (ref. 4.1.11) and OP-010-006, Outage Operations.

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)

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Waterford 3 SES EAL Basis Document Revision A 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.

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 WF3 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 WF3. 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 (SOCA)).

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Waterford 3 SES EAL Basis Document Revision A Hostile Force One or more individuals who are engaged in a determined assault, overtly or by stealth and deception, equipped with suitable weapons capable of killing, maiming, or causing destruction.

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

Impede(d)

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

Independent Spent Fuel Storage Installation (ISFSI)

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

Initiating Condition (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 The area encompassed by physical barriers (the security fence) and to which access is controlled into the vital areas of the plant (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.

Rupture(d)

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

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Waterford 3 SES EAL Basis Document Revision A 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.

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 area inside the SOCA Vehicle Barrier System (VBS) up to the PROTECTED AREA fence line. Access to this area is controlled by the SOCA Personnel Access Control Point. The SOCA is part of but not equal to the Owner Controlled Area as described or defined in the Waterford 3 Emergency Plan.

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 The border of the Exclusion Area or an area corresponding to a distance of 914 meters from the Waterford 3 reactor. Also referred to as the Exclusion Area Boundary (ref. 4.1.9).

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

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Waterford 3 SES EAL Basis Document Revision A Unplanned A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.

Unusual Event Events are in 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 indicators operability, the conditions existence, or the reports accuracy is removed. Implicit in this definition is the need for timely assessment.

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

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Waterford 3 SES EAL Basis Document Revision A 5.2 Abbreviations/Acronyms

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

° ........................................................................................................................... Degrees AC........................................................................................................ Alternating Current AFW................................................................................................... Auxiliary Feedwater ARM.............................................................................................. Area Radiation Monitor ARP ............................................................................. Annunciator Response Procedure ATWS ...................................................................... Anticipated Transient Without Scram CAS ................................................................................................. Central Alarm Station CCW ........................................................................................ Component Cooling Water CDE ....................................................................................... Committed Dose Equivalent CE............................................................................................... Combustion Engineering CEOG .................................................................Combustion Engineering Owners Group CET ............................................................................................ Core Exit Thermocouple CFR ..................................................................................... Code of Federal Regulations CNB ................................................................................................... Containment Barrier CPM..................................................................................................... Counts Per Minute CR ............................................................................................................... Control Room CTMT............................................................................................................. Containment DC ............................................................................................................... Direct Current DEF ..................................................................................................................... Defueled DBA ............................................................................................... Design Basis Accident DC ............................................................................................................... Direct Current D/G ......................................................................................................... Diesel Generator DRT .............................................................................................. Diversified Reactor Trip EAL ............................................................................................. Emergency Action Level ECCS............................................................................ Emergency Core Cooling System ECL.................................................................................. Emergency Classification Level ED...................................................................................................... Emergency Director EPIP ................................................................ Emergency Plan Implementing Procedure ele. ...................................................................................................................... Elevation EOF .................................................................................. Emergency Operations Facility EOP ............................................................................... Emergency Operating Procedure EPA .............................................................................. Environmental Protection Agency EPM .........................................................................................Emergency Plant Manager ERO ...........................................................................Emergency Response Organization Page 20 of 225

Waterford 3 SES EAL Basis Document Revision A ESF......................................................................................... Engineered Safety Feature FAA.................................................................................. Federal Aviation Administration FBI ................................................................................... Federal Bureau of Investigation FC ........................................................................................................... Fuel Clad Barrier FEMA............................................................... Federal Emergency Management Agency FHB ............................................................................................... Fuel Handling Building FSAR .................................................................................... Final Safety Analysis Report ft. ................................................................................................................................ Feet FTS ......................................................................................... Federal Telephone System GE ..................................................................................................... General Emergency HJTC ............................................................................... Heated Junction Thermocouple HPSI .................................................................................. High Pressure Safety Injection hr. .............................................................................................................................. Hour IC ..........................................................................................................Initiating Condition in. ................................................................................................................................ Inch IPEEE ................. Individual Plant Examination of External Events (Generic Letter 88-20)

ISFSI............................................................ Independent Spent Fuel Storage Installation Keff ......................................................................... Effective Neutron Multiplication Factor kV .......................................................................................................................... Kilovolt LCO .................................................................................. Limiting Condition of Operation LER................................................................................................ Licensee Event Report LOCA ......................................................................................... Loss of Coolant Accident LWR................................................................................................... Light Water Reactor MPC................................... Maximum Permissible Concentration/Multi-Purpose Canister mR, mRem, mrem, mREM .............................................. milli-Roentgen Equivalent Man MSIV ....................................................................................... Main Steam Isolation Valve MSL ........................................................................... Main Steam Line / Mean Sea Level MW .................................................................................................................... Megawatt NEI............................................................................................... Nuclear Energy Institute NEIC ................................................................... National Earthquake Information Center NESP ................................................................... National Environmental Studies Project NORAD................................................... North American Aerospace Defense Command (NO)UE ................................................................................ Notification of Unusual Event NPP .................................................................................................. Nuclear Power Plant NRC ................................................................................ Nuclear Regulatory Commission NSSS ................................................................................ Nuclear Steam Supply System Page 21 of 225

Waterford 3 SES EAL Basis Document Revision A OBE ...................................................................................... Operating Basis Earthquake OCA ............................................................................................... Owner Controlled Area ODCM............................................................................ Off-site Dose Calculation Manual OI ...................................................................................................... Operating Instruction OP .................................................................................................... Operating Procedure ORO ................................................................................. Offsite Response Organization OTCC ................................................................................... Once Through Core Cooling PA .............................................................................................................. Protected Area PIG ............................................................................... Particulate, Iodine, Gas (monitor)

PAG ........................................................................................ Protective Action Guideline PRA/PSA ..................... Probabilistic Risk Assessment / Probabilistic Safety Assessment PWR ....................................................................................... Pressurized Water Reactor PSIA ....................................................................... Pounds per Square Inch Atmosphere PTS........................................................................................ Pressurized Thermal Shock P/T ............................................................................................... Pressure - Temperature R ........................................................................................................................ Roentgen RAB .......................................................................................... Reactor Auxiliary Building RCB ................................................................................ Reactor Coolant System Barrier RCP ............................................................................................... Reactor Coolant Pump RCS ............................................................................................ Reactor Coolant System Rem, rem, REM ....................................................................... Roentgen Equivalent Man RHR ............................................................................................. Residual Heat Removal RMS..................................................................................... Radiation Monitoring System RPS ........................................................................................ Reactor Protection System RPV ........................................................................................... Reactor Pressure Vessel RVLMS .............................................................. Reactor Vessel Level Monitoring System SAR ............................................................................................... Safety Analysis Report SBO ......................................................................................................... Station Blackout SCBA ....................................................................... Self-Contained Breathing Apparatus SFP.......................................................................................................... Spent Fuel Pool S/G .........................................................................................................Steam Generator SI .............................................................................................................. Safety Injection SOCA .............................................................................. Security Owner Controlled Area SRO ............................................................................................ Senior Reactor Operator SSE ....................................................................................... Safe Shutdown Earthquake SWS ............................................................................................... Service Water System Page 22 of 225

Waterford 3 SES EAL Basis Document Revision A TED .................................................................................... Temporary Emergency Diesel TEDE ............................................................................... Total Effective Dose Equivalent TOAF .................................................................................................... Top of Active Fuel TSC .......................................................................................... Technical Support Center UFSAR ...................................................................Updated Final Safety Analysis Report USGS ............................................................................ United States Geological Survey UHS ..................................................................................................... Ultimate Heat Sink USGS ............................................................................ United States Geological Survey VBS ............................................................................................... Vehicle Barrier System WF3 ............................................................................ Waterford 3 Steam Electric Station Page 23 of 225

Waterford 3 SES EAL Basis Document Revision A 6.0 WF3-TO-NEI 99-01 Rev. 6 EAL CROSS-REFERENCE This cross-reference is provided to facilitate association and location of a WF3 EAL within the NEI 99-01 IC/EAL identification scheme. Further information regarding the development of the WF3 EALs based on the NEI guidance can be found in the EAL Comparison Matrix.

WF3 NEI 99-01 Rev. 6 Example EAL IC 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 Page 24 of 225

Waterford 3 SES EAL Basis Document Revision A WF3 NEI 99-01 Rev. 6 Example EAL IC EAL 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 CG1.2 CG1 2 EU1.1 E-HU1 1 FA1.1 FA1 1 FS1.1 FS1 1 FG1.1 FG1 1 HU1.1 HU1 1, 2, 3 Page 25 of 225

Waterford 3 SES EAL Basis Document Revision A WF3 NEI 99-01 Rev. 6 Example EAL IC EAL 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 2 SU5.1 SU4 1, 2, 3 SU6.1 SU5 1 Page 26 of 225

Waterford 3 SES EAL Basis Document Revision A WF3 NEI 99-01 Rev. 6 Example EAL IC EAL 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 Page 27 of 225

Waterford 3 SES EAL Basis Document Revision A 7.0 ATTACHMENTS 7.1 Attachment 1, Emergency Action Level Technical Bases 7.2 Attachment 2, Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases Page 28 of 225

Waterford 3 SES EAL Basis Document Revision A 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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Waterford 3 SES EAL Basis Document Revision A 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 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 A-1 Effluent Monitor Classification Thresholds Release Point Monitor GE SAE Alert UE 4.01 E+08 4.01 E+07 4.01 E+06 2.27 E+05 Plant Stack WRGM PRM-IRE-0110-4 Gaseous Ci/sec Ci/sec Ci/sec Ci/sec Fuel Handling Bldg. 1.48 E+10 1.48 E+09 1.48 E+08 2.27 E+05 PRM-IRE-3032-4 Exhaust WRGM Ci/sec Ci/sec Ci/sec Ci/sec Circulating Water 7.27 E-04 PRM-IRE-1900 N/A N/A N/A Discharge Monitor Ci/ml Liquid Waste 2.40 E-03 Liquid Management PRM-IRE-0647 N/A N/A N/A Discharge Monitor Ci/ml Boron Management 2.40 E-03 PRM-IRE-0627 N/A N/A N/A Discharge Monitor Ci/ml Page 30 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases 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 indicators operability, the conditions existence, or the reports 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 (ref. 1) for an extended period of time (e.g., an uncontrolled release). It includes any gaseous or liquid radiological release, monitored or un-monitored, including those for which a radioactivity discharge permit is normally prepared.

Nuclear power plants incorporate design features intended to control the release of radioactive effluents to the environment. Further, there are administrative controls established to prevent unintentional releases, and to control and monitor intentional releases. The occurrence of an extended, uncontrolled radioactive release to the environment is indicative of degradation in these features and/or controls.

Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.

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

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

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

Page 31 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases Escalation of the emergency classification level would be via IC AA1.

Reference(s):

1. UNT-005-014 Offsite Dose Calculation Manual

2. EC86890, EP-CALC-WF3-1701, Radiological Effluent EAL Values

3. NEI 99-01 AU1 Page 32 of 225

Waterford 3 SES EAL Basis Document Revision A 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.

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

Page 33 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases 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. UNT-005-014 Offsite Dose Calculation Manual

2. NEI 99-01 AU1 Page 34 of 225

Waterford 3 SES EAL Basis Document Revision A 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 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 A-1 Effluent Monitor Classification Thresholds Release Point Monitor GE SAE Alert UE 4.01 E+08 4.01 E+07 4.01 E+06 2.27 E+05 Plant Stack WRGM PRM-IRE-0110-4 Gaseous Ci/sec Ci/sec Ci/sec Ci/sec Fuel Handling Bldg. 1.48 E+10 1.48 E+09 1.48 E+08 2.27 E+05 PRM-IRE-3032-4 Exhaust WRGM Ci/sec Ci/sec Ci/sec Ci/sec Circulating Water 7.27 E-04 PRM-IRE-1900 N/A N/A N/A Discharge Monitor Ci/ml Liquid Waste 2.40 E-03 Liquid Management PRM-IRE-0647 N/A N/A N/A Discharge Monitor Ci/ml Boron Management 2.40 E-03 PRM-IRE-0627 N/A N/A N/A Discharge Monitor Ci/ml Page 35 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases 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 indicators operability, the conditions existence, or the reports 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.

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. EC86890, EP-CALC-WF3-1701, Radiological Effluent EAL Values

2. NEI 99-01 AA1 Page 36 of 225

Waterford 3 SES EAL Basis Document Revision A 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 - The border of the Exclusion Area or an area corresponding to a distance of 914 meters from the Waterford 3 reactor. Also referred to as the Exclusion Area Boundary.

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. EP-002-050 Offsite Dose Assessment

2. NEI 99-01 AA1 Page 37 of 225

Waterford 3 SES EAL Basis Document Revision A 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 - The border of the Exclusion Area or an area corresponding to a distance of 914 meters from the Waterford 3 reactor. Also referred to as the Exclusion Area Boundary.

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. 1)

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

Page 38 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases Reference(s):

1. UNT-005-014 Offsite Dose Calculation Manual

2. NEI 99-01 AA1 Page 39 of 225

Waterford 3 SES EAL Basis Document Revision A 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 - The border of the Exclusion Area or an area corresponding to a distance of 914 meters from the Waterford 3 reactor. Also referred to as the Exclusion Area Boundary.

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.

Page 40 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases Reference(s):

1. EP-002-060 Radiological Field Monitoring

2. NEI 99-01 AA1 Page 41 of 225

Waterford 3 SES EAL Basis Document Revision A 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 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 A-1 Effluent Monitor Classification Thresholds Release Point Monitor GE SAE Alert UE 4.01 E+08 4.01 E+07 4.01 E+06 2.27 E+05 Plant Stack WRGM PRM-IRE-0110-4 Gaseous Ci/sec Ci/sec Ci/sec Ci/sec Fuel Handling Bldg. 1.48 E+10 1.48 E+09 1.48 E+08 2.27 E+05 PRM-IRE-3032-4 Exhaust WRGM Ci/sec Ci/sec Ci/sec Ci/sec Circulating Water 7.27 E-04 PRM-IRE-1900 N/A N/A N/A Discharge Monitor Ci/ml Liquid Waste 2.40 E-03 Liquid Management PRM-IRE-0647 N/A N/A N/A Discharge Monitor Ci/ml Boron Management 2.40 E-03 PRM-IRE-0627 N/A N/A N/A Discharge Monitor Ci/ml Mode Applicability:

All Page 42 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases 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 indicators operability, the conditions existence, or the reports 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.

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. EC86890, EP-CALC-WF3-1701, Radiological Effluent EAL Values

2. NEI 99-01 AS1 Page 43 of 225

Waterford 3 SES EAL Basis Document Revision A 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 - The border of the Exclusion Area or an area corresponding to a distance of 914 meters from the Waterford 3 reactor. Also referred to as the Exclusion Area Boundary.

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. EP-002-050 Offsite Dose Assessment

2. NEI 99-01 AS1 Page 44 of 225

Waterford 3 SES EAL Basis Document Revision A 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 - The border of the Exclusion Area or an area corresponding to a distance of 914 meters from the Waterford 3 reactor. Also referred to as the Exclusion Area Boundary.

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.

Page 45 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases Reference(s):

1. EP-002-060 Radiological Field Monitoring

2. NEI 99-01 AS1 Page 46 of 225

Waterford 3 SES EAL Basis Document Revision A 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 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.

Page 47 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases Table A-1 Effluent Monitor Classification Thresholds Release Point Monitor GE SAE Alert UE 4.01 E+08 4.01 E+07 4.01 E+06 2.27 E+05 Plant Stack WRGM PRM-IRE-0110-4 Gaseous Ci/sec Ci/sec Ci/sec Ci/sec Fuel Handling Bldg. 1.48 E+10 1.48 E+09 1.48 E+08 2.27 E+05 PRM-IRE-3032-4 Exhaust WRGM Ci/sec Ci/sec Ci/sec Ci/sec Circulating Water 7.27 E-04 PRM-IRE-1900 N/A N/A N/A Discharge Monitor Ci/ml Liquid Waste 2.40 E-03 Liquid Management PRM-IRE-0647 N/A N/A N/A Discharge Monitor Ci/ml Boron Management 2.40 E-03 PRM-IRE-0627 N/A N/A N/A Discharge Monitor Ci/ml 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 indicators operability, the conditions existence, or the reports 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.

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):

Page 48 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases

1. EC86890, EP-CALC-WF3-1701, Radiological Effluent EAL Values

2. NEI 99-01 AG1 Page 49 of 225

Waterford 3 SES EAL Basis Document Revision A 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 - The border of the Exclusion Area or an area corresponding to a distance of 914 meters from the Waterford 3 reactor. Also referred to as the Exclusion Area Boundary.

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. EP-002-050 Offsite Dose Assessment

2. NEI 99-01 AG1 Page 50 of 225

Waterford 3 SES EAL Basis Document Revision A 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 - The border of the Exclusion Area or an area corresponding to a distance of 914 meters from the Waterford 3 reactor. Also referred to as the Exclusion Area Boundary.

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. EP-002-060 Radiological Field Monitoring

2. NEI 99-01 AG1 Page 51 of 225

Waterford 3 SES EAL Basis Document Revision A 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 SFP low water level alarm or visual observation AND UNPLANNED rise in corresponding area radiation levels as indicated by any Table A-2 radiation monitor Table A-2 Irradiated Fuel Radiation Monitors ARM-IRE-5024 ARM-IRE-5025 Containment Purge Isolation Monitors ARM-IRE-5026 ARM-IRE-5027 ARM-IRE-0300.1 ARM-IRE-0300.2 FHB Area Radiation Monitors ARM-IRE-0300.3 ARM-IRE-0300.4 PRM-IRE-5107A or B FHB PIG Gas Channel 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.

Page 52 of 225

Waterford 3 SES EAL Basis Document Revision A 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. Technical Specifications 3/4.9.11 Water Level - Spent Fuel Pool

2. Technical Specifications 3/4.9.10 Water Level - Reactor Vessel

3. OP-500-008 Control Room Cabinet H Window L-2 FUEL POOL LEVEL LO

4. W3-DBD-32, Radiation Monitoring System Design Basis Document

5. NEI 99-01 AU2 Page 53 of 225

Waterford 3 SES EAL Basis Document Revision A 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. The cask confinement boundary is considered the Multi-Purpose Canister (MPC) for the HI-STORM 100 dry storage 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.

Page 54 of 225

Waterford 3 SES EAL Basis Document Revision A 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 Page 55 of 225

Waterford 3 SES EAL Basis Document Revision A 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 A-2 radiation monitor Table A-2 Irradiated Fuel Radiation Monitors ARM-IRE-5024 ARM-IRE-5025 Containment Purge Isolation Monitors ARM-IRE-5026 ARM-IRE-5027 ARM-IRE-0300.1 ARM-IRE-0300.2 FHB Area Radiation Monitors ARM-IRE-0300.3 ARM-IRE-0300.4 PRM-IRE-5107A or B FHB PIG Gas Channel 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. The cask confinement boundary is considered the Multi-Purpose Canister (MPC) for the HI-STORM 100 dry storage 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 indicators operability, the conditions existence, or the reports accuracy is removed. Implicit in this definition is the need for timely assessment.

Page 56 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases 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. W3-DBD-32, Radiation Monitoring System Design Basis Document

2. NEI 99-01 AA2 Page 57 of 225

Waterford 3 SES EAL Basis Document Revision A 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 11 ft. (Level 2) on FS-ILI-3000(3001)

Mode Applicability:

All Definition(s):

None Basis:

FS-ILI-3000(3001) read Level 2 as 11ft. This corresponds to 31 MSL and 10 feet above level

3. The level as read on the instrument should be compared to the EAL criteria without adding any additional level.

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).

Spent Fuel Pool Level indicators FS-ILI-3000 and 3001 are read on the +21 Auxiliary Building just inside from the Q-deck.

Reference(s):

1. EC 48147 Attachment 10.002 SFPI Water Levels

2. OP-901-513 Spent Fuel Cooling Malfunction

3. OP-903-001 Technical Specification Surveillance Logs

4. NEI 99-01 AA2 Page 58 of 225

Waterford 3 SES EAL Basis Document Revision A 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 1 ft. (Level 3) on FS-ILI-3000(3001)

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:

FS-ILI-3000(3001) read Level 1 as 1 ft. This corresponds to 21 MSL. The level as read on the instrument should be compared to the EAL criteria without adding any additional level.

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).

Spent Fuel Pool Level indicators FS-ILI-3000 and 3001 are read on the +21 Auxiliary Building just inside from the Q-deck.

Reference(s):

1. EC 48147 Attachment 10.002 SFPI Water Levels

2. OP-901-513 Spent Fuel Cooling Malfunction

3. OP-903-001 Technical Specification Surveillance Logs

4. NEI 99-01 AS2 Page 59 of 225

Waterford 3 SES EAL Basis Document Revision A 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 1 ft. (Level 3) on FS-ILI-3000(3001) 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:

FS-ILI-3000(3001) read Level 1 as 1 ft. This corresponds to 21 MSL. The level as read on the instrument should be compared to the EAL criteria without adding any additional level.

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).

Spent Fuel Pool Level indicators FS-ILI-3000 and 3001 are read on the +21 Auxiliary Building just inside from the Q-deck.

Reference(s):

1. EC 48147 Attachment 10.002 SFPI Water Levels

2. OP-901-513 Spent Fuel Cooling Malfunction

3. OP-903-001 Technical Specification Surveillance Logs

4. NEI 99-01 AG2 Page 60 of 225

Waterford 3 SES EAL Basis Document Revision A 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 (ARM-IRE-5001)

CAS (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 is monitored for excessive radiation by ARM-IRE-5001 (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 EAL 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.

Reference(s):

1. W3-DBD-32, Radiation Monitoring System Design Basis Document

2. NEI 99-01 AA3 Page 61 of 225

Waterford 3 SES EAL Basis Document Revision A 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 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 A-3 Safe Operation & Shutdown Rooms/Areas Room/Area Mode Turbine Building (all elevations and rooms) 1 Polisher Building (all elevations and rooms) 1

-4 RCA Letdown Valve Gallery 3

+21 RAB Switchgears A or B 3

-4 RCA Wing Area 4

-15 RCA Valve Gallery 4

-35 RCA Safeguard Rooms 4

+21 RAB Switchgears A or B 4 Mode Applicability:

1 - Power Operation, 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.

Page 62 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases 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 4.

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.

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).

Page 63 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases EAL AA3.2 mode applicability has been limited to the mode limitations of Table A-3.

Reference(s):

1. Attachment 3 Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases

2. NEI 99-01 AA3 Page 64 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases Category C - Cold Shutdown / Refueling System Malfunction EAL Group: Cold Conditions (RCS temperature 200ºF); EALs in this category are applicable only in one or more cold operating modes.

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

Cold Shutdown, 6 - Refueling, 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 Safety Bus AC Power Loss of 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 4160 VAC safety 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 125-Volt vital DC 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.

Page 65 of 225

Waterford 3 SES EAL Basis Document Revision A 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 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. OP-001-003, Reactor Coolant System Drain Down, is used to capture required RCS water levels for plant conditions using RCS Level Control Requests sheets. 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.

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.

Page 66 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases Reference(s):

1. OP-001-003 Reactor Coolant System Drain Down

2. OI-040-000 Reactor Coolant System Leakage Monitoring

3. OP-901-111 Reactor Coolant System Leak

4. NEI 99-01 CU1 Page 67 of 225

Waterford 3 SES EAL Basis Document Revision A 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 Containment Sump or Reactor Drain Tank level due to loss of RCS inventory Visual observation of UNISOLABLE RCS leakage 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 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. 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 (ref. 1, 2).

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

Page 68 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases Reference(s):

1. OI-040-000 Reactor Coolant System Leakage Monitoring

2. OP-901-111 Reactor Coolant System Leak

3. NEI 99-01 CU1 Page 69 of 225

Waterford 3 SES EAL Basis Document Revision A 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 RCS level < 13.46 ft.

Mode Applicability:

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

None Basis:

13.46 ft. RCS level is at the elevation of the hot leg centerline. This is the minimum level for operation of Shutdown Cooling (ref. 1).

This IC addresses conditions that are precursors to a loss of the ability to adequately cool irradiated fuel (i.e., a precursor to a challenge to the fuel clad barrier). This condition represents a potential substantial reduction in the level of plant safety.

For this EAL, a lowering of RCS water level below the specified level indicates that operator actions have not been successful in restoring and maintaining reactor vessel/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 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). 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.

Reference(s):

1. OP-901-131 Shutdown Cooling Malfunction

2. NEI 99-01 CA1 Page 70 of 225

Waterford 3 SES EAL Basis Document Revision A 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 Containment Sump or Reactor Drain Tank level due to 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.

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 (ref. 1, 2).

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.

Page 71 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases Reference(s):

1. OI-040-000 Reactor Coolant System Leakage Monitoring

2. OP-901-111 Reactor Coolant System Leak

3. NEI 99-01 CA1 Page 72 of 225

Waterford 3 SES EAL Basis Document Revision A 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 upper plenum 0%

Mode Applicability:

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

CONTAINMENT CLOSURE - The procedurally defined actions taken to secure containment and its associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions. As applied to WF3, all Containment Penetrations are closed as described in the requirements of Technical Specifications section 3/4.9.4 Containment Building Penetrations and OP-010-006, Outage Operations.

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 an 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.

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.

Page 73 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases Reference(s):

1. UFSAR Section 1.9A Inadequate Core Cooling Instrumentation

2. NEI 99-01 CS1 Page 74 of 225

Waterford 3 SES EAL Basis Document Revision A 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 CONTAINMENT CLOSURE established AND Representative CETs indicate superheat Mode Applicability:

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

CONTAINMENT CLOSURE - The procedurally defined actions taken to secure containment and its associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions. As applied to WF3, all Containment Penetrations are closed as described in the requirements of Technical Specifications section 3/4.9.4 Containment Building Penetrations and OP-010-006, Outage Operations.

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 an 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 parameters of EALs CS1.1 and CS1.2 reflect the fact that with CONTAINMENT CLOSURE established, there is a lower probability of a fission product release to the environment.

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

Page 75 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases Escalation of the emergency classification level would be via IC CG1 or AG1.

Superheated conditions in the core can only occur with core uncovery. Therefore superheated conditions, as indicated on representative CETs, is used as an indicator of reactor vessel level below the top of active fuel (ref. 1, 2).

Reference(s):

1. UFSAR Section 1.9A Inadequate Core Cooling Instrumentation

2. TG-OP-902-002 Technical Guide for Loss of Coolant Accident Recovery Procedure.

3. NEI 99-01 CS1 Page 76 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - RCS Level Initiating Condition: Loss of RCS inventory affecting core decay heat removal capability EAL:

CS1.3 Site Area Emergency RCS level cannot be monitored for 30 min. (Note 1)

AND Core uncovery is indicated by any of the following:

UNPLANNED rise in Containment Sump or Reactor Drain Tank level of sufficient magnitude to indicate core uncovery Visual observation of UNISOLABLE RCS leakage of sufficient magnitude to indicate core uncovery Containment High Range Radiation Monitor (ARM-IRE-5400AS or BS) > 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.

Mode Applicability:

5 - Cold Shutdown, 6 - Refueling 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.

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 a significant and prolonged loss of RCS inventory control and makeup capability leading to IMMINENT fuel damage. The lost inventory may be due to an 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.

Page 77 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases 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 ARM-IRE-5400AS and BS 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. OI-040-000 Reactor Coolant System Leakage Monitoring

2. OP-901-111 Reactor Coolant System Leak

3. W3-DBD-32, Radiation Monitoring System Design Basis Document

4. NEI 99-01 CS1 Page 78 of 225

Waterford 3 SES EAL Basis Document Revision A 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 Representative CETs indicate superheat for 30 min. (Note 1)

AND Any Containment Challenge indication, Table C-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 6: If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, declaration of a General Emergency is not required.

Table C-1 Containment Challenge Indications CONTAINMENT CLOSURE not established (Note 6)

Containment hydrogen concentration > 4%

UNPLANNED rise in containment pressure Mode Applicability:

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

CONTAINMENT CLOSURE - The procedurally defined actions taken to secure containment and its associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions. As applied to WF3, all Containment Penetrations are closed as described in the requirements of Technical Specifications section 3/4.9.4 Containment Building Penetrations and OP-010-006, Outage Operations.

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.

Page 79 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases Basis:

This IC addresses the inability to restore and maintain reactor vessel level above the top of active fuel with containment challenged. This condition represents actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity.

Releases can be reasonably expected to exceed EPA PAG exposure levels offsite for more than the immediate site area.

Following an extended loss of core decay heat removal and inventory makeup, decay heat will cause reactor coolant boiling and a further reduction in reactor vessel level. If 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.

Superheated conditions in the core can only occur with core uncovery. Therefore superheated conditions, as indicated on representative CETs, is used as an indicator of reactor vessel level below the top of active fuel (ref. 1).

Reference(s):

1. TG-OP-902-002 Technical Guide for Loss of Coolant Accident Recovery Procedure.

2. Technical Specifications section 3/4.9.4 Containment Building Penetrations

3. UFSAR Section 6.2.5.5 Instrumentation Requirements

4. NEI 99-01 CG1 Page 80 of 225

Waterford 3 SES EAL Basis Document Revision A 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 Containment Sump or Reactor Drain Tank level of sufficient magnitude to indicate core uncovery Visual observation of UNISOLABLE RCS leakage of sufficient magnitude to indicate core uncovery Containment High Range Radiation Monitor (ARM-IRE-5400AS or BS) > 10 R/hr Erratic Source Range Monitor indication AND Any Containment Challenge indication, Table C-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 6: If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, declaration of a General Emergency is not required.

Table C-1 Containment Challenge Indications CONTAINMENT CLOSURE not established (Note 6)

Containment hydrogen concentration > 4%

UNPLANNED rise in containment pressure Mode Applicability:

5 - Cold Shutdown, 6 - Refueling Page 81 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases Definition(s):

CONTAINMENT CLOSURE - The procedurally defined actions taken to secure containment and its associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions. As applied to WF3, all Containment Penetrations are closed as described in the requirements of Technical Specifications section 3/4.9.4 Containment Building Penetrations and OP-010-006, Outage Operations.

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.

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 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.

Page 82 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases 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.

Containment High Range Radiation Monitors ARM-IRE-5400AS and BS 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. OI-040-000 Reactor Coolant System Leakage Monitoring

2. OP-901-111 Reactor Coolant System Leak

3. W3-DBD-32, Radiation Monitoring System Design Basis Document

4. Technical Specifications section 3/4.9.4 Containment Building Penetrations

5. UFSAR Section 6.2.5.5 Instrumentation Requirements

6. NEI 99-01 CG1 Page 83 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 2 - Loss of Safety Bus AC Power Initiating Condition: Loss of all but one AC power source to safety buses for 15 minutes or longer EAL:

CU2.1 Unusual Event AC power capability, Table C-2, to 4160 VAC safety buses 3A and 3B 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 C-2 AC Power Sources Emergency Diesel Generator A Onsite Emergency Diesel Generator B Temporary Emergency Diesels (TEDs)

(if already aligned)

Startup Transformer 3A Startup Transformer 3B Offsite Unit Auxiliary Transformer 3A (when back-fed from offsite)

Unit Auxiliary Transformer 3B (when back-fed from offsite)

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; Page 84 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases (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.

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

A loss of all offsite power with a concurrent failure of one emergency power source (e.g., an onsite diesel generator).

A loss of all offsite power and loss of all emergency power sources (e.g., onsite diesel generators) with a single train of safety buses being fed from TEDs.

A loss of emergency power sources (e.g., onsite diesel generators) with a train of safety buses being fed from a single offsite power source (Startup Transformer).

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.

Temporary Emergency Diesels (TEDs) can be credited if already installed in accordance with site procedures (ref. 5, 6).

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

Reference(s):

1. UFSAR Section 8.1, Onsite Power System

2. UFSAR Section 8.2, Offsite Power System

3. OP-901-310 Loss of Train A Safety Bus

4. OP-901-311 Loss of Train B Safety Bus

5. Technical Specifications 3/4.8.1 A.C. Sources

6. ME-001-012 Temporary Power from Temporary Diesel for 3A2 and 3B2 4KV Buses (Modes 1-6).

7. NEI 99-01 CU2 Page 85 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 2 - Loss of Safety Bus AC Power Initiating Condition: Loss of all offsite and all onsite AC power to safety buses for 15 minutes or longer EAL:

CA2.1 Alert Loss of all offsite and all onsite AC power to 4160 VAC safety buses 3A and 3B 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:

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 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.

Page 86 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases 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 safety 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.

Temporary Emergency Diesels (TEDs) can be credited if already installed in accordance with site procedures (ref. 5, 6).

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

Reference(s):

1. UFSAR Section 8.1, Onsite Power System

2. UFSAR Section 8.2, Offsite Power System

3. OP-901-310 Loss of Train A Safety Bus

4. OP-901-311 Loss of Train B Safety Bus

5. Technical Specifications 3/4.8.1 A.C. Sources

6. ME-001-012, Temporary Power from Temporary Diesel for 3A2 and 3B2 4KV Buses (Modes 1-6)

7. NEI 99-01 CU2 Page 87 of 225

Waterford 3 SES EAL Basis Document Revision A 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 procedurally defined actions taken to secure containment and its associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions. As applied to WF3, all Containment Penetrations are closed as described in the requirements of Technical Specifications section 3/4.9.4 Containment Building Penetrations and OP-010-006, Outage Operations.

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.

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.

Page 88 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases Reference(s):

1. Technical Specifications Table 1.2

2. OP-901-131 Shutdown Cooling Malfunction

3. NEI 99-01 CU3 Page 89 of 225

Waterford 3 SES EAL Basis Document Revision A 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 procedurally defined actions taken to secure containment and its associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions. As applied to WF3, all Containment Penetrations are closed as described in the requirements of Technical Specifications section 3/4.9.4 Containment Building Penetrations and OP-010-006, Outage Operations.

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.

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. Technical Specifications Table 1.2

2. OP-901-131 Shutdown Cooling Malfunction

3. OP-001-005 RCS Drain and Fill Below RCS Hot Leg Centerline

4. NEI 99-01 CU3 Page 90 of 225

Waterford 3 SES EAL Basis Document Revision A 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 C-3 duration (Note 1)

OR UNPLANNED RCS pressure rise > 10 psia (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 C-3 RCS Heat-up Duration Thresholds CONTAINMENT RCS Status Heat-up Duration CLOSURE Status Intact (but not lowered N/A 60 min.*

inventory) (< 20 ft MSL)

Not intact established 20 min.*

OR lowered inventory (< 20 ft MSL) not established 0 min.

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

Mode Applicability:

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

CONTAINMENT CLOSURE - The procedurally defined actions taken to secure containment and its associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions. As applied to WF3, all Containment Penetrations are closed as described in the requirements of Technical Specifications section 3/4.9.4 Containment Building Penetrations and OP-010-006, Outage Operations.

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

Page 91 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases 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 (ref. 2).

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). 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. Technical Specifications Table 1.2

2. OP-901-131 Shutdown Cooling Malfunction

3. OP-001-003 Reactor Coolant System Drain Down

4. NEI 99-01 CA3 Page 92 of 225

Waterford 3 SES EAL Basis Document Revision A 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 < 108 VDC on required vital DC 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 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.

As used in this EAL, required means the 3A-DC, 3B-DC or 3AB-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.

Page 93 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases Depending upon the event, escalation of the emergency classification level would be via IC CA1 or CA3, or an IC in Recognition Category A.

Less than 108 VDC bus voltage is based on the minimum bus voltage necessary for the operation of safety related equipment (ref. 1, 2, 3). Technical Specifications section 3/4.8.2.2 D.C. Sources - Shutdown, specifies required DC buses (associated station batteries and battery chargers) while in cold conditions (ref. 4).

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

Reference(s):

1. ECE91-058 Battery 3A-S A Train Calculation for Station Blackout

2. ECE91-059 Battery 3B-S B Train Calculation for Station Blackout

3. ECE91-060 Battery 3AB-S Calculation for Station Blackout

4. Technical Specifications section 3/4.8.2.2 D.C. Sources - Shutdown

5. NEI 99-01 CU4 Page 94 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 5 - Loss of Communications Initiating Condition: Loss of all onsite or offsite communications capabilities EAL:

CU5.1 Unusual Event Loss of all Table C-4 onsite communication methods OR Loss of all Table C-4 State and local agency communication methods OR Loss of all Table C-4 NRC communication methods Table C-4 Communication Methods Onsite State/ NRC System Local Telephone System X X X Operational Hotline X Plant Radio System (O&M) X Plant Paging System X Sound Powered Phone System X Civil Defense Radio System X Satellite Phones X X Emergency Notification System (ENS) X Mode Applicability:

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

None Page 95 of 225

Waterford 3 SES EAL Basis Document Revision A 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 St. Charles Parish Department of Homeland Security and Emergency Preparedness, St. Charles Parish Sheriffs Department 911 Center, St. John the Baptist Parish Office of Emergency Preparedness, St. John the Baptist Parish Sheriffs Department 911 Center, Louisiana Department of Environmental Quality and the Louisiana Governor's Office of Homeland Security and Emergency Preparedness. The third EAL condition 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. EP-003-070 Emergency Communications Systems

2. Waterford 3 SES Emergency Plan Section 7.5 Communications Systems

3. NEI 99-01 CU5 Page 96 of 225

Waterford 3 SES EAL Basis Document Revision A 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 C-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 C-5 Hazardous Events Seismic event (earthquake)

Internal or external FLOODING event High winds or tornado strike FIRE EXPLOSION Other events with similar hazard characteristics as determined by the Shift Manager Page 97 of 225

Waterford 3 SES EAL Basis Document Revision A 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 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.

Page 98 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases 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 Page 99 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases 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 WF3 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.

Page 100 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases Category: E - ISFSI Subcategory: ISFSI 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 > any Table E-1 dose rate limit.

Table E-1 ISFSI Dose Rate Limits HI-STORM (Note E) 126 mrem/hr Total Dose Rate Point A 338 mrem/hr Total Dose Rate Point B 326 mrem/hr Total Dose Rate Point C 60 mrem/hr Total Dose Rate Point D 60 mrem/hr Total Dose Rate Point E 374 mrem/hr Total Dose Rate Point F 136 mrem/hr Total Dose Rate Point G NOTE E: Survey points are described in DFS-007-003, Radiation Monitoring Requirements for Loading and Storage HI-STORM.

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. The cask confinement boundary is considered the Multi-Purpose Canister (MPC) for the HI-STORM 100 dry storage 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.

Page 101 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases Basis:

This IC addresses an event that results in damage to the CONFINEMENT BOUNDARY of a storage cask containing spent fuel. It applies to irradiated fuel that is licensed for dry storage beginning at the point that the loaded storage cask is sealed. The issues of concern are the creation of a potential or actual release path to the environment, degradation of one or more fuel assemblies due to environmental factors, and configuration changes which could cause challenges in removing the cask or fuel from storage.

The existence of damage is determined by radiological survey. The 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. UFSAR Section 9.1.5 Spent Fuel Dry Cask Storage

2. Holtec International Final Safety Analysis Report for the HI-STORM 100 Cask System -

Holtec Report No.: HI-2002444

3. DFS-007-003 Radiation Monitoring Requirements for Loading and Storage HI-STORM

4. NEI 99-01 E-HU1 Page 102 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases Category F - Fission Product Barrier Degradation EAL Group: Hot Conditions (RCS temperature > 200ºF); 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 containment building and connections up to and including the outermost containment isolation valves. This barrier also includes the main steam, feedwater, and blowdown line extensions outside the containment building up to and including the outermost secondary side isolation valve. Containment Barrier thresholds are used as criteria for escalation of the 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 Page 103 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases 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 WF3 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.

Page 104 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases Category: F - Fission Product Barrier Degradation Subcategory: N/A Initiating Condition: Any loss or any potential loss of either Fuel Clad or RCS barrier EAL:

FA1.1 Alert Any loss or any potential loss of either Fuel Clad or RCS barrier (Table F-1)

Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):

None Basis:

Fuel Clad, RCS and Containment comprise the fission product barriers. Table F-1 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 Page 105 of 225

Waterford 3 SES EAL Basis Document Revision A 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 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 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 (vice Loss), the Emergency Director would have greater assurance that escalation to a General Emergency is less IMMINENT.

Reference(s):

1. NEI 99-01 FS1 Page 106 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases Category: F - Fission Product Barrier Degradation Subcategory: N/A Initiating Condition: Loss of any two barriers and loss or potential loss of the third barrier EAL:

FG1.1 General Emergency Loss of any two barriers AND Loss or potential loss of the third barrier (Table F-1)

Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):

None Basis:

Fuel Clad, RCS and Containment comprise the fission product barriers. Table F-1 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 Page 107 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases Table F-1 Fission Product Barrier Threshold Matrix & Bases Table F-1 lists the threshold conditions that define the Loss and Potential Loss of the three fission product barriers (Fuel Clad, Reactor Coolant System, and Containment). The table is structured so that each of the three barriers occupies adjacent columns. Each fission product barrier column is further divided into two columns; one for Loss thresholds and one for Potential Loss thresholds.

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

A. RCS or S/G Tube Leakage B. Inadequate Heat removal C. Containment Radiation / RCS Activity D. Containment Integrity or Bypass E. Emergency Director Jugement Each category occupies a row in Table F-1 thus forming a matrix defined by the categories.

The intersection of each row with each Loss/Potential Loss column forms a cell in which one or more fission product barrier thresholds appear. If NEI 99-01 does not define a threshold for a barrier Loss/Potential Loss, the word None is entered in the cell.

Thresholds are assigned sequential numbers within each barrier column beginning with number one (ex., FCB1, FCB2FCB9).

If a cell in Table F-1 contains more than one numbered threshold, each of the numbered thresholds, if exceeded, signifies a Loss or Potential Loss of the barrier. It is not necessary to exceed all of the thresholds in a category before declaring a barrier Loss/Potential Loss.

Subdivision of Table F-1 by category facilitates association of plant conditions to the applicable fission product barrier Loss and Potential Loss thresholds. This structure promotes a systematic approach to assessing the classification status of the fission product barriers.

When equipped with knowledge of plant conditions related to the fission product barriers, the EAL-user first scans down the category column of Table F-1, locates the likely category and then reads across the fission product barrier Loss and Potential Loss thresholds in that category to determine if a threshold has been exceeded. If a threshold has not been exceeded, the EAL-user proceeds to the next likely category and continues review of the thresholds in the new category If the EAL-user determines that any threshold has been exceeded, by definition, the barrier is lost or potentially lost - even if multiple thresholds in the same barrier column are exceeded, only that one barrier is lost or potentially lost. The EAL-user must examine each of the three fission product barriers to determine if other barrier thresholds in the category are lost or potentially lost. For example, if containment radiation is sufficiently high, a Loss of the Fuel Clad and RCS Barriers and a Potential Loss of the Containment Barrier can occur. Barrier 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.

Page 108 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases 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.

Page 109 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases Table 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 RCB2 UNISOLABLE RCS leakage or S/G tube leakage > 44 gpm excluding normal reductions in RCS inventory (e.g.,

letdown, RCP seal leakoff)

A RCB1 An automatic or manual ECCS (SIAS) CNB1 S/G tube leakage > 44 gpm actuation required by EITHER: (excluding normal reductions in RCB3 RCS cooldown rate > 100°F/hr RCS or S/G None FCB1 RVLMS upper plenum level 0% RCS inventory) or that is None UNISOLABLE RCS leakage AND RUPTURED is also FAULTED Tube S/G tube RUPTURE Pressurizer pressure > maximum limits outside of containment Leakage of the RCS Pressure and Temperature Limits (OP-902-009 Attachments 2-A thru D)

FCB3 Representative CET readings > 700°F CNB2 Representative CET readings RCB4 Any OP-902-008 Functional Recovery > 1,200°F B FCB2 Representative CET readings FCB4 Any OP-902-008 Functional Recovery RCS/Core Heat Removal safety AND None None Inadequate > 1,200°F RCS/Core Heat Removal safety function criterion is not met for Restoration procedures not function criterion is not met for 15 min. (Note 1) effective within 15 min.

Heat Removal 15 min. (Note 1) (Note 1)

FCB5 Containment High Range C Radiation Monitor (ARM-IRE-5400AS or ARM-IRE-5400BS)

CTMT RCB5 Containment High Range Radiation CNB3 Containment High Range Radiation

> 900 R/hr. (Note 14)

None Monitor (ARM-IRE-5400AS or ARM- None None Monitor (ARM-IRE-5400AS or ARM-Radiation / IRE-5400BS) > 60 R/hr. (Note 14) IRE-5400BS) > 15,000 R/hr. (Note 14)

FCB6 Reactor coolant activity > 300 RCS Ci/gm dose equivalent I-131 as Activity indicated by Chemistry sample CNB4 Containment isolation is required AND EITHER:

CNB6 Containment pressure > 50 psia Containment integrity has D been lost based on CNB7 Containment hydrogen concentration Emergency Director judgment > 4%

CTMT None None None None CNB8 Containment pressure > 17.7 psia with UNISOLABLE pathway from Integrity or Containment to the < one full train of containment heat environment exists removal systems operating per design Bypass for 15 min. (Notes 1, 9)

CNB5 Indications of RCS leakage outside of Containment E FCB7 Any condition in the opinion of FCB8 Any condition in the opinion of the RCB6 Any condition in the opinion of the RCB7 Any condition in the opinion of the CNB9 Any condition in the opinion of the CNB10 Any condition in the opinion of the the Emergency Director that Emergency Director that indicates Emergency Director that indicates Emergency Emergency Director that indicates Emergency Director that indicates Emergency Director that indicates indicates loss of the Fuel Clad potential loss of the Fuel Clad potential loss of the Containment loss of the RCS barrier potential loss of the RCS barrier loss of the Containment barrier Director barrier barrier barrier Judgment Page 110 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases Barrier: Fuel Clad Category: A. RCS or S/G Tube Leakage Degradation Threat: Loss Threshold:

None Page 111 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases Barrier: Fuel Clad Category: A. RCS or S/G Tube Leakage Degradation Threat: Potential Loss Threshold:

FCB1 RVLMS upper plenum level 0%

Definition(s):

None Basis:

This reading indicates a reduction in reactor vessel water level sufficient to allow the onset of heat-induced cladding damage.

The closest indication of level near top of active fuel is provided by the RVLMS 0% sensor (#8),

~12.6 in. above the fuel alignment plate (ref. 1).

RVLMS Sensor #8 (RVLMS upper plenum level 0%) may briefly indicate voided during a normal response to a loss of inventory. Existing guidance for classifying transient events such as this addresses the period of time of event recognition and classification (15 minutes).

However, in cases when EAL declaration criteria may be met momentarily during the normal expected response of the plant, declaration requirements should not be considered to be met when the conditions are a part of the designed plant response, or result from appropriate operator actions (ref. 2, 3).

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

Reference(s):

1. UFSAR Section 1.9A Inadequate Core Cooling Instrumentation

2. SAMG-03A Severe Accident Management Guidelines Candidate High Level Action Implementation and Assessment Waterford Unit 3

3. CEOG Generic Accident Management Guidelines - Phase 1, Initial Diagnosis

4. NEI 99-01 RCS or S/G Tube Leakage Potential Loss 1.A Page 112 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases Barrier: Fuel Clad Category: B. Inadequate Heat Removal Degradation Threat: Loss Threshold:

FCB2 Representative CET readings > 1,200°F Definition(s):

None Basis:

This reading indicates temperatures within the core are sufficient to cause significant superheating of reactor coolant.

Reference(s):

1. CEOG Generic Accident Management Guidelines

2. UFSAR Section 1.9A Inadequate Core Cooling Instrumentation

3. NEI 99-01 Inadequate Heat Removal Loss 2.A Page 113 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases Barrier: Fuel Clad Category: B. Inadequate Heat Removal Degradation Threat: Potential Loss Threshold:

FCB3 Representative CET readings > 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. CEOG Generic Accident Management Guidelines

2. UFSAR Section 1.9A Inadequate Core Cooling Instrumentation

3. NEI 99-01 Inadequate Heat Removal Potential Loss 2.A Page 114 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases Barrier: Fuel Clad Category: B. Inadequate Heat Removal Degradation Threat: Potential Loss Threshold:

FCB4 Any OP-902-008 Functional Recovery RCS/Core Heat Removal safety function criterion is not met 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.

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 threshold RCB4, meeting this threshold results in a Site Area Emergency.

Inability to remove heat from the RCS to the ultimate heat sink is a loss of function required for hot shutdown with the reactor at pressure and temperature and thus represents potential loss of the Fuel Clad and RCS barriers. The RCS/Core Heat Removal safety function criteria from OP-902-008, Functional Recovery is used for this determination (ref. 1, 2).

The process of checking the safety functions in EOPs is periodic and continuous as long as the procedure is in use. The fifteen minute interval (subject to Note 1) provides a suitable opportunity to assess plant conditions with respect to the threshold criteria.

Reference(s):

1. OP-902-008 Functional Recovery

2. TG-OP-902-008 Technical Guide for Functional Recovery

3. NEI 99-01 Inadequate Heat Removal Potential Loss 2.B Page 115 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases Barrier: Fuel Clad Category: C. CTMT Radiation / RCS Activity Degradation Threat: Loss Threshold:

FCB5 Containment High Range Radiation Monitor (ARM-IRE-5400AS or ARM-IRE-5400BS)

> 900 R/hr (Note 14).

Note 14: Evaluate Containment High Range Radiation Monitor readings for potential erratic indications as a result of thermally induced currents.

Definition(s):

None Basis:

NRC Information Notice 97-045 Supplement 1 identifies the potential for erratic indications from the Containment High Range Radiation Monitors (CHRRMs) as a result of Thermally Induced Currents (TIC) which may cause the CHRRM to read falsely high on a rapid temperature rise, and fail low intermittently on a rapid temperature fall. The TICs induced in the Waterford CHRRM signal cable are anticipated to be negligible within 5 minutes. Because of this phenomenon, any trends or alarms on the CHRRMs 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 (943 R/hr rounded to 900 R/hr for readability) corresponds to an instantaneous release of all reactor coolant mass into the containment, assuming that reactor coolant activity equals 300 Ci/gm dose equivalent I-131. Reactor coolant activity above this level is greater than that expected for iodine spikes and corresponds to approximately 1% 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 Fuel Clad barrier Potential Loss threshold associated with CTMT Radiation/RCS Activity.

Reference(s):

1. EC86890, EP-CALC-WF3-1702, Cont. High Range Radiation EAL Threshold Values

2. NRC Information Notice 97-045 Supplement 1

3. NEI 99-01 CTMT Radiation / RCS Activity FC Loss 3.A Page 116 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases Barrier: Fuel Clad Category: C. CTMT Radiation / RCS Activity Degradation Threat: Loss Threshold:

FCB6 Reactor coolant activity > 300 Ci/gm dose equivalent I-131 as indicated by Chemistry sample Definition(s):

None Basis:

This threshold indicates that RCS radioactivity concentration is greater than 300 Ci/gm dose equivalent I-131. Reactor coolant activity above this level is greater than that expected for iodine spikes and corresponds to approximately 1% 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 Fuel Clad barrier Potential Loss threshold associated with CTMT Radiation/RCS Activity.

Reference(s):

1. NEI 99-01 CTMT Radiation / RCS Activity Fuel Clad Loss 3.B Page 117 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases Barrier: Fuel Clad Category: C. CTMT Radiation / RCS Activity Degradation Threat: Potential Loss Threshold:

None Page 118 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases Barrier: Fuel Clad Category: D. CTMT Integrity or Bypass Degradation Threat: Loss Threshold:

None Page 119 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases Barrier: Fuel Clad Category: D. CTMT Integrity or Bypass Degradation Threat: Potential Loss Threshold:

None Page 120 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases 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 Page 121 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases 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 Page 122 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases Barrier: Reactor Coolant System Category: A. RCS or S/G Tube Leakage Degradation Threat: Loss Threshold:

RCB1 An automatic or manual ECCS (SIAS) 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.

RUPTURE(D) - 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.

Reference(s):

1. NEI 99-01 RCS or SG Tube Leakage Reactor Coolant System Loss 1.A Page 123 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases 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 > 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 charging pump, but an ECCS (SIAS) actuation has not occurred.

Isolating letdown is a standard abnormal operating procedure action and may prevent unnecessary classifications when a non-RCS leakage path such as a CVCS leak exists. The intent of this condition is met if attempts to isolate Letdown are NOT successful. Additional charging pumps being required is indicative of a substantial RCS leak.

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. OP-901-111 Reactor Coolant System Leak

2. UFSAR Table 9.3-9 Design Parameters

3. NEI 99-01 RCS or SG Tube Leakage Reactor Coolant System Potential Loss 1.A Page 124 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases Barrier: Reactor Coolant System Category: A. RCS or S/G Tube Leakage Degradation Threat: Potential Loss Threshold:

RCB3 RCS cooldown rate > 100°F/hr AND Pressurizer pressure > maximum limits of the RCS Pressure and Temperature Limits (OP-902-009 Attachments 2-A thru D)

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 of greater than 100 F in less than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months while the RCS is in Mode 3 or higher (i.e., hot and pressurized).

Reference(s):

1. TG-OP-902-004 Technical Guide for Excess Steam Demand Recovery Procedure

2. Technical Specifications 3.4.8 Pressure/Temperature Limits

3. OP-902-008 Functional Recovery

4. OP-902-009 Standard Appendices, Attachments 2-A thru D RCS Pressure and Temperature Limits

5. NEI 99-01 RCS or SG Tube Leakage Reactor Coolant System Potential Loss 1.B Page 125 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases Barrier: Reactor Coolant System Category: B. Inadequate Heat Removal Degradation Threat: Loss Threshold:

None Page 126 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases Barrier: Reactor Coolant System Category: B. Inadequate Heat Removal Degradation Threat: Potential Loss Threshold:

RCB4 Any OP-902-008 Functional Recovery RCS/Core Heat Removal safety function criterion is not met 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.

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 threshold 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.

Inability to remove heat from the RCS to the ultimate heat sink is a loss of function required for hot shutdown with the reactor at pressure and temperature and thus represents potential loss of the Fuel Clad and RCS barriers. The RCS/Core Heat Removal safety function criteria from OP-902-008, Functional Recovery is used for this determination (ref. 1, 2).

The process of checking the safety functions in EOPs is periodic and continuous as long as the procedure is in use. The fifteen minute interval (subject to Note 1) provides a suitable opportunity to assess plant conditions with respect to the threshold criteria.

There is no RCS barrier Loss threshold associated with Inadequate Heat Removal.

Page 127 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases Reference(s):

1. OP-902-008 Functional Recovery

2. TG-OP-902-008 Technical Guide for Functional Recovery

3. NEI 99-01 Inadequate Heat Removal RCS Potential Loss 2.B Page 128 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases Barrier: Reactor Coolant System Category: C. CTMT Radiation/ RCS Activity Degradation Threat: Loss Threshold:

RCB5 Containment High Range Radiation Monitor (ARM-IRE-5400AS or ARM-IRE-5400BS)

> 60 R/hr (Note 14).

Note 14: Evaluate Containment High Range Radiation Monitor readings for potential erratic indications as a result of thermally induced currents.

Definition(s):

N/A Basis:

NRC Information Notice 97-045 Supplement 1 identifies the potential for erratic indications from the Containment High Range Radiation Monitors (CHRRMs) as a result of Thermally Induced Currents (TIC) which may cause the CHRRM to read falsely high on a rapid temperature rise, and fail low intermittently on a rapid temperature fall. The TICs induced in the Waterford CHRRM signal cable are anticipated to be negligible within 5 minutes. Because of this phenomenon, any trends or alarms on the CHRRMs 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 (61.9 R/hr rounded to 60 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 (ref. 1).

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 RCS barrier Potential Loss threshold associated with CTMT Radiation/RCS Activity.

Reference(s):

1. EC86890, EP-CALC-WF3-1702, Cont. High Range Radiation EAL Threshold Values

2. NRC Information Notice 97-045 Supplement 1

3. NEI 99-01 CTMT Radiation / RCS Activity RCS Loss 3.A Page 129 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases Barrier: Reactor Coolant System Category: B. CTMT Radiation/ RCS Activity Degradation Threat: Potential Loss Threshold:

None Page 130 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases Barrier: Reactor Coolant System Category: D. CTMT Integrity or Bypass Degradation Threat: Loss Threshold:

None Page 131 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases Barrier: Reactor Coolant System Category: D. CTMT Integrity or Bypass Degradation Threat: Potential Loss Threshold:

None Page 132 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases 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 Page 133 of 225

Waterford 3 SES EAL Basis Document Revision A 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 Page 134 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases Barrier: Containment Category: A. RCS or S/G Tube Leakage Degradation Threat: Loss Threshold:

CNB1 S/G tube leakage > 44 gpm (excluding normal 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 (S/G) 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, or to drive Emergency Feedwater Pump AB.

These conditions may exist when one steam generator is leaking or ruptured and the opposite steam generator is also faulted. These types of conditions will result in a significant and Page 135 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases 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.

Execution of a rapid RCS cooldown to less than 520°F Hot Leg temperature using both Atmospheric Dump valves, as directed by emergency operating procedures, does not meet this intent, provided the affected steam generator Atmospheric Dump valve can be closed when isolation of the RUPTURED steam generator is directed at less than 520°F. Steaming of Emergency Feedwater Pump AB prior to isolation of the RUPTURED steam generator does not meet this intent, provided MS-401 A(B) is closed when isolation of the RUPTURED steam generator is directed at less than 520°F. These short term radiological releases should be evaluated using Category A ICs.

Steam releases associated with the expected operation of a S/G Atmospheric Dump 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 S/G 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 S/G, are summarized below.

Affected S/G is FAULTED Outside of Containment?

P-to-S Leak Rate Yes No Less than or equal to 25 gpm No classification No classification Greater than 25 gpm Unusual Event per SU5.1 Unusual Event per SU5.1 Greater than 44 gpm (RCS Site Area Emergency per Alert per FA1.1 Barrier Potential Loss) FS1.1 Requires an automatic or manual Site Area Emergency per ECCS (SIAS) actuation (RCS Alert per FA1.1 FS1.1 Barrier Loss)

There is no Containment barrier 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 Page 136 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases Barrier: Containment Category: A. RCS or S/G Tube Leakage Degradation Threat: Potential Loss Threshold:

None Page 137 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases Barrier: Containment Category: B. Inadequate Heat Removal Degradation Threat: Loss Threshold:

None Page 138 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases Barrier: Containment Category: B. Inadequate Heat Removal Degradation Threat: Potential Loss Threshold:

CNB2 Representative CET readings > 1,200°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 of CET readings exceeding 1,200°F. 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 a greater 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.

There is no Containment barrier Loss threshold associated with Inadequate Heat Removal.

Page 139 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases Reference(s):

1. CEOG Generic Accident Management Guidelines

2. UFSAR Section 1.9A Inadequate Core Cooling Instrumentation

3. NEI 99-01 Inadequate Heat Removal Containment Potential Loss 2.A Page 140 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases Barrier: Containment Category: C. CTMT Radiation/RCS Activity Degradation Threat: Loss Threshold:

None Page 141 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases Barrier: Containment Category: C. CTMT Radiation/RCS Activity Degradation Threat: Potential Loss Threshold:

CNB3 Containment High Range Radiation Monitor (ARM-IRE-5400AS or ARM-IRE-5400BS)

> 15,000 R/hr (Note 14).

Note 14: Evaluate Containment High Range Radiation Monitor readings for potential erratic indications as a result of thermally induced currents.

Definition(s):

None Basis:

NRC Information Notice 97-045 Supplement 1 identifies the potential for erratic indications from the Containment High Range Radiation Monitors (CHRRMs) as a result of Thermally Induced Currents (TIC) which may cause the CHRRM to read falsely high on a rapid temperature rise, and fail low intermittently on a rapid temperature fall. The TICs induced in the Waterford CHRRM signal cable are anticipated to be negligible within 5 minutes. Because of this phenomenon, any trends or alarms on the CHRRMs 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 corresponds to an instantaneous release of all reactor coolant mass into the containment, assuming that 20% of the fuel cladding has failed (ref. 1). 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. EC86890, EP-CALC-WF3-1702, Cont. High Range Radiation EAL Threshold Values

2. NRC Information Notice 97-045 Supplement 1

3. NEI 99-01 CTMT Radiation / RCS Activity Containment Potential Loss 3.A Page 142 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases 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):

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.

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.).

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.

Page 143 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases 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.

A failure of a containment penetration will raise pressure in the Annulus. This condition will be indicated by more frequent cycling of the Shield Building Ventilation System to Exhaust Mode combined with rising radiation levels on the Plant Stack radiation monitors.

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 the second threshold would be met. Depending upon radiation monitor locations and sensitivities, this leakage could be detected by any of the four monitors depicted in the figure and cause the first threshold to be met as well.

Following the leakage of RCS mass into containment and a rise in containment pressure, there may be minor radiological releases associated with allowable containment leakage Page 144 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases 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 Page 145 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases Barrier: Containment Category: D. CTMT Integrity or Bypass Degradation Threat: Loss Threshold:

CNB5 Indications of RCS leakage outside of Containment Definition(s):

None Basis:

The status of the containment barrier during an event involving steam generator tube leakage is assessed using Loss Threshold CNB1.

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.

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.

Page 146 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases Figure 1: Containment Integrity or Bypass Examples 2nd Threshold-Airborne release from pathway Inside Reactor Effluent Auxiliary Building Monitor Building Vent Damper Filt Area Monitor Open valve Open valve Penetration Damper 1st Threshold-Airborne Airborne Monitor 2nd Threshold-reactor Open valve Open valve coolant 1st leakage Threshold- outside AB Interface leakage Airborne release from Process penetration Monitor Closed Cooling Open valve Open valve Pump RCP Seal Cooling Page 147 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases Barrier: Containment Category: D. CTMT Integrity or Bypass Degradation Threat: Potential Loss Threshold:

CNB6 Containment pressure > 50 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. Calculation ECS98-001 EOP Action value Basis Document

2. NEI 99-01 CTMT Integrity or Bypass Containment Potential Loss 4.A Page 148 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases 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. UFSAR Section 6.2.5.5 Instrumentation Requirements

2. NEI 99-01 CTMT Integrity or Bypass Containment Potential Loss 4.B Page 149 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases Barrier: Containment Category: D. CTMT Integrity or Bypass Degradation Threat: Potential Loss Threshold:

CNB8 Containment pressure > 17.7 psia with < one full train of containment heat removal systems operating per design for 15 min. (Notes 1, 9)

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 either:

One train of the Containment Spray System (operating with 1750 gpm flow) AND One train of the Containment Cooling System (one fan cooler required)

OR Two trains of the Containment Spray System (operating with 1750 gpm flow each).

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. OP-902-002 Loss of Coolant Accident Recovery

2. UFSAR Section 6.2.2 Containment Heat Removal Systems

3. Technical Specification 3.6.2 Containment Systems - Depressurization and Cooling Systems

4. NEI 99-01 CTMT Integrity or Bypass Containment Potential Loss 4.C Page 150 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases 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 Page 151 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases 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 Page 152 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases 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.

Page 153 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases 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 Security Shift Supervisor OR Notification of a credible security threat directed at the site OR A validated notification from the NRC providing information of an aircraft threat Mode Applicability:

All Definition(s):

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 WF3 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 WF3. 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 (SOCA)).

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

PROTECTED AREA - The area encompassed by physical barriers (the security fence) and to which access is controlled into the vital areas of the plant.

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):

Page 154 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases 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 area inside the SOCA Vehicle Barrier System (VBS) up to the PROTECTED AREA fence line. Access to this area is controlled by the SOCA Personnel Access Control Point. The SOCA is part of but not equal to the Owner Controlled Area as described or defined in the Waterford 3 Emergency Plan.

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 Supervisor because this is the individual 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 addresses the receipt of a credible security threat. The credibility of the threat is assessed in accordance with the WF3 Safeguards Contingency Plan (ref. 1).

The third threshold addresses the threat from the impact of an aircraft on the plant. The NRC Headquarters Operations Officer (HOO) will communicate to the licensee if the threat involves an aircraft. The status and size of the plane may also be provided by NORAD through the NRC. Validation of the threat is performed in accordance with the WF3 Safeguards Contingency Plan (ref.1).

Page 155 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases 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 Safeguards Contingency Plan for WF3 (ref. 1).

Escalation of the emergency classification level would be via IC HA1.

Reference(s):

1. WF3 Safeguards Contingency Plan

2. NEI 99-01 HU1 Page 156 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases 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 Security Shift Supervisor OR A validated notification from NRC of an aircraft attack threat within 30 min. of the site Mode Applicability:

All Definition(s):

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 WF3 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 WF3. 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 (SOCA)).

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 - The area encompassed by physical barriers (the security fence) and to which access is controlled into the vital areas of the plant.

SECURITY OWNER CONTROLLED AREA (SOCA) - The area inside the SOCA Vehicle Barrier System (VBS) up to the PROTECTED AREA fence line. Access to this area is controlled by the SOCA Personnel Access Control Point. The SOCA is part of but not equal to the Owner Controlled Area as described or defined in the Waterford 3 Emergency Plan.

Page 157 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases 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 the WF3 Safeguards Contingency Plan (ref.1).

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 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.

Page 158 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases 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 Safeguards Contingency Plan for WF3 (ref. 1).

Escalation of the emergency classification level would be via IC HS1.

Reference(s):

1. WF3 Safeguards Contingency Plan

2. NEI 99-01 HA1 Page 159 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases 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 Security Shift Supervisor 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 WF3 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 WF3. 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 (SOCA)).

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 - The area encompassed by physical barriers (the security fence) and to which access is controlled into the vital areas of the plant.

SECURITY OWNER CONTROLLED AREA (SOCA) - The area inside the SOCA Vehicle Barrier System (VBS) up to the PROTECTED AREA fence line. Access to this area is controlled by the SOCA Personnel Access Control Point. The SOCA is part of but not equal to the Owner Controlled Area as described or defined in the Waterford 3 Emergency Plan.

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.

Page 160 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and 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 Offsite Response Organization (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 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.

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 Safeguards Contingency Plan for WF3 (ref. 1).

Reference(s):

1. WF3 Safeguards Contingency Plan

2. NEI 99-01 HS1 Page 161 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 2 - Seismic Event Initiating Condition: Seismic event greater than OBE levels EAL:

HU2.1 Unusual Event Seismic event > OBE as indicated by RED light on the seismic monitor panel Mode Applicability:

All Definition(s):

None Basis:

A RED LIGHT on the Seismic Monitoring Panel indicates that the design limit for the Operating Basis Earthquake (OBE) (0.05g horizontal or 0.033g vertical) has been exceeded.

This IC addresses a seismic event that results in accelerations at the plant site greater than those specified for an Operating Basis Earthquake (OBE). An earthquake greater than an OBE but less than a Safe Shutdown Earthquake (SSE) should have no significant impact on safety-related systems, structures and components; however, some time may be required for the plant staff to ascertain the actual post-event condition of the plant (e.g., performs walk-downs and post-event inspections). Given the time necessary to perform walk-downs and inspections, and fully understand any impacts, this event represents a potential degradation of the level of safety of the plant.

Event verification with external sources should not be necessary during or following an OBE.

Earthquakes of this magnitude should be readily felt by on-site personnel and recognized as a seismic event (e.g., lateral accelerations in excess of 0.05g). The Shift Manager or Emergency Director may seek external verification if deemed appropriate (e.g., a call to the U.S. Geological Survey (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.

To avoid inappropriate emergency classification resulting from spurious actuation of the seismic instrumentation or felt motion not attributable to seismic activity, an offsite agency 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 WF3 coordinates: 29º 59' 42" north latitude, 90º 28' 16 west longitude (ref. 3). Seismic activity information can be accessed via the contact information in the Emergency Management Resources Book,Section IV, Offsite Support Organizations.

Page 162 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases Reference(s):

1. UFSAR Section 3.7.4 Seismic Instrumentation

2. OP-901-522 Seismic Event

3. UFSAR Section 1.2.1.1 Location and Population

4. NEI 99-01 HU2 Page 163 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases 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 - The area encompassed by physical barriers (the security fence) and to which access is controlled into the vital areas of the plant.

Basis:

This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant.

This EAL addresses a tornado striking (touching down) within the PROTECTED AREA.

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-901-521 Severe Weather and Flooding

2. NEI 99-01 HU3 Page 164 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases 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.

This EAL addresses FLOODING of a building room or area that results in operators isolating power to a SAFETY SYSTEM component due to water level or other wetting concerns.

Classification is also required if the water level or related wetting causes an automatic isolation of a SAFETY SYSTEM component from its power source (e.g., a breaker or relay trip). To warrant classification, operability of the affected component must be required by Technical Specifications for the current operating mode.

Escalation of the emergency classification level would be based on ICs in Recognition Categories 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):

Page 165 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases

1. UFSAR Section 3.6A.6 Flooding Analysis

2. NEI 99-01 HU3 Page 166 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases 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 PROTECTED 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 - The area encompassed by physical barriers (the security fence) and to which access is controlled into the vital areas of the plant.

Basis:

This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant.

This EAL addresses a hazardous materials event originating at a location outside the PROTECTED AREA and of sufficient magnitude to IMPEDE the movement of personnel within the PROTECTED AREA.

Escalation of the emergency classification level would be based on ICs in Recognition Categories A, F, S or C.

Reference(s):

1. NEI 99-01 HU3 Page 167 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases 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.

This EAL addresses a hazardous event that causes an on-site impediment to vehicle movement and significant enough to prohibit the plant staff from accessing the site using personal vehicles. Examples of such an event include site FLOODING caused by a hurricane, heavy rains, up-river water releases, dam failure, etc., or an on-site train derailment blocking the access road.

This EAL is not intended 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.

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 Page 168 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 4 - Fire Initiating Condition: FIRE potentially degrading the level of safety of the plant EAL:

HU4.1 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 12)

Field verification of a single fire alarm AND The FIRE is located within any Table 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 12: Bullet 2 of this EAL (multiple fire alarm indications) is not applicable when diagnosed into a LOCA or Excess Steam Demand event in Containment.

Table H-1 Fire Areas Containment Cooling Tower Areas Fuel Handling Building Reactor Auxiliary Building Mode Applicability:

All Page 169 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases 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 indicators operability, the conditions existence, or the reports 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.

The intent of the 15-minute duration is to size the FIRE and to discriminate against small FIRES that are readily extinguished (e.g., smoldering waste paper basket). In addition to alarms, other indications of a FIRE could be a drop in fire main pressure, automatic activation of a suppression system, etc.

Upon receipt, operators will take prompt actions to confirm the validity of an initial fire alarm, indication, or report. For EAL assessment purposes, the emergency declaration clock starts at the time that the initial alarm, indication, or report was received, and not the time that a subsequent verification action was performed. Similarly, the fire duration clock also starts at the time of receipt of the initial alarm, indication or report.

Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via EAL CA6.1 or SA9.1.

Table H-1 Fire Areas are those areas that contain equipment necessary for safe operation and shutdown of the plant (ref. 1).

Reference(s):

1. UFSAR Section 9.5.1.3.2 Fire Area-By-Fire Area Analysis

2. NEI 99-01 HU4 Page 170 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 4 - Fire Initiating Condition: FIRE potentially degrading the level of safety of the plant EAL:

HU4.2 Unusual Event Receipt of a single fire alarm (i.e., no other indications of a FIRE) (Note 13)

AND The fire alarm is indicating a FIRE within any Table H-1 area AND The existence of a FIRE is not verified (i.e., proved or disproved) within 30 min. of alarm receipt (Note 1)

Note 1: The 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: During Modes 1 and 2, HU4.2 is not applicable to a single fire alarm in containment. A fire in containment in these modes should be assessed under EAL HU4.1.

Table H-1 Fire Areas Containment Cooling Tower Areas Fuel Handling Building Reactor Auxiliary Building Mode Applicability:

All Definition(s):

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

Page 171 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases 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 indicators operability, the conditions existence, or the reports 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.

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 containment in Modes 1 and 2. The containment air flow design and Technical Specification requirements for operation of Containment Fan Coolers are such that multiple smoke detectors would be expected to alarm for a fire in containment. A fire in containment 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 is immediately applicable, and the emergency must be declared if the FIRE is not extinguished within 15-minutes of the report. If the alarm is verified to be due to an equipment failure or a spurious activation, and this verification occurs within 30-minutes of the receipt of the alarm, then this EAL is not applicable and no emergency declaration is warranted.

Basis-Related Fire Protection Requirements 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.

Page 172 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases In addition, 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. As used in HU4.2, the 30-minutes to verify a single alarm is well within this worst-case 1-hour time period.

Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via EAL CA6.1 or SA9.1.

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 H-1 Fire Areas are those areas that contain equipment necessary for safe operation and shutdown of the plant (ref. 1).

Reference(s):

1. OP-901-524 Fire in Areas Affecting Safe Shutdown

2. EP FAQ 2018-003

3. NEI 99-01 HU4 Page 173 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 4 - Fire Initiating Condition: FIRE potentially degrading the level of safety of the plant EAL:

HU4.3 Unusual Event A FIRE within the 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 - The area encompassed by physical barriers (the security fence) and to which access is controlled into the vital areas of the plant.

Basis:

This IC addresses the magnitude and extent of FIRES that may be indicative of a potential degradation of the level of safety of the plant.

In addition to a FIRE addressed by EAL HU4.1 or HU4.2, a FIRE within the plant PROTECTED AREA not extinguished within 60-minutes may also potentially degrade the level of plant safety.

Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via EAL CA6.1 or SA9.1.

Reference(s):

1. NEI 99-01 HU4 Page 174 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 4 - Fire Initiating Condition: FIRE potentially degrading the level of safety of the plant EAL:

HU4.4 Unusual Event A FIRE within the 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 - The area encompassed by physical barriers (the security fence) and to which access is controlled into the vital areas of the plant.

Basis:

This IC addresses the magnitude and extent of FIRES that may be indicative of a potential degradation of the level of safety of the plant.

If a FIRE within the plant PROTECTED AREA is of sufficient size to require a response by an offsite firefighting agency (e.g., a local town Fire Department), then the level of plant safety is potentially degraded. The dispatch of an offsite firefighting agency to the site requires an emergency declaration only if it is needed to actively support firefighting efforts because the fire is beyond the capability of the Fire Brigade to extinguish. Declaration is not necessary if the agency resources are placed on stand-by, or supporting post-extinguishment recovery or investigation actions.

Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via EAL CA6.1 or SA9.1.

Reference(s):

1. NEI 99-01 HU4 Page 175 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases 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 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 H-2 Safe Operation & Shutdown Rooms/Areas Room/Area Mode Turbine Building (all elevations and rooms) 1 Polisher Building (all elevations and rooms) 1

-4 RCA Letdown Valve Gallery 3

+21 RAB Switchgears A or B 3

-4 RCA Wing Area 4

-15 RCA Valve Gallery 4

-35 RCA Safeguard Rooms 4

+21 RAB Switchgears A or B 4 Mode Applicability:

1 - Power Operation, 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).

Page 176 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases 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. This gas release could originate either on site or off site.

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 Directors 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 room or area (e.g., fire suppression system testing).

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.

Page 177 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases 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 H-2 (Modes 1, 3 and 4 only).

Reference(s):

1. Attachment 2 Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases

2. NEI 99-01 HA5 Page 178 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 6 - Control Room Evacuation Initiating Condition: Control Room evacuation resulting in transfer of plant control to alternate locations EAL:

HA6.1 Alert An event has resulted in plant control being transferred from the Control Room to the Remote Shutdown Panel (LCP-43)

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-901-502 Evacuation of Control Room and Subsequent Plant Shutdown

2. NEI 99-01 HA6 Page 179 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 6 - Control Room Evacuation Initiating Condition: Inability to control a key safety function from outside the Control Room EAL:

HS6.1 Site Area Emergency An event has resulted in plant control being transferred from the Control Room to the Remote Shutdown Panel (LCP-43)

AND Control of any of the following key safety functions is not re-established within 15 min.

(Note 1):

Reactivity Control (Modes 1, 2 and 3 only)

Core Heat Removal 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 minutes whether or not the operating staff has control of key safety functions from the remote safe shutdown location(s).

Transfer of plant control and the time period to establish control begins when the last licensed operator leaves the Control Room.

Page 180 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases OP-901-502 Evacuation of Control Room and Subsequent Plant Shutdown provides specific instructions for evacuating the Control Room and establishing plant control at the local control panel LCP-43. The 15 minute limit is established to ensure control is established at LCP-43 in sufficient time to allow completion of the remaining time critical actions in OP-901-502 (ref. 1).

Escalation of the emergency classification level would be via IC FG1 or CG1 Reference(s):

1. OP-901-502 Evacuation of Control Room and Subsequent Plant Shutdown

2. EP FAQ 2015-014

3. NEI 99-01 HS6 Page 181 of 225

Waterford 3 SES EAL Basis Document Revision A 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 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 an UNUSUAL EVENT.

Reference(s):

1. NEI 99-01 HU7 Page 182 of 225

Waterford 3 SES EAL Basis Document Revision A 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 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 WF3 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 WF3. 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 (SOCA)).

PROTECTED AREA - The area encompassed by physical barriers (the security fence) and to which access is controlled into the vital areas of the plant.

SECURITY OWNR CONTROLLED AREA (SOCA) - The area inside the SOCA Vehicle Barrier System (VBS) up to the PROTECTED AREA fence line. Access to this area is controlled by the SOCA Personnel Access Control Point. The SOCA is part of but not equal to the Owner Controlled Area as described or defined in the Waterford 3 Emergency Plan.

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 Page 183 of 225

Waterford 3 SES EAL Basis Document Revision A 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 levels 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 WF3 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 WF3. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the SECUFRITY OWNER CONTROLLED AREA (SOCA)).

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

PROTECTED AREA - The area encompassed by physical barriers (the security fence) and to which access is controlled into the vital areas of the plant.

SECURITY OWNR CONTROLLED AREA (SOCA) - The area inside the SOCA Vehicle Barrier System (VBS) up to the PROTECTED AREA fence line. Access to this area is controlled by the SOCA Personnel Access Control Point. The SOCA is part of but not equal to the Owner Controlled Area as described or defined in the Waterford 3 Emergency Plan.

SITE BOUNDARY - The border of the Exclusion Area or an area corresponding to a distance of 914 meters from the Waterford 3 reactor. Also referred to as the Exclusion Area Boundary.

Page 184 of 225

Waterford 3 SES EAL Basis Document Revision A 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 a SITE AREA EMERGENCY.

Reference(s):

1. NEI 99-01 HS7 Page 185 of 225

Waterford 3 SES EAL Basis Document Revision A 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 WF3 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 WF3. 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 (SOCA)).

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 - The area encompassed by physical barriers (the security fence) and to which access is controlled into the vital areas of the plant.

SECURITY OWNR CONTROLLED AREA (SOCA) - The area inside the SOCA Vehicle Barrier System (VBS) up to the PROTECTED AREA fence line. Access to this area is controlled by the SOCA Personnel Access Control Point. The SOCA is part of but not equal to the Owner Controlled Area as described or defined in the Waterford 3 Emergency Plan.

Page 186 of 225

Waterford 3 SES EAL Basis Document Revision A 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 a GENERAL EMERGENCY.

Reference(s):

1. NEI 99-01 HG7 Page 187 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases Category S - System Malfunction EAL Group: Hot Conditions (RCS temperature > 200ºF); EALs in this category are applicable only in one or more hot operating modes.

Numerous system-related equipment failure events that warrant emergency classification have been identified in this category. They may pose actual or potential threats to plant safety.

The events of this category pertain to the following subcategories:

1. Loss of Safety Bus AC Power Loss of emergency electrical power can compromise plant SAFETY SYSTEM operability including decay heat removal and emergency core cooling systems which may be necessary to ensure fission product barrier integrity. This category includes loss of onsite and offsite sources for 4160 VAC safety 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.

Page 188 of 225

Waterford 3 SES EAL Basis Document Revision A 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.

Page 189 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases Category: S - System Malfunction Subcategory: 1 - Loss of Safety Bus AC Power Initiating Condition: Loss of all offsite AC power capability to safety buses for 15 minutes or longer EAL:

SU1.1 Unusual Event Loss of all offsite AC power capability, Table S-1, to 4160 VAC safety buses 3A and 3B 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 S-1 AC Power Sources Emergency Diesel Generator A Emergency Diesel Generator B Onsite Temporary Emergency Diesels (TEDs)

(if already aligned)

Unit Auxiliary Transformer 3A Unit Auxiliary Transformer 3B Startup Transformer 3A Startup Transformer 3B Offsite Unit Auxiliary Transformer 3A (when back-fed from offsite)

Unit Auxiliary Transformer 3B (when back-fed from offsite)

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 safety 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 safety 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.

Page 190 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases Escalation of the emergency classification level would be via IC SA1.

Reference(s):

1. UFSAR Section 8.2, Offsite Power System

2. NEI 99-01 SU1 Page 191 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases Category: S - System Malfunction Subcategory: 1 - Loss of Safety Bus AC Power Initiating Condition: Loss of all but one AC power source to safety buses for 15 minutes or longer EAL:

SA1.1 Alert AC power capability, Table S-1, to 4160 VAC safety buses 3A and 3B 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 S-1 AC Power Sources Emergency Diesel Generator A Emergency Diesel Generator B Onsite Temporary Emergency Diesels (TEDs)

(if already aligned)

Unit Auxiliary Transformer 3A Unit Auxiliary Transformer 3B Startup Transformer 3A Startup Transformer 3B Offsite Unit Auxiliary Transformer 3A (when back-fed from offsite)

Unit Auxiliary Transformer 3B (when back-fed from offsite)

Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 3 - Hot Shutdown Definition(s):

SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10 CFR 50.2):

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

Page 192 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases (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 safety bus. Some examples of this condition are presented below.

A loss of all offsite power with a concurrent failure of all but one emergency power source (e.g., an onsite diesel generator).

A loss of all offsite power and loss of all emergency power sources (e.g., onsite diesel generators) with a single train of safety buses being back-fed from the unit main generator via a Unit Auxiliary Transformer.

A loss of emergency power sources (e.g., onsite diesel generators) with a single train of safety buses being fed from an offsite power source (Startup Transformer).

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.

Temporary Emergency Diesels (TEDs) can be credited if already installed in accordance with site procedures (ref. 5, 6).

This EAL is the hot condition equivalent of the cold condition EAL CU2.1.

Reference(s):

1. UFSAR Section 8.1, Onsite Power System

2. UFSAR Section 8.2, Offsite Power System

3. OP-901-310 Loss of Train A Safety Bus

4. OP-901-311 Loss of Train B Safety Bus

5. Technical Specifications 3/4.8.1 A.C. Sources

6. ME-001-012, Temporary Power from Temporary Diesel for 3A2 and 3B2 4KV Buses (Modes 1-6)

7. NEI 99-01 SA1 Page 193 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases Category: S - System Malfunction Subcategory: 1 - Loss of Safety Bus AC Power Initiating Condition: Loss of all offsite power and all onsite AC power to safety buses for 15 minutes or longer EAL:

SS1.1 Site Area Emergency Loss of all offsite and all onsite AC power to 4160 VAC safety buses 3A and 3B 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:

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 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.

Page 194 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases Escalation of the emergency classification level would be via IC AG1, FG1 or SG1.

Temporary Emergency Diesels (TEDs) can be credited if already installed in accordance with site procedures (ref. 4, 5).

This EAL is the hot condition equivalent of the cold condition EAL CA2.1.

Reference(s):

1. UFSAR Section 8.1, Onsite Power System

2. UFSAR Section 8.2, Offsite Power System

3. OP-902-005 Station Blackout Recovery

4. Technical Specifications 3/4.8.1 A.C. Sources

5. ME-001-012, Temporary Power from Temporary Diesel for 3A2 and 3B2 4KV Buses (Modes 1-6)

6. NEI 99-01 SS1 Page 195 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases Category: S -System Malfunction Subcategory: 1 - Loss of Safety Bus AC Power Initiating Condition: Prolonged loss of all offsite and all onsite AC power to safety buses EAL:

SG1.1 General Emergency Loss of all offsite and all onsite AC power to 4160 VAC safety buses 3A and 3B AND EITHER:

Long-term RCS heat removal capability is not likely to be established and maintained per FIG-001, FLEX Implementing Guideline, Extended Loss of AC Power (Note 1)

Representative CETs reading > 1,200ºF 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 safety buses that result in degraded core cooling. 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.

Page 196 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases 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 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 long-term RCS heat removal capability is not likely to be established and maintained per FIG-001, FLEX Implementing Guideline, Extended Loss of AC Power. FIG-001 contains attachments that list execution timelines for use of Train A or Train B power. These timelines describe the execution times for Flex actions, such as Deep Load Shed and aligning Flex DG, to assist in evaluating the success of Flex actions.

The estimate for restoring at least one safety bus or establishing long term cooling with Flex equipment 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.

Temporary Emergency Diesels (TEDs) can be credited if already installed in accordance with site procedures (ref. 4, 5).

Reference(s):

1. UFSAR Section 8.1, Onsite Power System

2. UFSAR Section 8.2, Offsite Power System

3. OP-902-005 Station Blackout Recovery

4. Technical Specifications 3/4.8.1 A.C. Sources

5. ME-001-012, Temporary Power from Temporary Diesel for 3A2 and 3B2 4KV Buses (Modes 1-6)

6. F I G - 0 0 1 , FLEX Implementing Guideline, Extended Loss of AC Power

7. UFSAR Section 1.9A Inadequate Core Cooling Instrumentation

8. CEOG Generic Accident Management Guidelines - Phase 1, Initial Diagnosis

9. NEI 99-01 SG1 Page 197 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases Category: S -System Malfunction Subcategory: 1 - Loss of Safety Bus AC Power Initiating Condition: Loss of all safety bus 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 4160 VAC safety buses 3A and 3B for 15 min. (Note 1)

AND Indicated voltage is < 108 VDC on all vital DC 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:

This IC addresses a concurrent and prolonged loss of both safety bus AC and vital DC power.

A loss of all safety bus AC power compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink. A loss of vital DC power compromises the ability to monitor and control SAFETY SYSTEMS. A sustained loss of both safety bus AC and vital DC power will lead to multiple challenges to fission product barriers.

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

The 15-minute emergency declaration clock begins at the point when both EAL thresholds are met.

Page 198 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases Temporary Emergency Diesels (TEDs) can be credited if already installed in accordance with site procedures (ref. 4, 5).

This IC refers to loss of vital DC power from the 3A-DC, 3B-DC, and 3AB-DC buses. Less than 108 VDC bus voltage is based on the minimum bus voltage necessary for the operation of safety related equipment (ref. 5, 6, 7).

Reference(s):

1. UFSAR Section 8.1, Onsite Power System

2. UFSAR Section 8.2, Offsite Power System

3. OP-902-005 Station Blackout Recovery

4. Technical Specifications 3/4.8.1 A.C. Sources

5. ME-001-012, Temporary Power from Temporary Diesel for 3A2 and 3B2 4KV Buses (Modes 1-6)

6. ECE91-058 Battery 3A-S A Train Calculation for Station Blackout

7. ECE91-059 Battery 3B-S B Train Calculation for Station Blackout

8. ECE91-060 Battery 3AB-S Calculation for Station Blackout

9. NEI 99-01 SG8 Page 199 of 225

Waterford 3 SES EAL Basis Document Revision A 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 < 108 VDC on all vital DC 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:

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.

This IC refers to loss of vital DC power from the 3A-DC, 3B-DC, and 3AB-DC buses.

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

Escalation of the emergency classification level would be via IC AG1, FG1 or SG1.

Less than 108 VDC bus voltage is based on the minimum bus voltage necessary for the operation of safety related equipment (ref. 1, 2, 3).

This EAL is the hot condition equivalent of the cold condition EAL CU4.1.

Page 200 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases Reference(s):

1. ECE91-058 Battery 3A-S A Train Calculation for Station Blackout

2. ECE91-059 Battery 3B-S B Train Calculation for Station Blackout

3. ECE91-060 Battery 3AB-S Calculation for Station Blackout

4. Technical Specifications section 3/4.8.2.2 D.C. Sources - Shutdown

5. NEI 99-01 SS8 Page 201 of 225

Waterford 3 SES EAL Basis Document Revision A 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 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 S-2 Safety System Parameters Reactor power RCS level RCS pressure Core exit temperature Level in at least one S/G S/G emergency feed water 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.

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

Page 202 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases 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 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 RCS level cannot be determined from the indications and recorders on a main control board, the SPDS or the plant computer, the availability of other parameter values may be compromised as well.

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

Escalation of the emergency classification level would be via EAL SA3.1.

Reference(s):

1. NEI 99-01 SU2 Page 203 of 225

Waterford 3 SES EAL Basis Document Revision A 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 with a significant transient in progress EAL:

SA3.1 Alert An UNPLANNED event results in the inability to monitor one or more Table S-2 parameters from within the Control Room for 15 min. (Note 1)

AND Any significant transient is in progress, Table S-3 Note 1: The Emergency 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 S-2 Safety System Parameters Reactor power RCS level RCS pressure Core exit temperature Level in at least one S/G S/G emergency feed water flow to at least one S/G Table S-3 Significant Transients Turbine runback > 25% reactor power Electrical load rejection > 25% full electrical load (300 MWE)

Reactor trip ECCS (SI) activation Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Page 204 of 225

Waterford 3 SES EAL Basis Document Revision A 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 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 RCS level cannot be determined from the indications and recorders on a main control board, the SPDS or the plant computer, the availability of other parameter values may be compromised as well.

Page 205 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases 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 FS1 or AS1 Reference(s):

1. NEI 99-01 SA2 Page 206 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases Category: S - System Malfunction Subcategory: 4 - RCS Activity Initiating Condition: Reactor coolant activity greater than Technical Specification allowable limits EAL:

SU4.1 Unusual Event RCS sample activity > 60 Ci/gm dose equivalent I-131 OR RCS sample activity > 1.0 Ci/gm dose equivalent I-131 for > 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months during one continuous time interval (Note 1)

OR RCS sample activity >100/ Ci/gm 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 FA1 or the Recognition Category A ICs.

Reference(s):

1. Technical Specification 3.4.7 Reactor Coolant System - Specific Activity

2. NEI 99-01 SU3 Page 207 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases 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 are focused on a loss of mass from the RCS due to unidentified leakage", "pressure boundary leakage" or "identified leakage (as these leakage types are defined in the plant Technical Specifications). The third condition addresses an RCS mass loss caused by an UNISOLABLE leak through an interfacing system. These conditions thus apply to leakage into the containment, a secondary-side system (e.g., steam generator tube leakage) or a location outside of containment (ref. 1).

Page 208 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases The leak rate values for each condition were selected because they are usually observable with normal Control Room indications. Lesser values typically require time-consuming calculations to determine (e.g., a mass balance calculation). The first condition uses a lower value that reflects the greater significance of unidentified or pressure boundary leakage (ref.

2).

The release of mass from the RCS due to the as-designed/expected operation of a relief valve does not warrant an emergency classification. An emergency classification would be required if a mass loss is caused by a relief valve that is not functioning as designed/expected (e.g., a relief valve sticks open and the line flow cannot be isolated).

The 15-minute threshold duration allows sufficient time for prompt operator actions to isolate the leakage, if possible.

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

Reference(s):

1. Technical Specifications 3.4.5 Reactor Coolant Leakage

2. OP-901-111 Reactor Coolant System Leakage

3. NEI 99-01 SU4 Page 209 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases 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 (manual reactor trip push buttons or DRT) 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:

Consistent with OP-902-000, Standard Post Trip Actions, and its associated technical guide TGOP-902-000, a successful reactor trip includes inserting CEAs. Actions are included for failure of individual CEAs to insert. Should an automatic reactor trip fail to occur, the operator is directed to use the manual pushbuttons on the main control board. Should this fail, the operator is directed to manually initiate Diverse Reactor Trip (DRT), also from the main control board. Should this also fail, the operator is directed to open the feeder breakers for the 32 bus, which supply power to the CEA motor generator set and then reclose the breakers.

After a successful reactor trip neutron power should immediately drop to approximately 6%

due to prompt drop. Therefore, for the purpose of emergency classification, reactor power less than or equal to 5% is used to identify a successful reactor trip (ref. 2).

This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor trip that results in a reactor shutdown, and either a subsequent operator manual action taken at the reactor control consoles or an automatic trip is successful in shutting down the reactor.

This event is a precursor to a more significant condition and thus represents a potential degradation of the level of safety of the plant.

Page 210 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases 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 of an automatic reactor trip, operators will promptly initiate manual actions at the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor trip). If these manual actions are successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plants decay heat removal systems.

If an initial manual reactor trip is unsuccessful, operators will promptly take manual action at another location(s) on the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor trip) using a different switch). Depending upon several factors, the initial or subsequent effort to manually trip the reactor, or a concurrent plant condition, may lead to the generation of an automatic reactor trip signal. If a subsequent manual or automatic trip is successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plants decay heat removal systems.

A manual action at the reactor control consoles is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor trip). Use of the Reactor Trip Pushbuttons at either CP-2 or CP-8 or the Diverse Reactor Trip Pushbuttons at CP-2 satisfy this requirement. This action does not include manually driving in control rods or implementation of boron injection strategies. Actions taken at back-panels or other locations (opening A32 and B32 Bus Feeders) within the Control Room, or any location outside the Control Room, are not considered to be at the reactor control consoles (ref. 1).

The plant response to the failure of an automatic or manual reactor trip will vary based upon several factors including the reactor power level prior to the event, availability of the condenser, performance of mitigation equipment and actions, other concurrent plant conditions, etc. If subsequent operator manual actions taken at the reactor control consoles are also unsuccessful in shutting down the reactor, then the emergency classification level will escalate to an Alert via IC SA6. Depending upon the plant response, escalation is also possible via IC FA1. Absent the plant conditions needed to meet either IC SA6 or FA1, an Unusual Event declaration is appropriate for this event.

Page 211 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases Should a reactor trip 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 and the RPS fails to automatically shutdown the reactor, then this IC and associated 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 failure is determined through other means (e.g., assessment of test results),

then this IC and associated EALs are not applicable and no classification is warranted.

Reference(s):

1. OP-902-000 Standard Post Trip Actions

2. Calculation No. EC-S98-001 EOP Value Basis Document Application X.01

3. NEI 99-01 SU5 Page 212 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases 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 (manual reactor trip push buttons or DRT) 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):

None Basis:

Consistent with OP-902-000, Standard Post Trip Actions, and its associated technical guide TGOP-902-000, a successful reactor trip includes inserting CEAs. Actions are included for failure of individual CEAs to insert. Should an automatic reactor trip fail to occur, the operator is directed to use the manual pushbuttons on the main control board. Should this fail, the operator is directed to manually initiate Diverse Reactor Trip (DRT), also from the main control board. Should this also fail, the operator is directed to open the feeder breakers for the 32 bus, which supply power to the CEA motor generator set and then reclose the breakers.

After a successful reactor trip neutron power should immediately drop to approximately 6%

due to prompt drop. Therefore, for the purpose of emergency classification, reactor power less than or equal to 5% is used to identify a successful reactor trip (ref. 2).

This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor trip that results in a reactor shutdown, and either a subsequent operator manual action taken at the reactor control consoles or an automatic trip is successful in shutting down the reactor.

This event is a precursor to a more significant condition and thus represents a potential degradation of the level of safety of the plant.

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.

Page 213 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases Following the failure on an automatic reactor trip, operators will promptly initiate manual actions at the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor trip). If these manual actions are successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plants decay heat removal systems.

If an initial manual reactor trip is unsuccessful, operators will promptly take manual action at another location(s) on the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor trip) using a different switch). Depending upon several factors, the initial or subsequent effort to manually trip the reactor, or a concurrent plant condition, may lead to the generation of an automatic reactor trip signal. If a subsequent manual or automatic trip is successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plants decay heat removal systems.

A manual action at the reactor control consoles is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor trip). Use of the Reactor Trip Pushbuttons at either CP-2 or CP-8 or the Diverse Reactor Trip Pushbuttons at CP-2 satisfy this requirement. This action does not include manually driving in control rods or implementation of boron injection strategies. Actions taken at back-panels or other locations (opening A32 and B32 Bus Feeders) within the Control Room, or any location outside the Control Room, are not considered to be at the reactor control consoles (ref. 1).

The plant response to the failure of an automatic or manual reactor trip will vary based upon several factors including the reactor power level prior to the event, availability of the condenser, performance of mitigation equipment and actions, other concurrent plant conditions, etc. If subsequent operator manual actions taken at the reactor control consoles are also unsuccessful in shutting down the reactor, then the emergency classification level will escalate to an Alert via IC SA6. Depending upon the plant response, escalation is also possible via IC FA1. Absent the plant conditions needed to meet either IC SA6 or FA1, an Unusual Event declaration is appropriate for this event.

Should a reactor trip 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 and the RPS fails to automatically shutdown the reactor, then this IC and associated 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 failure is determined through other means (e.g., assessment of test results), then this IC and associated EALs are not applicable and no classification is warranted.

Reference(s):

1. OP-902-000 Standard Post Trip Actions

2. Calculation No. EC-S98-001 EOP Value Basis Document Application X.01

3. NEI 99-01 SU5 Page 214 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases 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 control console (manual reactor trip push buttons and DRT) 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:

Consistent with OP-902-000, Standard Post Trip Actions, and its associated technical guide TGOP-902-000, a successful reactor trip includes inserting CEAs. Actions are included for failure of individual CEAs to insert. Should an automatic reactor trip fail to occur, the operator is directed to use the manual pushbuttons on the main control board. Should this fail, the operator is directed to manually initiate Diverse Reactor Trip (DRT), also from the main control board. Should this also fail, the operator is directed to open the feeder breakers for the 32 bus, which supply power to the CEA motor generator set and then reclose the breakers.

After a successful reactor trip neutron power should immediately drop to approximately 6%

due to prompt drop. Therefore, for the purpose of emergency classification, reactor power less than or equal to 5% is used to identify a successful reactor trip (ref. 2).

Page 215 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor trip that results in a reactor shutdown, and subsequent operator manual actions taken at the reactor control consoles to shutdown the reactor are also unsuccessful. This condition represents an actual or potential substantial degradation of the level of safety of the plant. An emergency declaration is required even if the reactor is subsequently shutdown by an action taken away from the reactor control consoles since this event entails a significant failure of the 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). Use of the Reactor Trip Pushbuttons at either CP-2 or CP-8 or the Diverse Reactor Trip Pushbuttons at CP-2 satisfy this requirement. This action does not include manually driving in control rods or implementation of boron injection strategies. If this action(s) is unsuccessful, operators would immediately pursue additional manual actions at locations away from the reactor control console (e.g., locally opening breakers). Actions taken at back panels or other locations within the Control Room (opening A32 and B32 Bus Feeders), or any location outside the Control Room, are not considered to be at the reactor control console.

The plant response to the failure of an automatic or manual reactor trip will vary based upon several factors including the reactor power level prior to the event, availability of the condenser, performance of mitigation equipment and actions, other concurrent plant conditions, etc. If the failure to shut down the reactor is prolonged enough to cause a challenge to the core cooling or RCS heat removal safety functions, the emergency classification level will escalate to a Site Area Emergency via IC SS6. Depending upon plant responses and symptoms, escalation is also possible via IC FS1. Absent the plant conditions needed to meet either IC SS6 or FS1, an Alert declaration is appropriate for this event.

It is recognized that plant responses or symptoms may also require an Alert declaration in accordance with the Recognition Category F ICs; however, this IC and EAL are included to ensure a timely emergency declaration.

Reference(s):

1. OP-902-000 Standard Post Trip Actions

2. Calculation No. EC-S98-001 EOP Value Basis Document Application X.01

3. NEI 99-01 SA5 Page 216 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases 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:

Representative CET readings > 1,200ºF Any OP-902-008 Functional Recovery RCS/Core Heat Removal safety function criterion is not met 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 Definition(s):

None Basis:

Consistent with OP-902-000, Standard Post Trip Actions, and its associated technical guide TGOP-902-000, a successful reactor trip includes inserting CEAs. Actions are included for failure of individual CEAs to insert. Should an automatic reactor trip fail to occur, the operator is directed to use the manual pushbuttons on the main control board. Should this fail, the operator is directed to manually initiate Diverse Reactor Trip (DRT), also from the main control board. Should this also fail, the operator is directed to open the feeder breakers for the 32 bus, which supply power to the CEA motor generator set and then reclose the breakers.

After a successful reactor trip neutron power should immediately drop to approximately 6%

due to prompt drop. Therefore, for the purpose of emergency classification, reactor power less than or equal to 5% is used to identify a successful reactor trip (ref. 2).

Page 217 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor trip that results in a reactor shutdown, all subsequent operator actions to manually shutdown the reactor are unsuccessful, and continued power generation is challenging the capability to adequately remove heat from the core and/or the RCS. This condition will lead to fuel damage if additional mitigation actions are unsuccessful and thus warrants the declaration of a Site Area Emergency.

In some instances, the emergency classification resulting from this IC/EAL may be higher than that resulting from an assessment of the plant responses and symptoms against the Recognition Category F ICs/EALs. This is appropriate in that the Recognition Category F ICs/EALs do not address the additional threat posed by a failure to shutdown 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.

The process of checking the safety functions in EOPs is periodic and continuous as long as the procedure is in use. The fifteen minute interval (subject to Note 1) provides a suitable opportunity to assess plant conditions with respect to the threshold criteria.

Escalation of the emergency classification level would be via IC AG1 or FG1.

Reference(s):

1. OP-902-000 Standard Post Trip Actions

2. Calculation No. EC-S98-001 EOP Value Basis Document Application X.01

3. OP-902-008 Functional Recovery

4. UFSAR Section 1.9A Inadequate Core Cooling Instrumentation

5. CEOG Generic Accident Management Guidelines - Phase 1, Initial Diagnosis

6. TG-OP-902-008 Technical Guide for Functional Recovery

7. NEI 99-01 SS5 Page 218 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases Category: S - System Malfunction Subcategory: 7 - Loss of Communications Initiating Condition: Loss of all onsite or offsite communications capabilities EAL:

SU7.1 Unusual Event Loss of all Table S-4 onsite communication methods OR Loss of all Table S-4 State and local agency communication methods OR Loss of all Table S-4 NRC communication methods Table S-4 Communication Methods Onsite State/ NRC System Local Telephone System X X X Operational Hotline X Plant Radio System (O&M) X Plant Paging System X Sound Powered Phone System X Civil Defense Radio System X Satellite Phones X X Emergency Notification System (ENS) X Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):

None Page 219 of 225

Waterford 3 SES EAL Basis Document Revision A 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 St. Charles Parish Department of Homeland Security and Emergency Preparedness, St. Charles Parish Sherriffs Department 911 Center, St. John the Baptist Parish Office of Emergency Preparedness, St. John the Baptist Parish Sherriffs Department 911 Center, Louisiana Department of Environmental Quality and the Louisiana Governor's Office of Homeland Security and Emergency Preparedness. The third EAL condition 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. EP-003-070 Emergency Communications Systems

2. Waterford 3 SES Emergency Plan Section 7.5 Communications Systems

3. NEI 99-01 SU6 Page 220 of 225

Waterford 3 SES EAL Basis Document Revision A 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 a required actuation signal OR Containment pressure > 17.7 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 either:

One train of the Containment Spray System (operating with 1750 gpm flow) AND One train of the Containment Cooling System (one fan cooler required)

OR Two trains of the Containment Spray System (operating with 1750 gpm flow each).

Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):

None Basis:

This EAL addresses a failure of one or more containment penetrations to automatically isolate (close) when required by an actuation signal. It also addresses an event that results in high containment pressure with a concurrent failure of containment pressure control systems.

Absent challenges to another fission product barrier, either condition represents potential degradation of the level of safety of the plant.

For the first condition, the containment isolation signal must be generated as the result on an off-normal/accident condition (e.g., a safety injection or high containment pressure); a failure resulting from testing or maintenance does not warrant classification. The determination of containment and penetration status - isolated or not isolated - should be made in accordance with the appropriate criteria contained in the plant AOPs and EOPs. The 15-minute criterion is included to allow operators time to manually isolate the required penetrations, if possible.

The second condition addresses a condition where containment pressure is greater than the setpoint at which containment energy (heat) removal systems are designed to automatically actuate, and less than one full train of equipment is capable of operating per design.

Page 221 of 225

Waterford 3 SES EAL Basis Document Revision A Attachment 1 - Emergency Action Level Technical Bases 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) 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-902-002 Loss of Coolant Accident Recovery

2. UFSAR Section 6.2.2 Containment Heat Removal Systems

3. Technical Specification 3.6.2 Containment Systems - Depressurization and Cooling Systems

4. NEI 99-01 SU7 Page 222 of 225

Waterford 3 SES EAL Basis Document Revision A 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 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 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 Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Page 223 of 225