NG-12-0157, Response to Supplemental Request for Additional Information Amendment to Change Emergency Action Levels

Response to Supplemental Request for Additional Information Amendment to Change Emergency Action Levels

ML121000327
Person / Time
Site:	Duane Arnold
Issue date:	04/05/2012
From:	Wells P Nextera Energy, NextEra Energy Duane Arnold
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NG-12-0157
Download: ML121000327 (220)

Text

NExTera ENERGYC 4o S DUANE ARNOLD April 5, 2012 NG-12-0157 10 CFR 50.90 10 CFR 50 Appendix E U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C. 20555-0001 Duane Arnold Energy Center Docket 50-331 Renewed License No. DPR-49 Response to Supplemental Request for Additional Information Re: Amendment to Change Emergency Action Levels

References:

1. Letter, Christopher R. Costanzo (NextEra Energy Duane Arnold, LLC) to Document Control Desk (USNRC), Proposed Changes to the Emergency Plan (TSCR-126), dated May 31, 2011, NG-11-0117 (ML111540279 [package])

2. Letter, Karl Feintuch (USNRC) to Peter Wells (NextEra Energy Duane Arnold, LLC), Request for Additional Information Re:

Amendment to Change Emergency Action Levels, dated January 17, 2012 (TAC ME6508) (ML120120318)

3. Letter, Peter Wells (NextEra Energy Duane Arnold, LLC) to Document Control Desk (USNRC), Response to Request for Additional Information Re: Amendment to Change Emergency Action Levels, dated March 16, 2012, NG-12-0064 Reference 1 provided an amendment request to revise the Duane Arnold Energy Center (DAEC) emergency plan. The proposed changes involve upgrading selected DAEC Emergency Action Levels (EALs) based on NEI 99-01, Revision 5, "Methodology for Development of Emergency Action Levels," using the guidance of NRC Regulatory Issue Summary 2003-18, Supplement 2, "Use of Nuclear Energy Institute (NEI) 99-01, Methodology for Development of Emergency Action Levels."

In Reference 2, the Staff issued a request for additional information regarding Reference 1. The response to this request was provided in Reference 3. 4W/S NextEra Energy Duane Arnold, LLC, 3277 DAEC Road, Palo, IA 52324

NG-12-0157 April 5, 2012 Page 2 of 2 In an email dated March 22, 2012, the Staff issued a supplemental request for additional information regarding Reference 1. The response to the request is provided in Enclosure 1 to this letter. An additional minor change to EAL Bases Document EBD F is also included with this submittal and discussed in Enclosure 1.

This minor change was discussed with the staff in a phone call on April 2, 2012.

The following enclosures are provided to this letter:

" Enclosure 1 - Response to Supplemental Request for Additional Information Re:

Amendment to Change Emergency Action Levels

• Enclosure 2 - Comparison Matrix of NEI 99-01, Rev. 5 generic guidance to proposed NextEra Energy Duane Arnold Emergency Classification System

" Enclosure 3 - Emergency Action Level Design Basis Document for NextEra Energy Duane Arnold NextEra Energy Duane Arnold, LLC requests approval of the amendment request of Reference 1 by June 1,2012.

This letter contains no new commitments nor revises any previous commitments.

In accordance with 10 CFR 50.91, a copy of this application, with attachments, is being provided to the designated Iowa State Official.

If you have any questions, please contact Steve Catron at (319) 851-7234.

I decl der penalty perjury that the foregoing is true and correct. Executed on 5

20 Ar Peter [Is Vice President, Duane Arnold Energy Center NextEra Energy Duane Arnold, LLC Enclosures 1. Response to Supplemental Request for Additional Information Re: Amendment to Change Emergency Action Levels

2. Comparison Matrix of NEI 99-01, Rev. 5 generic guidance to proposed NextEra Energy Duane Arnold Emergency Classification System

3. Emergency Action Level Design Basis Document for NextEra Energy Duane Arnold cc: NRC Regional Administrator NRC Resident Inspector NRC Project Manager M. Rasmusson (State of Iowa)

Enclosure 1 to NG-12-0157 Page 1 of 4 ENCLOSURE I RESPONSE TO SUPPLEMENTAL REQUEST FOR ADDITIONAL INFORMATION RE: AMENDMENT TO CHANGE EMERGENCY ACTION LEVELS

Enclosure 1 to NG-12-0157 Page 2 of 4 Response to Supplemental Request for Additional Information Re: Amendment to Change Emergency Action Levels NRC Question #1:

[The reviewer] need[s] to have [the licensee's] entire proposed EAL Technical Basis Document (the "Document") on the docket as part of [its] response, even if they are not changing some of the sections. Please submit [this Document] so we (NRC) can re-baseline the its approval for the entire EAL scheme (as encompassed in [DAEC's]

EAL Technical Basis Document).

The front section, ISFSI EAL, fission barrier matrix section, and the systems section were not provided with [the recent RAIl response.

[The NRC staff needs the docketed Document in ADAMS to reference it as] the single document found acceptable for DAECs EAL Scheme. [This reference is necessary] so that future inspectors, and site licensing staff, can readily determine the approved baseline scheme when performing future 10 CFR 50.54(q) reviews/changes.

NextEra Energy Duane Arnold Response: of this submittal provides the Comparison Matrix of NEI 99-01, Rev. 5 Generic Guidance to the Proposed Nextera Energy Duane Arnold Emergency Classification System. Enclosure 3 of this submittal provides the entire proposed EAL Technical Basis Document.

NRC Question #2:

The response to RAI #6, while adequate, does not answer the question, i.e., the incorporation of information to EAL EU1. Note that this EAL was not provided with this response (see <1> above.)

NextEra Energy Duane Arnold Response:

The following was added as the first paragraph in the basis discussion of EUI:

"Security related events for the ISFSI are to be covered under the Security EALs in the Hazard recognition category."

NRC Question #3:

The response to RAI #10 requires more information to support the justification for why the Control Room is not an area for consideration, as well as documenting what

Enclosure 1 to NG-12-0157 Page 3 of 4 areas were considered and the basis for why each area was determined to be not applicable.

NextEra Energy Duane Arnold Response:

NextEra Energy Duane Arnold staff have reviewed all of the Safe ShutdownNital Areas (listed in the table below) for local actions required by Operators during normal plant shutdown and cooldown. None of the areas reviewed require local actions except for the Torus Room and Control Room.

" Regarding the Torus Room, NextEra Energy Duane Arnold reviewed the requirements for local manual operation of the Low Pressure Cooling Injection (LPCI) system RHR crosstie during Mode 3 for initiation of Shutdown Cooling.

Since LPCI is inoperable/removed from service as a result of this evolution, this also would not apply to this EAL when considering the NEI guidance in HA3. 1:

o If the equipment in the stated area was already inoperable,or out of service, before the event occurred,then this EAL should not be declaredas it will have no adverse impact on the ability of the plant to safely operate or safely shutdown beyond that already allowed by Technical Specifications at the time of the event.

• Regarding the Control Room, the control building ventilation system has two redundant standby filter unit trains to supply filtered air to the Control Room.

The Control Building Envelope is the combination of walls, floor, roof, ducting, doors, penetrations and equipment that physically form a boundary. The filtration units and envelope provide a boundary in which a positive pressure can be maintained that would minimize infiltration of gases into the Control Room. In the event a Control Room evacuation is needed for any reason, existing EAL HA5.1 would be declared.

.. SafeStutdown/Vital Areas Category Area 1G31 DG and Day Tank Rooms, 1G21 DG and Day Tank Power Rooms, Battery Rooms, Essential Switchgear Rooms, Cable Spreading Room Heat Colant Sink /

Suy Torus Room, Intake Structure, Pumphouse Coolant Supply Containment Drywell, Torus Emergency NE, NW, SE Corner Rooms, HPCI Room, RCIC Room, RHR Systems Valve Room, North CRD Area, South CRD Area, CSTs Control Building, Remote Shutdown Panel 1C388 Area, Panel 1C55/56 Area, SBGT Room

Enclosure 1 to NG-12-0157 Page 4 of 4 Additional Change:

Additionally, EBD F is being changed to remove the ambiguous 3rd paragraph that begins with, "The "Failure of both isolation valves..." loss threshold..." in the Primary Containment barrier (Leakage) for Containment Isolation Valve Status after Containment Isolation Signal" section (page 26 of 31).

NG-12-0157 Enclosure 2 Enclosure 2 COMPARISON MATRIX OF NEI 99-01, REV. 5 GENERIC GUIDANCE TO PROPOSED NEXTERA ENERGY DUANE ARNOLD EMERGENCY CLASSIFICATION SYSTEM 37 Pages follow

NG-12-01657 Enclosure 3 Enclosure 3 EMERGENCY ACTION LEVEL DESIGN BASIS DOCUMENT FOR NEXTERA ENERGY DUANE ARNOLD 175 Pages follow

EAL BASES DOCUMENT EBD-C Rev. 1 next COLD SHUTDOWN/REFUELING - SYSTEMS MALFUNCTION Page 1 of 28 IIUsage

~ LevelI Information Use Effective Date:

>>~ .:'TECHNICAL REVIEW Prepared by: Date:

Reviewed by: Date:

Emergency Planning Staff PROCEDURE APPROV~AL I am responsible for the technical content of this procedure.

Approved by: Date:

Manager, Emergency Planning

EAL BASES DOCUMENTE EBD-C

+

Rev. 1 next COLD SHUTDOWN/REFUELING - SYSTEMS MALFUNCTION Page 2 of 28 CUI RCS Leakage EVENT TYPE: Coolant Leakage OPERATING MODE APPLICABILITY: Cold S/D EAL THRESHOLD VALUE:

Note: The Emergency Director (OSM, EC or ER&RD) 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.

CU1.1 RCS leakage results in the inability to maintain or restore RPV level GREATER THAN 170 inches for 15 minutes or longer.

DAEC EAL INFORMATION:

This IC is included as a NOUE because it is considered to be a potential degradation of the level of safety of the plant. The 10 gpm value for the unidentified and pressure boundary leakage was selected as it is sufficiently large to be observable via normally installed instrumentation or reduced inventory instrumentation such as level hose indication. Lesser values must generally be determined through time-consuming surveillance tests (e.g., mass balances). The EAL for identified leakage is set at a higher value due to the lesser significance of identified leakage in comparison to unidentified or pressure boundary leakage. Prolonged loss of RCS Inventory may result in escalation to the Alert level via either IC CA1 or CA4.

Relief valve normal operation should be excluded from this IC. However, a relief valve that operates and fails to close per design should be considered applicable to this IC if the relief valve cannot be isolated.

The difference between CU1 and CU2 deals with the RCS conditions that exist between cold shutdown and refueling mode applicability. In cold shutdown, the RCS will normally be intact and RCS inventory and level monitoring means such as makeup volume control tank levels are normally available. In the refueling mode the RCS is not intact and RPV level and inventory are monitored by different means.

REFERENCES:

cul

EAL BASES DOCUMENT. . EBD-C Rev. 1 next COLD SHUTDOWN/REFUELING - SYSTEMS MALFUNCTION Page 3 of 28

1. NEI 99-01 Rev. 5, NEI Methodology for Development of Emergency Action Levels cul

EAL BASES DOCUMENT EBD-C Rev. 1 next COLD SHUTDOWN/REFUELING - SYSTEMS MALFUNCTION Page 4 of 28 CU2 Unplanned Loss of RCS/RPV Inventory EVENT TYPE: RCS Level OPERATING MODE APPLICABILITY: Refueling EAL THRESHOLD VALUE:

Note: The Emergency Director (OSM, EC or ER&RD) 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.

CU2.1 UNPLANNED RCS/RPV level drop as indicated by ALL of the following conditions being met:

" RCS/RPV level band is established above the RPV flange

• RCS/RPV water level drops below the RPV flange for 15 minutes or longer OR CU2.2 UNPLANNED RCS/RPV level drop as indicated by ALL of the following conditions being met:

• RCS/RPV level band is established below the RPV flange.

• RCS/RPV water level drops below the RCS level band for 15 minutes or longer OR CU2.3 RCS/RPV level cannot be monitored with a loss of RCS/RPV inventory as indicated by an unexplained level rise in Drywell/Reactor Building Equipment or Floor Drain sump, or Suppression Pool.

DAEC EAL INFORMATION:

This IC is included as a NOUE because it may be a precursor of more serious conditions and, as result, is considered to be a potential degradation of the level of safety of the plant. Refueling evolutions that decrease RCS water level below the RPV flange are carefully planned and procedurally controlled. An UNPLANNED event that CU2

EAL BASES DOCUMENT' EBD-C Rev. 1 next COLD SHUTDOWN/REFUELING - SYSTEMS MALFUNCTION Page 5 of 28 results in water level decreasing below the RPV flange, or below the planned RCS water level for the given evolution (if the RCS water level is already below the RPV flange), warrants declaration of a NOUE due to the reduced RCS inventory that is available to keep the core covered. The allowance of 15 minutes was chosen because it is reasonable to assume that level can be restored within this time frame using one or more of the redundant means of refill that should be available. If level cannot be restored in this time frame then it may indicate a more serious condition exists.

Continued loss of RCS Inventory will result in escalation to the Alert level via either IC CA2 or CA4.

The difference between CU1 and CU2 deals with the RCS conditions that exist between cold shutdown and refueling modes. In cold shutdown the RCS will normally be intact and standard RCS inventory and level monitoring means are available. In the refueling mode the RCS is not intact and RPV level and inventory are monitored by different means.

EAL 1 involves a decrease in RCS level below the top of the RPV flange that continues for 15 minutes due to an UNPLANNED event. This EAL is not applicable to decreases in flooded reactor cavity level (covered by RU2 ) until such time as the level decreases to the level of the vessel flange. If RPV level continues to decrease and reaches the Low-Low ECCS Actuation Setpoint then escalation to CA1 would be appropriate.

In the refueling mode, normal means of core temperature indication and RCS level indication may not be available. Redundant means of RPV level indication will normally be installed (including the ability to monitor level visually) to assure that the ability to monitor level will not be interrupted. However, if all level indication were to be lost during a loss of RCS inventory event, the operators would need to determine that RPV inventory loss was occurring by observing sump and Suppression Pool level changes.

The drywell floor and equipment drain sumps, reactor building equipment and floor drain sumps receive all liquid waste from floor and equipment drains inside the primary containment and reactor building. A rise in Suppression Pool water level may be indicative of valve misalignment or leakage in systems that discharge to the Torus.

Sump and Suppression Pool level increases must be evaluated against other potential sources of leakage such as cooling water sources inside the containment to ensure they are indicative of RCS leakage. Escalation to Alert would be via either CA1 or RCS heatup via CA4.

REFERENCES:

CU2

EAL BASES DOCUMENT EBD-C Rev. 1 next COLD SHUTDOWN/REFUELING - SYSTEMS MALFUNCTION Page 6 of 28

1. NEI 99-01 Rev. 5, NEI Methodology for Development of Emergency Action Levels CU2

EAL BASES DOCUMENT EBD-C Rev. 1 next COLD SHUTDOWN/REFUELING - SYSTEMS MALFUNCTION Page 7 of 28 CU3 AC power capability to essential busses reduced to a single power source for 15 minutes or longer such that any additional single failure would result in station blackout EVENT TYPE: Loss of Power OPERATING MODE APPLICABILITY: Cold S/D, Refueling EAL THRESHOLD VALUE:

Note: The Emergency Director (OSM, EC or ER&RD) 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.

CU3.1 AC power capability to 1A3 or 1A4 busses reduced to a single power source for 15 minutes or longer AND Any additional single power source failure will result in station blackout.

DAEC EAL INFORMATION:

Prolonged loss of AC power reduces required redundancy and potentially degrades the level of safety of the plant by rendering the plant more vulnerable to a complete Loss of AC Power (e.g., Station Blackout). Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. Subsequent loss of the single power source would escalate the event to an Alert via CA3.

The DAEC EAL is written to address the underlying concern, i.e., only one AC power source remains and if it is lost, a Station Blackout will occur. Under the conditions of concern, entry into AOP 301, Loss of Essential Electrical Power, would be made under Tab 1, Loss of One Essential 4160V Bus, and/or under Tab 3, Loss of Offsite Power.

Indications/alarms related to degraded AC power are displayed on control room panel 1C08 and are listed in AOP 301 under "Probable Indications."

At DAEC, the Essential Busses of concern are 4160V Busses 1A3 and 1A4. Each of these busses feed their associated 480V and 120V AC busses through step down CU3

BASES DOCUMENT AL EBD-C Rev. 1 next COLD SHUTDOWN/REFUELING - SYSTEMS MALFUNCTION Page 8 of 28 transformers. Onsite power sources at DAEC include the A and B Diesel Generators, 1G-31 and 1G-21, respectively.

REFERENCES:

1. Abnormal Operating Procedure (AOP) 301, Loss of Essential Electrical Power

2. UFSAR Section 8.2, Offsite Power System

3. NEI 99-01 Rev. 5, NEI Methodology for Development of Emergency Action Levels CU3

EAL BASES DOCUMENT EBD-C Rev. 1 next COLD SHUTDOWN/REFUELING - SYSTEMS MALFUNCTION Page 9 of 28 CU4 UNPLANNED loss of decay heat removal capability with irradiated fuel in the RPV EVENT TYPE: RCS Temperature OPERATING MODE APPLICABILITY: Cold S/D, Refueling EAL THRESHOLD VALUE:

Note: The Emergency Director (OSM, EC or ER&RD) 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.

CU4.1 An UNPLANNED event results in RCS temperature GREATER THAN 212 IF OR CU4.2 Loss of ALL RCS temperature and RCS/RPV level indication for 15 minutes or longer.

DAEC EAL INFORMATION:

This IC is included as a NOUE because it may be a precursor of more serious conditions and, as a result, is considered to be a potential degradation of the level of safety of the plant. In cold shutdown the ability to remove decay heat relies primarily on forced cooling flow. Operation of the systems that provide this forced cooling may be jeopardized due to the unlikely loss of electrical power or RCS inventory. Since the RCS usually remains intact in the cold shutdown mode a large inventory of water is available to keep the core covered. In cold shutdown the decay heat available to raise RCS temperature during a loss of inventory or heat removal event may be significantly greater than in the refueling mode. Entry into cold shutdown conditions may be attained within hours of operating at power. Entry into the refueling mode procedurally may not occur for typically 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> or longer after the reactor has been shutdown. Thus the heatup threat and therefore the threat to damaging the fuel clad may be lower for events that occur in the refueling mode with irradiated fuel in the RPV (note that the heatup threat could be lower for cold shutdown conditions if the entry into cold shutdown was following a refueling). In addition, the operators should be able to monitor RCS CU4

EBD-C Rev. 1 next COLD SHUTDOWN/REFUELING - SYSTEMS MALFUNCTION Page 10 of 28 temperature and RPV level so that escalation to the alert level via CA4 or CA1 will occur if required.

During refueling the level in the RPV will normally be maintained above the RPV flange.

Refueling evolutions that decrease water level below the RPV flange are carefully planned and procedurally controlled. Loss of forced decay heat removal at reduced inventory may result in more rapid increases in RCS/RPV temperatures depending on the time since shutdown. Escalation to the Alert level via CA4 is provided should an UNPLANNED event result in RCS temperature exceeding the Technical Specification cold shutdown temperature limit with CONTAINMENT CLOSURE not established.

Unlike the cold shutdown mode, normal means of core temperature indication and RCS level indication may not be available in the refueling mode. Redundant means of RPV level indication are therefore procedurally installed to assure that the ability to monitor level will not be interrupted. However, if all level and temperature indication were to be lost in either the cold shutdown or refueling modes, EAL 2 would result in declaration of a NOUE if either temperature or level indication cannot be restored within 15 minutes from the loss of both means of indication. Escalation to Alert would be via CA1 based on an inventory loss or CA4 based on exceeding its temperature criteria.

The Emergency Director must remain attentive to events or conditions that lead to the conclusion that exceeding the EAL threshold is imminent. If, in the judgment of the Emergency Director, an imminent situation is at hand, the classification should be made as if the threshold has been exceeded.

REFERENCES:

1. NEI 99-01 Rev. 5, NEI Methodology for Development of Emergency Action Levels CU4

EAL BASES.DOCUMENT EBD-C Rev. 1 next COLD SHUTDOWN/REFUELING - SYSTEMS MALFUNCTION Page 11 of 28 CU6 Loss of all onsite or offsite communications capabilities EVENT TYPE: Communication OPERATING MODE APPLICABILITY: Cold S/D, Refueling, Defueled EAL THRESHOLD VALUE:

CU6.1 Loss of ALL of the following onsite communication methods affecting the ability to perform routine operations:

• Plant Operations Radio System

• In-Plant Telephones

• Plant Paging System OR CU6.2 Loss of ALL of the following offsite communication methods affecting the ability to perform offsite notifications:

" All telephone lines (commercial)

" Cell phones (including fixed cell phone system)

• Control Room fixed satellite phones

" Microwave Phone System

" Emergency Notification System (ENS)

• FTS Phone System DAEC EAL INFORMATION:

The purpose of this IC and its associated EALs is to recognize a loss of communications capability that either defeats the plant operations staff ability to perform routine tasks necessary for plant operations or the ability to communicate problems with offsite authorities. The loss of offsite communications ability is expected to be significantly more comprehensive than the condition addressed by 10 CFR 50.72.

CU6

EAL..BASESDOCUMENT EBD-C Rev. 1 next COLD SHUTDOWN/REFUELING - SYSTEMS MALFUNCTION Page 12 of 28 The availability of one method of ordinary offsite communications is sufficient to inform state and local authorities of plant problems. This EAL is intended to be used only when extraordinary means (e.g., use of personal cell phones, relaying of information from radio transmissions, individuals being sent to offsite locations, etc.) are being utilized to make communications possible.

REFERENCES:

1. Emergency Plan, Section F, Emergency Communications

2. Abnormal Operating Procedure (AOP) 399, Loss of Communication

3. NEI 99-01 Rev. 5, NEI Methodology for Development of Emergency Action Levels CU6

EAL BASES DOCUMENT EBD-C Rev. 1 next COLD SHUTDOWN/REFUELING - SYSTEMS MALFUNCTION Page 13 of 28 CU7 Loss of required 125 VDC power for 15 minutes or longer EVENT TYPE: Loss of Power OPERATING MODE APPLICABILITY: Cold S/D, Refueling EAL THRESHOLD VALUE:

Note: The Emergency Director (OSM, EC or ER&RD) 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.

CU7.1 LESS THAN 105 VDC indicated on required vital DC busses for 15 minutes or longer.

DAEC EAL INFORMATION:

The purpose of this IC and its associated EALs is to recognize a loss of DC power compromising the ability to monitor and control the removal of decay heat during Cold Shutdown or Refueling operations. This EAL is intended to be anticipatory in as much as the operating crew may not have necessary indication and control of equipment needed to respond to the loss.

Routinely plants will perform maintenance on a Train related basis during shutdown periods. It is intended that the loss of the required (operable) train is to be considered.

If this loss results in the inability to maintain cold shutdown, the escalation to an Alert will be per CA4.

Bus voltage should be based on the minimum bus voltage necessary for the operation of safety related equipment. This voltage value should incorporate a margin of at least 15 minutes of operation before the onset of inability to operate those loads. This voltage is usually near the minimum voltage selected when battery sizing is performed.

Typically the value for the entire battery set is approximately 105 VDC.

REFERENCES:

1. NEI 99-01 Rev. 5, NEI Methodology for Development of Emergency Action Levels CU7

EAL BASES DOCUMENT. EBD-C Rev. 1 next COLD SHUTDOWN/REFUELING - SYSTEMS MALFUNCTION Page 14 of 28 CU8 Inadvertent criticality EVENT TYPE: Inadvertent Criticality OPERATING MODE APPLICABILITY: Cold S/D, Refueling EAL THRESHOLD VALUE:

CU8.1 UNPLANNED sustained positive period observed on nuclear instrumentation.

DAEC EAL INFORMATION:

This IC addresses criticality events that occur in Cold Shutdown or Refueling modes (NUREG 1449, Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the United States) such as fuel misloading events and inadvertent dilution events. This IC indicates a potential degradation of the level of safety of the plant, warranting a NOUE classification.

This condition can be identified using period monitors. The terms "extended" is used in order to allow exclusion of expected short term positive periods from planned fuel bundle or control rod movements during core alteration for BWRs. These short term positive periods are the result of the increase in neutron population due to subcritical multiplication.

Escalation would be by Emergency Director Judgment.

REFERENCES:

1. NEI 99-01 Rev. 5, NEI Methodology for Development of Emergency Action Levels CU8

EAL'BASES DOCUMENT EBD-C Rev. 1 next COLD SHUTDOWN/REFUELING - SYSTEMS MALFUNCTION Page 15 of 28 CAI Loss of RCS/RPV inventory EVENT TYPE: RCS Level OPERATING MODE APPLICABILITY: Cold S/D, Refueling EAL THRESHOLD VALUE:

Note: The Emergency Director (OSM, EC or ER&RD) 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.

CA1.1 Loss of RCS/RPV inventory as indicated by RPV level LESS THAN 119.5 inches.

OR CA1.2 RCS/RPV level cannot be monitored for 15 minutes or longer with a loss of RCS/RPV inventory as indicated by an unexplained level rise in Drywell/Reactor Building Equipment or Floor Drain sump, or Suppression Pool.

DAEC EAL INFORMATION:

These example EALs serve as precursors to a loss of ability to adequately cool the fuel.

The magnitude of this loss of water indicates that makeup systems have not been effective and may not be capable of preventing further RPV level decrease and potential core uncovery. This condition will result in a minimum classification of Alert. The inability to restore and maintain level after reaching 119.5 inches would therefore be indicative of a failure of the RCS barrier.

In cold shutdown the decay heat available to raise RCS temperature during a loss of inventory or heat removal event may be significantly greater than in the refueling mode.

Entry into cold shutdown conditions may be attained within hours of operating at power or hours after refueling is completed. Entry into the refueling mode procedurally may not occur for typically 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> or longer after the reactor has been shutdown. Thus the heatup threat and therefore the threat to damaging the fuel clad may be lower for events that occur in the refueling mode with irradiated fuel in the RPV (note that the CA1

EAL BASESDOCUMENT EBD-C Rev. 1 next COLD SHUTDOWN/REFUELING - SYSTEMS MALFUNCTION Page 16 of 28 heatup threat could be lower for cold shutdown conditions if the entry into cold shutdown was following a refueling). The above forms the basis for needing both a cold shutdown specific IC (CA1) and a refueling specific IC (CA2).

In the cold shutdown mode, normal RCS level and RPV level instrumentation systems will normally be available. In the refueling mode, normal means of RPV level indication may not be available. Redundant means of RPV level indication will be normally installed (including the ability to monitor level visually) to assure that the ability to monitor level will not be interrupted. However, if all level indication were to be lost during a loss of RCS inventory event, the operators would need to determine that RPV inventory loss was occurring by observing sump and Suppression Pool level changes.

The drywell floor and equipment drain sumps, reactor building equipment and floor drain sumps receive all liquid waste from floor and equipment drains inside the primary containment and reactor building. A rise in Suppression Pool water level may be indicative of valve misalignment or leakage in systems that discharge to the Torus.

Sump and Suppression Pool level increases must be evaluated against other potential sources of leakage such as cooling water sources inside the containment to ensure they are indicative of RCS leakage. The 15-minute duration for the loss of level indication was chosen because it is half of the CS1 Site Area Emergency EAL duration. The 15-minute duration allows CA1 to be an effective precursor to CS1. Significant fuel damage is not expected to occur until the core has been uncovered for greater than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> per the analysis referenced in the CG1 basis. Therefore this EAL meets the definition for an Alert emergency.

The difference between CA1 and CA2 deals with the RCS conditions that exist between cold shutdown and refueling mode applicability. In cold shutdown the RCS will normally be intact and standard RCS inventory and level monitoring means are available. In the refueling mode the RCS is not intact and RPV level and inventory are monitored by different means.

If RPV level continues to decrease then escalation to Site Area will be via CS1.

REFERENCES:

1. NEI 99-01 Rev. 5, NEI Methodology for Development of Emergency Action Levels CA1

EAL BASES DOCUMENT EBD-C Rev. 1 next COLD SHUTDOWN/REFUELING - SYSTEMS MALFUNCTION Page 17 of 28 CA3 Loss of all offsite and all onsite AC power to essential busses for 15 minutes or longer EVENT TYPE: Loss of Power OPERATING MODE APPLICABILITY: Cold S/D, Refueling, Defueled EAL THRESHOLD VALUE:

Note: The Emergency Director (OSM, EC or ER&RD) 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.

CA3.1 Loss of all offsite and all onsite AC power to 1A3 and 1A4 for 15 minutes or longer.

DAEC EAL INFORMATION:

Loss of all AC power compromises all plant safety systems requiring electric power including RHR, ECCS, Containment Heat Removal, Spent Fuel Heat Removal and the Ultimate Heat Sink. When in cold shutdown or refueling mode the event can be classified as an Alert, because of the significantly reduced decay heat, lower temperature and pressure, increasing the time to restore one of the emergency busses, relative to that specified for the Site Area Emergency EAL. Escalating to Site Area Emergency, if appropriate, is by Abnormal Rad Levels / Radiological Effluent ICs.

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

Consideration should be given to operable loads necessary to remove decay heat or provide Reactor Vessel makeup capability when evaluating loss of AC power to essential busses. Even though an essential bus may be energized, if necessary loads (i.e., loads that if lost would inhibit decay heat removal capability or Reactor Vessel makeup capability) are not operable on the energized bus then the bus should not be considered operable.

REFERENCES:

CA3

EAL BASES DOCUMENT. EBD-C Rev. 1 next COLD SHUTDOWN/REFUELING - SYSTEMS MALFUNCTION Page 18 of 28

1. NEI 99-01 Rev. 5, NEI Methodology for Development of Emergency Action Levels CA3

EAL BASES DOCUMENT EBD-C Rev. 1 next COLD SHUTDOWN/REFUELING - SYSTEMS MALFUNCTION Page 19 of 28 CA4 Inability to maintain plant in cold shutdown EVENT TYPE: RCS Temperature OPERATING MODE APPLICABILITY: Cold S/D, Refueling EAL THRESHOLD VALUE:

CA4.1 With Secondary Containment and RCS integrity NOT established, an UNPLANNED event results in RCS temperature GREATER THAN 212 OF OR CA4.2 With Secondary Containment established and RCS integrity NOT established, an UNPLANNED event results in RCS temperature GREATER THAN 212 OF for 20 minutes or longer. Note: If an RCS heat removal system is in operation within the specified time frames and RCS temperatureis being reduced then this EAL is not applicable.

OR CA4.3 With RCS integrity established, an UNPLANNED event results in RCS temperature GREATER THAN 212 IF for 60 minutes or longer. Note: If an RCS heat removal system is in operationwithin the specified time frames and RCS temperatureis being reduced then this EAL is not applicable.

OR CA4.4 An UNPLANNED event results in an RCS pressure rise GREATER THAN 10 psig due to a loss of RCS cooling.

DAEC EAL INFORMATION:

If an UNPLANNED event results in RCS temperature GREATER THAN 212 IF, use the table to select the appropriate EAL.

RCS Integrity Secondary Containment Duration temperature EAL GREATER THAN 212 OF Not Established Not Established 0 minutes CA4.1 Not Established Established 20 minutes CA4.2 Established NIA 60 minutes CA4.3 CA4

EAL BASES DOCUMENT EBD-C Rev. 1 next COLD SHUTDOWN/REFUELING - SYSTEMS MALFUNCTION Page 20 of 28 EAL 1 addresses complete loss of functions required for core cooling during refueling and cold shutdown modes when neither Secondary Containment nor RCS integrity are established. RCS integrity is in place 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). No delay time is allowed for EAL1 because the evaporated reactor coolant that may be released into the Containment during this heatup condition could also be directly released to the environment.

EAL 2 addresses the complete loss of functions required for core cooling for > 20 minutes during refueling and cold shutdown modes when Secondary Containment is established but RCS integrity is not established or RCS inventory is reduced. As in EAL 1, RCS integrity should be assumed to be in place 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). The allowed 20 minute time frame was included to allow operator action to restore the heat removal function, if possible. The allowed time frame is consistent with the guidance provided by Generic Letter 88-17, "Loss of Decay Heat Removal" (discussed later in this basis) and is believed to be conservative given that a low pressure Containment barrier to fission product release is established. EAL 2 is not applicable if actions are successful in restoringan RCS heat removal system to operation and RCS temperatureis being reduced within the 20 minute time frame.

EAL 3 addresses complete loss of functions required for core cooling for > 60 minutes during refueling and cold shutdown modes when RCS integrity is established. As in EAL 1 and 2, RCS integrity should be considered to be in place 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). The status of Secondary Containment in this EAL is immaterial given that the RCS is providing a high pressure barrier to fission product release to the environment. The 60 minute time frame should allow sufficient time to restore cooling without there being a substantial degradation in plant safety. EAL 3 is not applicableif actions are successful in restoringan RCS heat removal system to operation and RCS temperature is being reduced within the 60 minute time frame assuming that the RCS pressure increase has remainedless than the site specific pressure value.

For EAL 4, the 10 psi pressure increase addresses situations where, due to high decay heat loads, the time provided to restore temperature control, should be less than 60 CA4

EAL BASES DOCUMENT' EBD-C Rev. 1 next COLD SHUTDOWN/REFUELING - SYSTEMS MALFUNCTION Page 21 of 28 minutes. The RCS pressure setpoint chosen is 10 psig since this is the lowest value that can be read on installed Control Board instrumentation.

Escalation to Site Area would be via CS1 should boiling result in significant RPV level loss leading to core uncovery.

A loss of Technical Specification components alone is not intended to constitute an Alert. The same is true of a momentary UNPLANNED excursion above 212 IF when the heat removal function is available.

The Emergency Director must remain alert to events or conditions that lead to the conclusion that exceeding the EAL threshold is IMMINENT. If, in the judgment of the Emergency Director, an IMMINENT situation is at hand, the classification should be made as if the threshold has been exceeded.

REFERENCES:

1. NEI 99-01 Rev. 5, NEI Methodology for Development of Emergency Action Levels

2. NEP 2004-0034, EAL Submittal - Containment Pressure Indicator Justification CA4

EAL BASES DOCUMENT EBD-C Rev. 1 next COLD SHUTDOWN/REFUELING - SYSTEMS MALFUNCTION Page 22 of 28 CS1 Loss of RCS/RPV Inventory affecting core decay heat removal capability EVENT TYPE: RCS Level OPERATING MODE APPLICABILITY: Cold S/D, Refueling EAL THRESHOLD VALUE:

Note: The Emergency Director (OSM, EC or ER&RD) 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.

CS1.1 With Secondary Containment NOT established, RPV level LESS THAN 113.5 inches.

OR CS1.2 With Secondary Containment established, RPV level LESS THAN +15 inches.

OR CS1.3 RPV level cannot be monitored for 30 minutes or longer with a loss of RCS/RPV inventory as indicated by ANY of the following:

• Containment High Range Rad Monitor reading GREATER THAN 10 Rem/hr

" Erratic Source Range Monitor Indication

" Unexplained level rise in Drywell/Reactor Building Equipment or Floor Drain sump, or Suppression Pool DAEC EAL INFORMATION:

Under the conditions specified by this IC, continued decrease in RPV level is indicative of a loss of inventory control. Inventory loss may be due to an RPV breach, pressure boundary leakage, or continued boiling in the RPV. Since BWRs have RCS CS1

EBD-C Rev. 1 next COLD SHUTDOWN/REFUELING - SYSTEMS MALFUNCTION Page 23 of 28 penetrations below the setpoint, continued level decrease may be indicative of pressure boundary leakage.

Escalation to a General Emergency is via CG1 or RG1.

In cold shutdown the decay heat available to raise RCS temperature during a loss of inventory or heat removal event may be significantly greater than in the refueling mode.

Entry into cold shutdown conditions may be attained within hours of operating at power or hours after refueling is completed. Entry into the refueling mode procedurally may not occur for typically 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> or longer after the reactor has been shutdown. Thus the heatup threat and therefore the threat to damaging the fuel clad may be lower for events that occur in the refueling mode with irradiated fuel in the RPV (note that the heatup threat could be lower for cold shutdown conditions if the entry into cold shutdown was following a refueling).

For EAL 3 in the refueling mode, normal means of RPV level indication may not be available. Redundant means of RPV level indication will be normally installed (including the ability to monitor level visually) to assure that the ability to monitor level will not be interrupted.

The 30-minute duration allows sufficient time for actions to be performed to recover needed cooling equipment and is considered to be conservative.

As water level in the RPV lowers, the dose rate above the core will increase. The dose rate due to this core shine will result in significantly increased Containment High Range Radiation Monitor readings. An unexplained reading of greater than 10 Rem/hr may be indicative of fuel damage. The basis for 10 Rem/hr is that it is sufficiently above the normal shutdown levels to avoid an unnecessary entry into the EAL. The 10 Rem/hr is also well below the containment radiation monitor reading of 2E+2 R/hr that would be indicative of 1% clad failure found in the following calculation:

Calculation of Drywell Radiation Monitor Reading Assuming 1% Gap Release NG-88-0966 value 20% Gap Release at 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for drywell = 2.9E+3Rem/hr Drywell reading = 2.9E+3Rem/hr x [1 % / 20 %] = 1.45E+2 Rem/hr, round off as 2E+2 Rem/hr CS1

EAL BASES DOCUMENT EBD-C Rev. 1 next COLD SHUTDOWN/REFUELING - SYSTEMS MALFUNCTION Page 24 of 28 Post-TMI studies indicated that the installed nuclear instrumentation will operate erratically when the core is uncovered and that this should be used as a tool for making such determinations.

In the cold shutdown mode, normal RCS level indication systems will normally be available. However, if all level indication were to be lost during a loss of RCS inventory event, the operators would need to determine that RPV inventory loss was occurring by observing sump and Suppression Pool level changes. The drywell floor and equipment drain sumps, reactor building equipment and floor drain sumps receive all liquid waste from floor and equipment drains inside the primary containment and reactor building. A rise in Suppression Pool water level may be indicative of valve misalignment or leakage in systems that discharge to the Torus. Sump and Suppression Pool level increases must be evaluated against other potential sources of leakage such as cooling water sources inside the containment to ensure they are indicative of RCS leakage.

REFERENCES:

1. NEI 99-01 Rev. 5, NEI Methodology for Development of Emergency Action Levels CS1

EAL BASES DOCUMENT EBD-C Rev. 1 next COLD SHUTDOWN/REFUELING - SYSTEMS MALFUNCTION Page 25 of 28 CGI Loss of RCS/RPV inventory affecting fuel clad integrity with containment challenged EVENT TYPE: Inability to Reach or Maintain Shutdown Conditions OPERATING MODE APPLICABILITY: Cold S/D, Refueling EAL THRESHOLD VALUE:

Note: The Emergency Director (OSM, EC or ER&RD) 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.

CG1.1 RPV level LESS THAN +15 inches for 30 minutes or longer with irradiated fuel in the RPV AND ANY Secondary Containment challenge indication (see Table):

OR CG1.2 RPV level cannot be monitored for 30 minutes or longer AND ANY of the following indications are present (indicating core uncovery):

• Containment High Range Rad Monitor reading GREATER THAN 10 Rem/hr.

• Erratic source range monitor indication

• UNPLANNED level rise in Drywell/Reactor Building Equipment or Floor Drain sump, or Torus AND ANY Secondary Containment challenge indication (see Table):

CG1

EAL BASES DOCUMENT EBD-C Rev. 1 next COLD SHUTDOWN/REFUELING - SYSTEMS MALFUNCTION Page 26 of 28 Table: Secondary Containment Challenge Indications

• Secondary Containment NOT established

• Drywell Hydrogen or Torus Hydrogen GREATER THAN 6% AND Drywell Oxygen or Torus Oxygen GREATER THAN 5%

• UNPLANNED rise in containment pressure

• Two or more Reactor Building areas exceed Max Safe Radiation Levels DAEC EAL INFORMATION:

These EALs represent the inability to restore and maintain RPV level to above the top of irradiated active fuel. Fuel damage is probable if RPV level cannot be restored, as available decay heat will cause boiling, further reducing the RPV level. With Secondary Containment breached or challenged then the potential for unmonitored fission product release to the environment is high. This represents a direct path for radioactive inventory to be released to the environment. This is consistent with the definition of a GE. The GE is declared on the occurrence of the loss or IMMINENT loss of function of all three barriers.

For EAL 2 in the cold shutdown mode, normal RCS level and RPV level instrumentation systems will normally be available. In the refueling mode, normal means of RPV level indication may not be available and redundant means of RPV level indication will be normally installed (including the ability to monitor level visually) to assure that the ability to monitor level will not be interrupted. However, if all level indication were to be lost during a loss of RCS inventory event, the operators would need to determine that RPV inventory loss was occurring by observing sump and Torus level changes. Sump and Torus level increases must be evaluated against other potential sources of leakage such as cooling water sources inside the containment to ensure they are indicative of RCS leakage.

These example EALs are based on 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 Operationat Commercial Nuclear Power Plantsin the United States, and, NUMARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management.

CG1

EAL BASES DOCUMENT EBD-C Rev. 1 next COLD SHUTDOWN/REFUELING - SYSTEMS MALFUNCTION Page 27 of 28 A number of variables, (BWRs - e.g., such as initial vessel level, or shutdown heat removal system design) can have a significant impact on heat removal capability challenging the fuel clad barrier. Analysis in the above references indicates that core damage may occur within an hour following continued core uncovery therefore, conservatively, 30 minutes was chosen. If CONTAINMENT CLOSURE is re-established prior to exceeding the 30 minute core uncovery time limit, then escalation to GE would not occur.

For EAL 2, as water level in the RPV lowers, the dose rate above the core will increase.

The dose rate due to this core shine will result in significantly increased Containment High Range Radiation Monitor readings. An unexplained reading of greater than 10 Rem/hr may be indicative of fuel damage. The basis for 10 Rem/hr is that it is sufficiently above the normal shutdown levels to avoid an unnecessary entry into the EAL. The 10 Rem/hr is also well below the containment radiation monitor reading of 2E+2 R/hr that would be indicative of 1% clad failure found in the following calculation:

Calculation of Drywell Radiation Monitor Reading Assuming 1% Gap Release NG-88-0966 value 20% Gap Release at 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for drywell = 2.9E+3Rem/hr Drywell reading = 2.9E+3Rem/hr x [1 % / 20 %] = 1.45E+2 Rem/hr, round off as 2E+2 Rem/hr Based on the above discussion, RCS barrier failure resulting in core uncovery for 30 minutes or more may cause fuel clad failure.

Secondary Containment closure is the action taken to secure containment and its associated structures, systems, and components as a functional barrier to fission product release under existing plant conditions. Secondary Containment should not be confused with refueling containment integrity as defined in technical specifications. Site shutdown contingency plans typically provide for re-establishing Secondary Containment following a loss of heat removal or RCS inventory functions. If Secondary Containment is re-established prior to exceeding the temperature or level thresholds of the RCS Barrier and Fuel Clad Barrier EALs, escalation to GE would not occur.

For BWRs, the use of Secondary Containment radiation monitors should provide indication of increased release that may be indicative of a challenge to secondary containment. The site-specific radiation monitor values should be based on the EOP CGI

EAL BASES DOCUMENT, EBD-C Rev. 1 next COLD SHUTDOWN/REFUELING - SYSTEMS MALFUNCTION Page 28 of 28 "maximum safe values" because these values are easily recognizable and have an emergency basis.

In the early stages of a core uncovery event, it is unlikely that hydrogen buildup due to a core uncovery could result in an explosive mixture of dissolved gasses in Secondary Containment. However, Secondary Containment monitoring and/or sampling should be performed to verify this assumption and a General Emergency declared if it is determined that an explosive mixture exists.

REFERENCES:

1. NEI 99-01 Rev. 5, NEI Methodology for Development of Emergency Action Levels CG1

EAL BASES DOCUMENT EBD-E Rev. 2 next ISFSI ABNORMAL EVENTS CATEGORY PAGE 1 of 3 I os,. ,I Usage Level INFORMATION Effective Date:

STECHNICAL REVIEW, Prepared and Verified by: Date:

Validated by: Date:

Emergency Planning Staff

'Si PROCEDURE APPROVAL I am responsible for the technical content of this procedure.

Approved by: Date:

Manager, Emergency Planning

EAL BASES DOCUMENT EBD-E Rev. 2 next ISFSI ABNORMAL EVENTS CATEGORY PAGE 2 of 3 EU1 Damage to a loaded cask CONFINEMENT BOUNDARY EVENT TYPE: Independent Spent Fuel Storage Installation (ISFSI)

OPERATING MODE APPLICABILITY: Not Applicable EAL THRESHOLD VALUE:

EU 1.1 Damage to the Dry Shielded Canister of a loaded cask.

DAEC EAL INFORMATION:

Security related events for the ISFSI are to be covered under the Security EALs in the Hazard recognition category.

The CONFINEMENT BOUNDARY is the barrier (Dry Shielded Canister (DSC)) that separates areas containing radioactive substances, spent nuclear fuel or high-level waste, and the environment.

A NOUE in this IC is categorized on the basis of the occurrence of an event of sufficient magnitude that a loaded cask CONFINEMENT BOUNDARY is damaged or violated. This includes classification based on a loaded fuel storage cask CONFINEMENT BOUNDARY loss leading to degradation of the fuel during storage or posing an operational safety problem with respect to its removal from storage.

The results of the ISFSI Safety Analysis Report (SAR) per NUREG-1 536 or SAR referenced in the cask's Certificate of Compliance and the related NRC Safety Evaluation Report were used to develop the DAEC list of natural phenomena events and accident conditions. This EAL addresses responses to a dropped cask, a tipped over cask, or natural phenomena affecting a cask (e.g.,

seismic event, tornado, etc.) or a dropped cask. (Reference Action Request OTH026062 for credible and non-credible event analysis.)

During all packaging, transfer, and storage activities, the DSC is completely enclosed in one of two additional containers, the DSC transfer cask or the horizontal storage module, and is never exposed to the environment. Both of these devices provide physical missile protection and radiation shielding for the DSC.

Because the DSC is not directly accessible for visible inspection, damage to the DSC is defined as:

damage to the DSC transfer cask or horizontal storage module that is readily observable without measurements, testing, or analysis. Damage is sufficient to cause concern regarding the continued operability or reliability of the DSC inside the transfer cask or horizontal storage module. Example damage includes: deformation due to heat, impact, or unplanned movement, denting, penetration, EU1

EAL. BASES DOCUMENT EBD-E Rev. 2 next ISFSI ABNORMAL EVENTS CATEGORY PAGE 3 of 3 rupture, cracking, or spalling of concrete to expose concrete reinforcing bar, or reduction in depth or configuration of radiation shielding materials. Surface blemishes (e.g., paint fading, paint chipping, concrete cracks or scratches) are not included in visible damage.

REFERENCES:

1. "Methodology for Development of Emergency Action Levels," NEI 99-01 Revision 5, February 2008

2. NUREG-1536, "Standard Review Plan for Dry Cask Storage Systems"

3. SAR referenced in the cask's Certificate of Compliance and the related NRC Safety Evaluation Report

4. Action Request OTH026062 EU1

EA L BASES DOCUMENT EBD F Rev. 8 next FISSION PRODUCT BARRIER DEGRADATION PAGE 1 of 31 Usage Level INFORMATION Approved for 'Point-of-Use' printing IF NO Temporary Changes are in effect for this procedure.

(on designated printers)

Record the following: Date / Time: Initials:

NOTE: A check to ensure currentrevision and no temporary changes shall be performed and documented every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> if active document use exceeds a 24 hourperiod as determined from the date and time recorded above.

Prepared By: Date:

Print Signature ICROSS-DISCIPLINE REVIEW (AS REQUIRED)

Reviewed By: Date:

Print Signature Reviewed By: Date:

Print Signature PROCEDURE APPROVAL BY QUALIFIED REVIEWER~

Approved By Date:

Print Signature

EAL BASES DOCUMENT EBD F Rev. 8 next FISSION PRODUCT BARRIER DEGRADATION PAGE 2 of 31 FU1 ANY Loss or ANY Potential Loss of Primary Containment Barrier EVENT TYPE: See Fission Barrier Table OPERATING MODE APPLICABILITY: Power Operation, Startup, Hot Shutdown EAL Threshold Values:

See the Fission Barrier Table indicators discussed later in this section.

DAEC INFORMATION:

The entry conditions for this Initiating Condition are shown by the logic chart shown below.

REFERENCES:

See the Fission Barrier Table indicators discussed later in this section.

FUI

EAL BASES. DOCUMENT EBD F Rev. 8 next FISSION PRODUCT BARRIER DEGRADATION PAGE 3 of 31 FA1 ANY Loss or ANY Potential Loss of EITHER Fuel Clad OR RCS Barrier EVENT TYPE: See Fission Barrier Table OPERATING MODE APPLICABILITY: Power Operation, Startup, Hot Shutdown EAL Threshold Values:

See the Fission Barrier Table indicators discussed later in this section.

DAEC INFORMATION:

The entry conditions for this Initiating Condition are shown by the logic chart shown below.

REFERENCES:

See the Fission Barrier Table indicators discussed later in this section.

FA1

EAL BASES DOCUMENT EBDF Rev. 8 next FISSION PRODUCT BARRIER DEGRADATION PAGE 4 of 31 FS1 Loss or Potential Loss of ANY Two Barriers EVENT TYPE: See Fission Barrier Table OPERATING MODE APPLICABILITY: Power Operation, Startup, Hot Shutdown EAL Threshold Values:

See the Fission Barrier Table indicators discussed later in this section.

DAEC INFORMATION:

The entry conditions for this Initiating Condition are shown by the logic chart shown below.

REFERENCES:

See the Fission Barrier Table indicators discussed later in this section.

FS1

EAL BASES DOCUMENT EBD F Rev. 8 next FISSION PRODUCT BARRIER DEGRADATION PAGE 5 of 31 FG1 Loss of ANY Two Barriers AND Loss or Potential Loss of the Third Barrier EVENT TYPE: See Fission Barrier Table OPERATING MODE APPLICABILITY: Power Operation, Startup, Hot Shutdown EAL Threshold Values:

See the Fission Barrier Table indicators discussed later in this section.

DAEC INFORMATION:

The entry conditions for this Initiating Condition are shown by the logic chart shown below.

REFERENCES:

See the Fission Barrier Table indicators discussed later in this section.

FG1

EAL BASES DOCUMENT EBD F Rev. 8 next FISSION PRODUCT BARRIER DEGRADATION PAGE 6 of 31 FISSION BARRIER: Fuel Clad DAEC INDICATOR: Radiation/Core Damage EAL THRESHOLD VALUE:

Clad Damage Determination LOSS: Fuel damage assessment (PASAP 7.2) indicates at least 5% fuel clad damage.

POTENTIAL LOSS: None DAEC INFORMATION:

As a site-specific loss indicator, DAEC uses determination of at least 5% fuel clad damage, which is consistent with the containment rad monitor reading indicators. This can be determined per FUEL DAMAGE ASSESMENT, PASAP 7.2.

REFERENCES:

1. Post Accident Sampling and Analysis Procedure (PASAP) 7.2, Fuel Damage Assessment

2. NEI 99-01 Rev. 5, NEI Methodology for Development of Emergency Action Levels Fuel Clad Barrier Radiation/Core Damage

EAL BASES DOCUMENT EBD F Rev. 8 next FISSION PRODUCT BARRIER DEGRADATION PAGE 7 of 31 FISSION BARRIER: Fuel Clad DAEC INDICATOR: Radiation/Core Damage EAL THRESHOLD VALUE:

Drywell/Torus Radiation Monitoring LOSS: Drywell Area Hi Range Rad Monitor RIM-9184A or B reading GREATER THAN 700 Rem/hr OR LOSS: Torus Area Hi Range Rad Monitor RIM-9185A or B reading GREATER THAN 30 Rem/hr POTENTIAL LOSS: None DAEC INFORMATION:

Valid means that the reading is from instrumentation determined to be operable in accordance with the Technical Specifications or has been verified by other independent methods such as indications displayed on the control panels, reports from plant personnel, coolant sampling or radiological survey results.

There is no significant deviation from the generic "loss" indicator. Per NEI 99-01, the (site-specific) reading (Drywell/Torus Rad - above) is a value that indicates release into the drywell of reactor coolant with elevated activity corresponding to about 2% to 5% fuel clad damage. This activity level is well above that expected from iodine spiking. It is intended that determination of barrierloss be made whenever the indicatorthreshold is reached until such time that core damage assessment is performed, at which time direct use of containment rad monitor readings is no longerrequired.

As documented by NG-88-0966, General Electric performed a study to predict dose rate readings from fuel damage calculations for emergency planning. The calculations were performed to obtain gamma ray dose rates at the locations of the containment atmospheric monitoring system radiation detectors in the drywell and torus locations for assumed releases of gap activity from the core. These calculations were based on "nominal" estimates of fuel rod gap fission product inventory fractions, which are considered to be more appropriate for determining a minimum threshold reading than inventory assumptions found in the NRC Regulatory Guides. The Regulatory Guide inventory assumptions applicable to dose assessments are larger and therefore non-conservative for determination of this EAL threshold. Two separate cases were evaluated.

Fuel Clad Barrier RadiationlCore Damage

EAL BASES DOCUMENT EBD F Rev. 8 next FISSION PRODUCT BARRIER DEGRADATION PAGE 8 of 31 In the first case, the released activity was assumed to be contained in the drywell atmosphere. This case is considered representative of conditions following a line break in which activity is released directly into the drywell. In the second case, the released activity was assumed to be contained in the torus. This could be applied for an event which results in vessel isolation and blowdown to the suppression chamber. The results for each case were provided for each case in the form of gamma ray dose rate versus time profiles for assumed releases of 100% and 20% of the gap activity from the core. The dose rate calculations were carried out independent of any specific information on details of construction or response characteristics of the detector systems. The figures show a drywell reading of about 2.9 x 103 Rem/hr or a torus reading of about 1.1 x 102 Rem/hr associated with 20% gap release at two hours after shutdown. Scaling this down to 5%

gap release:

Calculation of Drywell and Torus Monitor Readings Assuming 5% Gap Release NG-88-0966 value 20% Gap Release at 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for drywell = 2.9 x 103 Rem/hr Drywell reading = 2.9 x 103 Rem/hr x [5 % / 20 %] = 7.25 x 102 Rem/hr, round off as 7 E+2 Rem/hr NG-88-0966 value 20% Gap Release at 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for torus = 1.1 x 102 Rem/hr Torus reading = 1.1 x 102 Rem/hr x [5 % /20 %] = 2.75 x 101 Rem/hr, round off as 3 E+1 Rem/hr The results are rounded off for ease of reading the respective radiation monitors' scales.

The two-hour point was picked because it allows ample time for the Technical Support Center to be operational and core damage assessment to begin. These indicators correspond to about 2.5% gap release ifthey occur immediately after shutdown. Thus, the indicators address the 2%-5% fuel clad damage range of concern described by the generic guidance.

REFERENCES:

1. Office Memo NG-88-0966, G.E. Fuel Damage Documentation/Dose Rate Calculations, 03/18/88

2. NEI 99-01 Rev. 5, NEI Methodology for Development of Emergency Action Levels Fuel Clad Barrier RadiationlCore Damage

EAL BASES DOCUMENT EBD F Rev. 8 next FISSION PRODUCT BARRIER DEGRADATION PAGE 9 of 31 FISSION BARRIER: Fuel Clad DAEC INDICATOR: Radiation/Core Damage EAL THRESHOLD VALUE:

Primary Coolant Activity Level LOSS: Coolant activity GREATER THAN 300 gtCi/gm dose equivalent 1-131.

POTENTIAL LOSS: None DAEC INFORMATION:

There is no significant deviation from the generic indicator. Consistent with the generic methodology, DAEC uses a coolant activity value of 300 [tCi/gm 1-131 equivalent. This value is well above that expected for iodine spikes and would indicate fuel clad damage has occurred.

REFERENCES:

1. Post Accident Sampling and Analysis Procedure (PASAP) 7.2, Fuel Damage Assessment

2. NEI 99-01 Rev. 5, NEI Methodology for Development of Emergency Action Levels Fuel Clad Barrier Radiation/Core Damage

EAL BASES DOCUMENT EBD F Rev. 8 next FISSION PRODUCT BARRIER DEGRADATION PAGE 10 of 31 FISSION BARRIER: Fuel Clad DAEC INDICATOR: RPV Level EAL THRESHOLD VALUE:

Reactor Vessel Water Level LOSS: RPV Level cannot be restored and maintained above -25 inches POTENTIAL LOSS: RPV Level cannot be restored and maintained above +15 inches or cannot be determined.

DAEC INFORMATION:

The loss indicator is based on a value that corresponds to the minimum value to assure core cooling without further degradation of the fuel clad. DAEC uses the Minimum Steam Cooling RPV Water Level of -25 inches. This is defined to be the lowest RPV water level at which the covered portion of the reactor core will generate sufficient steam to preclude any clad temperature in the uncovered portion of the core from exceeding 1500 0 F.

Consistent with the EOPs, an indicated RPV level below -25 inches is used.

The potential loss indicator corresponds to the water level at the top of the active fuel (TAF). Consistent with the EOPs, an indicated RPV level below +15 inches.

REFERENCES:

1. Emergency Operating Procedure (EOP)-1, RPV Control, Sheet 1 of 1

2. ATWS Emergency Operating Procedure (EOP)-RPV Control, Sheet 1 of 1

3. Emergency Operating Procedure (EOP) Basis, Curves and Limits, C5, Minimum Steam Cooling RPV Water Level

4. NEI 99-01 Rev. 5, NEI Methodology for Development of Emergency Action Levels Fuel Clad Barrier RPV Level

EAL BASES DOCUMENT . . EBDF Rev. 8 next FISSION PRODUCT BARRIER DEGRADATION PAGE 11 of 31 FISSION BARRIER: Fuel Clad DAEC INDICATOR: Emergency Director Judgment EAL THRESHOLD VALUE:

Emergency Director Judgment Any condition in the opinion of the Emergency Director (OSM, EC or ER&RD) that indicates LOSS or POTENTIAL LOSS of the Fuel Clad Barrier DAEC INFORMATION:

There is no significant deviation from the generic indicator.

Emergency Director considerations for determining whether any barrier "Loss" or "Potential Loss" include imminent barrier degradation, degraded barriermonitoring capability, and consideration of dominant accident sequences.

Any condition which in the judgment of the Emergency Director indicates a LOSS or POTENTIAL LOSS of the FUEL CLAD barrier such as, but not limited to:

• Degraded barriermonitoringcapability from loss of/lack of reliable indicators.

• Consideration for instrumentation operability.

• Portable instrumentation readings.

• Offsite monitoring results.

• Complete loss of 125 VDC.

• Loss of decay heat removal.

• ATWS with failure of Standby Liquid Control.

• Prolonged station blackout.

• Loss of offsite power with early HPCI/RCIC failure Imminent means that no turnaround in safety system performance is expected and that General Emergency conditions can be expected to occur within two hours. Imminent fission barrier degradation must be considered by the Emergency Director to assure timely declaration of a General Emergency and to better assure that offsite protective actions can be effectively accomplished.

Degraded barriermonitoringcapability from loss of/lack of reliable indicators must also be considered by the Emergency Director when determining if a fission barrier loss or potential loss has occurred.

Fuel Clad Barrier ED Judgment

EAL BASES DOCUMENT EBD F Rev. 8 next FISSION PRODUCT BARRIER DEGRADATION PAGE 12 of 31 This assessment should also include consideration for instrumentation operability and portable instrumentation readings.

Offsite monitoring results may be an indication of Fission Product Barrier degradation causing an unmonitored release.

Dominant accident sequences can lead to loss of all Fission Barriers. Based on the IPE, the dominant accident sequences leading to core damage at DAEC include complete loss of 125 VDC, loss of decay heat removal, ATWS with failure of Standby Liquid Control, prolonged station blackout, and loss of offsite power with early HPCI/RCIC failure. The Emergency Director should also consult System Malfunction EALs, as appropriate, to assure timely emergency classification declaration.

REFERENCES:

1. Emergency Plan Implementing Procedure (EPIP) 2.5, Control Room Emergency Response Operation

2. Duane Arnold Energy Center Individual Plant Examination (IPE) November 1992

3. NEI 99-01 Rev. 5, NEI Methodology for Development of Emergency Action Levels Fuel Clad Barrier ED Judgment

EAL BASES DOCUMENT EBD F Rev. 8 next FISSION PRODUCT BARRIER DEGRADATION PAGE 13 of 31 FISSION BARRIER: RCS DAEC INDICATOR: Radiation/Core Damage EAL THRESHOLD VALUE:

Drywell Radiation Monitoring LOSS: Drywell Area Hi Range Rad Monitor RIM-9184A or B reading GREATER THAN 5 Rem/hr after Reactor Shutdown POTENTIAL LOSS: None DAEC INFORMATION:

This loss indicator is based on conditions after reactor shutdown to assure that it is not misapplied, i.e., to exclude readings due to N-16 effects which are typically 5 to 8 Rem/hr at full power conditions.

The 5 Rem/hr value for this loss indicator corresponds to instantaneous release and dispersal of the reactor coolant noble gas and iodine inventory associated with normal operating concentrations (i.e., within Technical Specifications) into the drywell atmosphere.

The reading will be less than that specified for the loss indicator for Radiation/Core Damage that applies to the Fuel Clad barrier. Thus, this indicator would be indicative of a RCS leak only. Ifthe radiation monitor reading increased to that value specified by the Radiation/Core indicator applying to the Fuel Clad barrier, fuel damage would also be indicated.

As documented by NG-88-0966, General Electric performed a study to predict dose rate readings from fuel damage calculations for emergency planning. The calculations were performed to obtain gamma ray dose rates at the locations of the containment atmosphere monitoring system radiation detectors in the drywell and torus locations for assumed releases of gap activity from the core. These calculations were based on "nominal" estimates of fuel rod gap fission product inventory fractions, which are considered to be more appropriate for determining a minimum threshold reading than inventory assumptions found in the NRC Regulatory Guides. The Regulatory Guide inventory assumptions applicable to dose assessments are larger and therefore non-conservative for determination of this EAL threshold. Two separate cases were evaluated. In the first case, the released activity was assumed to be contained in the drywell atmosphere. This case is considered representative of conditions following a line break in which activity is released directly into the drywell. In the second case, the released activity was assumed to be contained in the torus. This could be applied for an event which results in vessel isolation and blowdown to the suppression chamber. The results for each case were RCS Barrier Radiation/Core Damage

EAL BASES'DOCUMENT EBD F Rev. 8 next FISSION PRODUCT BARRIER DEGRADATION PAGE 14 of 31 provided for each case in the form of gamma ray dose rate versus time profiles for assumed releases of 100% and 20% of the gap activity from the core. The dose rate calculations were carried out independent of any specific information on details of construction or response characteristics of the detector systems. The figures show a drywell reading of about 2.1 x 104 Rem/hr associated with a 100% gap release immediately after shutdown. Assuming 99.99% fuel clad integrity (0.01% gap release) and uniform dispersal of radionuclides into the drywell immediately after shutdown, a drywell monitor reading is calculated:

Calculation of Drywell Monitor Reading Assuming 0.01% Gap Release NG-88-0966 value for 100% Gap Release at 0.01 minutes = 2.1 x 104 Rem/hr (2.1 x 104 ) Rem/hr x [(1 x 10-2 ) percent / 100 percent] = (2.1) x 104-4 Rem/hr = 2.1 x 100 Rem/hr = 2 Rem/hr To assure an indicator that is readily discernible on the drywell radiation monitor scale, DAEC uses a valid reading above 5 Rem/hr after reactor shutdown.

REFERENCES:

1. Office Memo NG-88-0966, G.E. Fuel Damage Documentation/Dose Rate Calculations, 03/18/88

2. Technical Specification 3.4.5, Drywell Leak Detection Instrumentation

3. NEI 99-01 Rev. 5, NEI Methodology for Development of Emergency Action Levels RCS Barrier Radiation/Core Damage

EAL BASES DOCUMENT EBD F Rev. 8 next FISSION PRODUCT BARRIER DEGRADATION PAGE 15 of 31 FISSION BARRIER: RCS DAEC INDICATOR: RPV Level EAL THRESHOLD VALUE:

Reactor Vessel Water Level LOSS: RPV Level cannot be restored and maintained above +15 inches or cannot be determined POTENTIAL LOSS: None DAEC INFORMATION:

There is no significant deviation from the generic indicator. This loss indicator corresponds to the water level at the top of the active fuel (TAF). In order to provide normal means to cool the fuel, water level must be maintained above the top of active fuel otherwise extraordinary means must be taken to assure that adequate core cooling exists. In certain failure event sequences reactor vessel water level may be procedurally lowered to the top of active fuel and the reactor coolant system depressurized to allow for steam cooling of the core. Even though fuel clad damage is not predicted under these conditions several safety system failures need to have occurred to reach the condition where steam cooling would be procedurally required. Therefore this is indicative of a loss of the reactor coolant system boundary. Water levels below this value indicate a challenge to core cooling which is a precursor to more serious events.

REFERENCES:

1. Emergency Operating Procedures (EOP) Basis, Breakpoints

2. NEI 99-01 Rev. 5, NEI Methodology for Development of Emergency Action Levels RCS Barrier RPV Level

EAL BASES DOCUMENT, EBD F Rev. 8 next FISSION PRODUCT BARRIER DEGRADATION PAGE 16 of 31 FISSION BARRIER: RCS DAEC INDICATOR: Leakage EAL THRESHOLD VALUE:

RCS Leak Rate LOSS: UNISOLABLE Main Steamline, HPCI, Feedwater, RWCU or RCIC break as indicated by the failure of both isolation valves in any one line to close AND EITHER:

" High MSL flow or high steam tunnel temperature annunciators

" Direct report of steam release OR LOSS: Emergency RPV Depressurization is required.

POTENTIAL LOSS: RCS leakage GREATER THAN 50 GPM inside the drywell.

OR POTENTIAL LOSS: UNISOLABLE primary system leakage outside drywell as indicated by area temperatures or ARMs exceeding the Max Normal Limits per EOP 3, Table 6.

DAEC INFORMATION:

There are no significant deviations from the generic potential loss indicators applying to RCS leakage and indications of UNISOLABLE primary system leakage.

The EC/OSM should also consult SU5, RCS Leakage, to determine if RCS leakage exceeds the threshold required for declaration of an Unusual Event.

An UNISOLABLE MSL break is a breach of the RCS barrier. Thus, this EAL is included for consistency with the Alert emergency classification. Other large high-energy line breaks such as HPCI, Feedwater, RWCU and RCIC that are UNISOLABLE also represent a significant loss of the RCS barrier and are therefore considered as MSL breaks, for purposes of classification. This threshold is concerned with lines penetrating primary containment and have PCIS valves. Leakage inside primary containment is covered under thresholds for drywell leakage or drywell pressure.

RCS Barrier Leakage

EAL BASES DOCUMENT, EBID F Rev. 8 next FISSION PRODUCT BARRIER DEGRADATION PAGE 17 of 31 Plant symptoms requiring Emergency RPV Depressurization per the site specific EOPs are indicative of a loss of the RCS barrier. If Emergency RPV depressurization is required, the plant operators are directed to open safety relief valves (SRVs) and keep them open. Even though the RCS is being vented into the suppression pool, a loss of the RCS should be considered to exist due to the diminished effectiveness of the RCS pressure barrier to a release of fission products beyond its boundary.

If an SRV is stuck open or is cycling and no other emergency conditions exist, an emergency declaration may not be appropriate. RCS leakage inside the drywell excludes Safety-Relief Valve (SRV) dischargethrough the SRV dischargepiping into the torus below the water line. However, if the fuel is damaged and the SRV is allowing fission products to escape into primary containment, a loss of RCS should be determinedas having occurred.

UNISOLABLE primary system leakage is considered a Potential loss of RCS based on RCS leakage outside the drywell. Site-specific RCS leakage is determined from temperature or area radiation alarms (ARMs) exceeding the Max Normal limits listed in Table 6, EOP 3. UNISOLABLE primary system leakage in the areas of the steam tunnel, main turbine generator, RCIC, HPCI, etc., indicates a direct path from the RCS to areas outside primary containment. It should be confirmed that the indicators are caused by RCS leakage. Area temperatures or area radiation alarms above Max Normal limits are the criteria for declaration of an Alert classification. An UNISOLABLE leak which is indicated by exceeding Max Safe limits escalates to a Site Area Emergency when combined with Primary Containment Barrier loss (after a containment isolation) and a General Emergency when the Fuel Clad Barrier criteria is also exceeded.

REFERENCES:

1. Alarm Response Procedure (ARP) 1C04B, Reactor Water Cleanup and Recirculation

2. Alarm Response Procedure (ARP) 1C04C, Reactor Water Cleanup and Recirculation

3. Emergency Operating Procedure (EOP) 3, Secondary Containment Control

4. UFSAR Section 15.6.6, Loss-of-Coolant-Accident

5. NEI 99-01 Rev. 5, NEI Methodology for Development of Emergency Action Levels RCS Barrier Leakage

EAL BASES DOCUMENT EBD F Rev. 8 next FISSION PRODUCT BARRIER DEGRADATION PAGE 18 of 31 FISSION BARRIER: RCS DAEC INDICATOR: Primary Containment Atmosphere EAL THRESHOLD VALUE:

Drywell Pressure LOSS: Drywell Pressure GREATER THAN 2 psig due to RCS Leakage (not caused by a loss of DW Cooling).

POTENTIAL LOSS: None DAEC INFORMATION:

There is no significant deviation from the generic indicator. The value for this loss indicator corresponds to the drywell high pressure ECCS initiation signal setpoint of 2.0 psig. The indicator also ensures this threshold is not misapplied to conditions that do not indicate RCS leakage into the drywell, i.e., the drywell pressure increase is not due to loss of drywell cooling.

DAEC uses a GE Mark I Containment. During reactor operation, with drywell cooling in operation and the drywell inerted, the normal operating pressure in the drywell is between 0.5 and 1.0 psig. Analysis at the DAEC shows that a 50 gpm RCS leak would result in a 2 to 3 psig pressure rise over a six minute time period. Since a 2 psig rise would place DAEC above the ECCS initiation setpoint, (2 psig) it is necessary to select the DAEC ECCS initiation setpoint of 2 psig to indicate an actual loss of the RCS. Drywell cooling is not isolated at the 2 psig ECCS initiation setpoint, therefore further pressure rise would be indicative of a RCS leak.

REFERENCES:

1. Emergency Operating Procedures (EOP) Bases, Breakpoints

2. Emergency Operating Procedures (EOP) -1, RPV Control

3. Emergency Operating Procedures (EOP) -2, Primary Containment Control

4. NEI 99-01 Rev. 5, NEI Methodology for Development of Emergency Action Levels RCS Barrier Pri. Cont. Atmosphere

EBD F Rev. 8 next FISSION PRODUCT BARRIER DEGRADATION PAGE 19 of 31 FISSION BARRIER: RCS DAEC INDICATOR: Emergency Director Judgment EAL THRESHOLD VALUE:

Any condition in the opinion of the Emergency Director (OSM, EC or ER&RD) that indicates LOSS or POTENTIAL LOSS of the RCS Barrier.

DAEC INFORMATION:

There is no significant deviation from the generic EAL. Emergency Director considerations for determining whether any barrier "Loss" or "Potential Loss" include imminent barrier degradation, degraded barriermonitoringcapability, and consideration of dominant accident sequences.

Any condition which in the judgment of the Emergency Director indicates a LOSS or POTENTIAL LOSS of the RCS barrier such as, but not limited to:

• Degraded barriermonitoringcapability from loss of/lack of reliable indicators.

• Consideration for instrumentation operability.

• Portable instrumentation readings.

• Offsite monitoring results.

• Complete loss of 125 VDC.

• Loss of decay heat removal.

• ATWS with failure of Standby Liquid Control.

• Prolonged station blackout.

• Loss of offsite power with early HPCI/RCIC failure Imminent means that no turnaround in safety system performance is expected and that General Emergency conditions can be expected to occur within two hours. Imminent fission barrier degradation must be considered by the Emergency Director to assure timely declaration of a General Emergency and to better assure that offsite protective actions can be effectively accomplished.

Degraded barriermonitoringcapability from loss of/lack of reliable indicators must also be considered by the Emergency Director when determining ifa fission barrier loss or potential loss has occurred.

RCS Barrier ED Judgment

This assessment should also include consideration for instrumentation operability and portable instrumentation readings.

Offsite monitoring results may be an indication of Fission Product Barrier degradation causing an unmonitored release.

Dominantaccident sequences can lead to loss of all Fission Barriers. Based on the IPE, the dominant accident sequences leading to core damage at DAEC include complete loss of 125 VDC, loss of decay heat removal, ATWS with failure of Standby Liquid Control, prolonged station blackout, and loss of offsite power with early HPCI/RCIC failure. The Emergency Director should also consult System Malfunction EALs, as appropriate, to assure timely emergency classification Forthe RCS barrier,the Emergency Director should also considersafety-relief valves (SRVs) open or cycling. If an SRV is stuck open or is cycling and no other emergency conditions exist, an emergency declaration may not be appropriate. However, if the fuel is damaged and the SRV is allowing fission products to escape into primary containment, a loss of RCS should be determined as having occurred.

REFERENCES:

1. Emergency Plan Implementing Procedure (EPIP) 2.5, Control Room Emergency Response Operation

2. Duane Arnold Energy Center Individual Plant Examination (IPE) November 1992

3. NEI 99-01 Rev. 5, NEI Methodology for Development of Emergency Action Levels RCS Barrier ED Judgment

EAL BASES DOCUMENT EBD F Rev. 8 next FISSION PRODUCT BARRIER DEGRADATION PAGE 21 of 31 FISSION BARRIER: Primary Containment DAEC INDICATOR: Radiation/Core Damage EAL THRESHOLD VALUE:

Significant Radioactive Inventory in Containment LOSS: None POTENTIAL LOSS: Drywell Area Hi Range Rad Monitor RIM-9184A or B reading GREATER THAN 3000 Rem/hr OR POTENTIAL LOSS: Torus Area Hi Range Rad Monitor RIM-9185A or B reading GREATER THAN 100 Rem/hr DAEC INFORMATION:

There is no significant deviation from the generic indicators. The potential loss (site-specific) indicator value corresponds to at least 20% fuel clad damage with release into the primary containment. This indicator corresponds to loss of both the Fuel Clad and RCS barriers with Potential Loss of the Primary Containment barrier, and would result in declaration of a General Emergency. The basis for the 20% fuel clad damage threshold is described under the 20% core damage assessment indicator. It is intended that determination of barrierpotentialloss be made whenever the indicatorthreshold is reached until such time that core damage assessment is performed, at which time direct use of containment rad monitorreadings is no longerrequired.

Primary Containment Barrier Radiation/Core Damage

EBD F Rev. 8 next FISSION PRODUCT BARRIER DEGRADATION PAGE 22 of 31 As documented by NG-88-0966, General Electric performed a study to predict dose rate readings from fuel damage calculations for emergency planning. The calculations were performed to obtain gamma ray dose rates at the locations of the containment atmospheric monitoring system radiation detectors in the drywell and torus locations for assumed releases of gap activity from the core. These calculations were based on "nominal" estimates of fuel rod gap fission product inventory fractions, which are considered to be more appropriate for determining a minimum threshold reading than inventory assumptions found in the NRC Regulatory Guides. The Regulatory Guide inventory assumptions applicable to dose assessments are larger and therefore non-conservative for determination of this EAL threshold. Two separate cases were evaluated.

In the first case, the released activity was assumed to be contained in the drywell atmosphere. This case is considered representative of conditions following a line break in which activity is released directly into the drywell. In the second case, the released activity was assumed to be contained in the torus. This could be applied for an event which results in vessel isolation and blowdown to the suppression chamber. The results for each case were provided for each case in the form of gamma ray dose rate versus time profiles for assumed releases of 100% and 20% of the gap activity from the core. The dose rate calculations were carried out independent of any specific information on details of construction or response characteristics of the detector systems. The figures show a drywell reading of about 2.9 x 103 Rem/hr and a torus reading of about 1.1 x 102 Rem/hr associated with 20% gap release at two hours after shutdown. These values are rounded to 3 E+3 Rem/hr and 1 E+2 Rem/hr, respectively. The two hour point was picked because it allows ample time for the Technical Support Center to be operational and core damage assessment to begin.

REFERENCES:

1. Office Memo NG-88-0966, G.E. Fuel Damage Documentation/Dose Rate Calculations, 03/18/88

2. NEI 99-01 Rev. 5, NEI Methodology for Development of Emergency Action Levels Primary Containment Barrier Radiation/Core Damage

EAL BASES DOCUMENT EBD F Rev. 8 next FISSION PRODUCT BARRIER DEGRADATION PAGE 23 of 31 FISSION BARRIER: Primary Containment DAEC INDICATOR: Radiation/Core Damage EAL THRESHOLD VALUE:

Clad Damage Determination LOSS: None POTENTIAL LOSS: Fuel Damage assessment (PASAP 7.2) indicates at least 20%

fuel clad damage.

DAEC INFORMATION:

As a site-specific "potential loss" indicator, DAEC uses determination of at least 20% fuel clad damage, which is consistent with the level of fuel damage indicated by the drywell and torus radiation monitor readings used earlier with this Indicator. This can be determined using appropriate fuel damage assessment procedures. Regardless of whether primary containment integrityis challenged,it is possible for significant radioactivitywithin the primary containment to result in EPA PAG plume exposure levels being exceeded even assuming that the primary containment is within technical specification allowable leakage rates. With or without primary containment challenge, however, a major release of radioactivity requiring off-site protective actions from core damage is not possible unless a major failure of the fuel clad barrier allows radioactive material to be released from core into the reactor coolant. NUREG-1228 indicates that such conditions do not exist when the amount of fuel clad damage is less than 20%.

Other indicators were also considered. No other reliable indicators for Primary Containment "loss" or "potential loss" could be determined.

REFERENCES:

1. Post Accident Sampling and Analysis Procedure (PASAP) 7.2, Fuel Damage Assessment

2. NUREG-1228, Source Term Estimations DuringIncident Response to Severe Nuclear PowerPlantAccidents, October 1988

3. NEI 99-01 Rev. 5, NEI Methodology for Development of Emergency Action Levels Primary Containment Barrier Radiation/Core Damage

EAL BASES DOCUMENT  : EBD F Rev. 8 next FISSION PRODUCT BARRIER DEGRADATION PAGE 24 of 31 FISSION BARRIER: Primary Containment DAEC INDICATOR: RPV Level EAL THRESHOLD VALUE:

Reactor Vessel Water Level LOSS: None POTENTIAL LOSS: Primary Containment flooding required.

DAEC INFORMATION:

The entry into the Primary Containment Flooding emergency procedure indicates reactor vessel water level can not be restored and that a core melt sequence is possible. SAGs direct the operators to enter Containment Flooding when Reactor Vessel Level cannot be restored to greater than a Site Specific value (generally 2/3 core height) or is unknown. The condition in this potential loss EAL represent a potential core melt sequence which, if not corrected, could lead to vessel failure and increased potential for containment failure. In conjunction with Reactor Vessel water level "Loss" thresholds in the Fuel Clad and RCS barrier columns, this threshold will result in the declaration of a General Emergency -- loss of two barriers and the potential loss of a third. If the emergency operating procedures have been ineffective in restoring reactor vessel level above the RCS and Fuel Clad Barrier Threshold Values, there is not a "success" path and a core melt sequence is in progress. Entry into Containment flooding procedures is a logical escalation in response to the inability to maintain reactor vessel level. Severe accident analysis (e.g., NUREG-1150) have concluded that function restoration procedures can arrest core degradation with the reactor vessel in a significant fraction of the core damage scenarios, and the likelihood of containment failure is very small in these events. Given this, it is appropriate to provide a reasonable period to allow emergency operating procedures to arrest the core melt sequence.

Whether or not the procedures will be effective should be apparent within the time provided. The Emergency Director should make the declaration as soon as it is determined that the procedures have been, or will be, ineffective. There is no "loss" EAL associated with this item.

REFERENCES:

1. Emergency Operating Procedure (EOP) RPV/F - RPV Flooding

2. NEI 99-01 Rev. 5, NEI Methodology for Development of Emergency Action Levels Primary Containment Barrier Leakage

EAL BASES DOCUMENT EBD F Rev. 8 next FISSION PRODUCT BARRIER DEGRADATION PAGE 25 of 31 FISSION BARRIER: Primary Containment DAEC INDICATOR: Leakage EAL THRESHOLD VALUE:

Containment Isolation Valve Status After Containment Isolation Signal LOSS: Failure of both isolation valves in any one line to close AND a direct downstream pathway to the environment exists after primary containment isolation signal.

OR LOSS: UNISOLABLE primary system leakage outside the drywell as indicated by area temps or ARMs exceeding the Max Safe Limits per EOP 3, Table 6, when Containment Isolation is required.

OR LOSS: Intentional Primary Containment venting per EOPs.

POTENTIAL LOSS: None DAEC INFORMATION:

This EAL is intended to cover the inability to isolate primary containment when primary containment isolation is required either by an automatic Primary Containment Isolation System (PCIS) signal or manual operator action to isolate the pathway. In addition, the presence of area radiation or temperature alarms above the Max Safe limits listed in Table 6, EOP 3 after a containment isolation, indicate an UNISOLABLE primary system leakage outside the drywell. The indicators should be confirmed to be caused by RCS leakage.

The first part of the "Failure of both isolation valves..." loss threshold describes a condition where one isolation valve is not fully closed when required AND the other isolation valve is also not fully closed when required. The second part of this threshold is met when the two isolation valve failures described previously create a pathway to the environment (no barriers exist to prevent an actual or potential release).

Primary Containment Barrier Leakage

EAL BASES DOCUMENT EBD"F Rev. 8 next FISSION PRODUCT BARRIER DEGRADATION PAGE 26 of 31 The use of the modifier "direct" in defining the release path discriminates against release paths through interfacing liquid systems. In-line charcoal filter systems do not make a release path indirect since the filter is not effective at removing fission product noble gases. Since the fission product release would be driven by boiling in the reactor vessel, the high humidity in the release steam can be expected to render the filters ineffective in a short period.

A downstream pathway to the environment exists whenever the contents of primary containment can reach the environment through the failed barrier, whether or not those contents are treated or released via an elevated release point. This means any leak path into secondary containment will reach the environment.

Venting of the primary containment can be performed in accordance with EOP 2 and SAGs irrespective of the offsite radioactivity release rate that will occur and by defeating isolation interlocks as necessary. The consequences of not doing so may be the loss of primary containment integrity, core damage, and an uncontrolled radioactive release much greater than might otherwise occur. Primary containment venting is performed only as necessary to reduce and then maintain torus pressure below the Primary Containment Pressure Limit (PCPL) of 53 psig.

Intentional venting of primary containment for primary containment pressure or combustible gas control per EOPs or SAGs to the secondary containment and/or the environment is considered a loss of containment. Containment venting for pressure when not in an accident situation should not be considered.

If UNISOLABLE primary system leakage is occurring outside the drywell when primary containment isolation is required, the Primary Containment fission product barrier is not considered to be lost until area temps or ARMs exceed Max Safe Limits per EOP 3, Table

6. This generally will result in escalation from an ALERT (the RCS barrier is considered to be potentially lost when UNISOLABLE primary system leakage outside the drywell is indicated by area temperatures or ARMs exceeding Max Normal Limits per EOP-3, Table

6) to a Site Area Emergency (Primary Containment is considered to be lost when Max Safe Limits per EOP-3, Table 6 are exceeded) due to the loss or potential loss of 2 out of 3 fission product barriers.

Primary Containment Barrier Leakage

EAL BASES DOCUMENT EBD F Rev. 8 next FISSION PRODUCT BARRIER DEGRADATION PAGE 27 of 31

REFERENCES:

1. Emergency Operating Procedure (EOP) 2, Primary Containment Control

2. Emergency Operating Procedure (EOP) 3, Secondary Containment Control

3. Emergency Operating Procedures (EOP) Bases, Breakpoints

4. NEI 99-01 Rev. 5, NEI Methodology for Development of Emergency Action Levels Primary Containment Barrier Leakage

EAL BASES DOCUMENT EBD F Rev. 8 next FISSION PRODUCT BARRIER DEGRADATION PAGE 28 of 31 FISSION BARRIER: Primary Containment DAEC INDICATOR: Primary Containment Atmosphere EAL THRESHOLD VALUE:

Drywell Pressure/Atmosphere LOSS: Drywell pressure rise followed by a rapid unexplained drop in drywell pressure.

OR LOSS: Drywell pressure response NOT consistent with LOCA conditions.

POTENTIAL LOSS: Torus Pressure reaches 53 PSIG and rising.

OR POTENTIAL LOSS: Drywell or Torus H 2 cannot be determined to be LESS THAN 6% and Drywell or torus 02 cannot be determined to be LESS THAN 5%.

OR POTENTIAL LOSS: RPV pressure and Torus water temperature cannot be maintained below the Heat Capacity Limit (EOP Graph 4).

DAEC INFORMATION:

There are no significant deviations from the generic indicators. The "loss" indicators used at DAEC directly correspond to the generic indicators.

Rapid unexplained loss of pressure (i.e., not attributable to drywell spray or condensation effects) following an initial pressure increase from a high energy line break indicates a loss of containment integrity. Primary containment pressure should increase as a result of mass and energy release into containment from a LOCA. Thus, primary containment pressure not increasing under these conditions indicates a loss of containment integrity.

The LOSS thresholds describe a condition where there is an absence of a pressure rise that would be expected for LOCA conditions. The implication is that the containment atmosphere is vented through an unknown leakage path resulting in little to no pressure increase.

Primary Containment Barrier Primary Containment Atmosphere

EAL BASES DOCUMENT EBD F Rev. 8 next FISSION PRODUCT BARRIER DEGRADATION PAGE 29 of 31 Other pressure anomalies may indicate malfunctions internal to containment that do not indicate a LOSS of the containment barrier. For example, the BWR Mark I containment relies on the pressure suppression function to limit containment pressure rise. If this suppression capability is lost or bypassed, it may result in containment pressures exceeding the maximum design basis pressure. Exceeding the maximum design basis containment pressure would result in a POTENTIAL LOSS of containment.

The first "potential loss" indicator is torus pressure of 53 psig, which is the Primary Containment Pressure Limit (PCPL) used in the EOPs.

The second "potential loss" indicator is based on determination of explosive mixture in accordance with the SAGs. DAEC SAGs require control of drywell and torus atmosphere gas concentrations to less than 6% H 2 and less than 5% 02 to assure that an explosive mixture does not exist. This "potential loss" indicator is written to be consistent with the SAGs.

The third "potential loss" indicator uses the Heat Capacity Limit (HCL) which is defined to be the highest torus temperature at which initiation of RPV depressurization will not result in exceeding the Primary Containment Pressure Limit (the PCPL is 53 psig at the DAEC) before the rate of energy transfer from the RPV to the primary containment is within the capacity of the containment vent.

REFERENCES:

1. Emergency Operating Procedure (EOP) 2, Primary Containment Control

2. Severe Accident Guideline - 3 (SAG-3), Hydrogen Control

3. NEI 99-01 Rev. 5, NEI Methodology for Development of Emergency Action Levels Primary Containment Barrier Primary Containment Atmosphere

EAL BASES DOCUMENT EBD F Rev. 8 next FISSION PRODUCT BARRIER DEGRADATION PAGE 30 of 31 FISSION BARRIER: Primary Containment DAEC INDICATOR: Emergency Director Judgment EAL THRESHOLD VALUE:

Any condition in the opinion of the Emergency Director (OSM, EC or ER&RD) that indicates LOSS or POTENTIAL LOSS of the Containment Barrier.

DAEC INFORMATION:

There is no significant deviation from the generic indicator. Emergency Director considerations for determining whether any barrier "Loss" or "Potential Loss" include imminent barrier degradation, degraded barriermonitoringcapability, and consideration of dominant accidentsequences.

Any condition which in the judgment of the Emergency Director that indicates LOSS or POTENTIAL LOSS of the Primary Containment Barrier such as, but not limited to:

• Degraded barriermonitoring capability from loss of/lack of reliable indicators.

• Consideration for instrumentation operability.

• Portable instrumentation readings.

• Offsite monitoring results.

• Complete loss of 125 VDC.

• Loss of decay heat removal.

• ATWS with failure of Standby Liquid Control.

• Prolonged station blackout.

• Loss of offsite power with early HPCI/RCIC failure Imminent means that no turnaround in safety system performance is expected and that General Emergency conditions can be expected to occur within two hours. Imminent fission barrier degradation must be considered by the Emergency Director to assure timely declaration of a General Emergency and to better assure that offsite protective actions can be effectively accomplished.

Degraded barriermonitoring capability from loss of/lack of reliable indicators must also be considered by the Emergency Director when determining if a fission barrier loss or potential loss has occurred.

Primary Containment Barrier EC/OSM Judgement

EAL BASES DOCUMENT EBD F Rev. 8 next FISSION PRODUCT BARRIER DEGRADATION PAGE 31 of 31 This assessment should also include consideration for instrumentation operability and portable instrumentation readings.

Offsite monitoring results may be an indication of Fission Product Barrier degradation causing an unmonitored release.

Dominant accident sequences can lead to loss of all Fission Barriers. Based on the IPE, the dominant accident sequences leading to core damage at DAEC include complete loss of 125 VDC, loss of decay heat removal, ATWS with failure of Standby Liquid Control, prolonged station blackout, and loss of offsite power with early HPCI/RCIC failure. The Emergency Director should also consult System Malfunction EALs, as appropriate, to assure timely emergency classification

REFERENCES:

1. Emergency Plan Implementing Procedure (EPIP) 2.5, Control Room Emergency Response Operation

2. Duane Arnold Energy Center Individual Plant Examination (IPE) November 1992

3. NEI 99-01 Rev. 5, NEI Methodology for Development of Emergency Action Levels Primary Containment Barrier EC/OSM Judgement

EAL BASES DOCUMENT EBD-H Rev. 11 next HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY PAGE 1 of 30 I ~ Usage Level INFORMATION USE Effective Date:

Approved for 'Point-of-Use' printing IF NO Temporary Changes are in effect for this procedure.

(on designated printers)

Record the following: Date / Time: / Initials:

NOTE: A check to ensure currentrevision and no temporary changes shall be performed and documented every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> if active document use exceeds a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period as determined from the date and time recordedabove.

Prepared By: Michael Winchester / Date:

Print Signature CROSS-DISCIPLINE REV IEW (AS REQUIRED) -

Reviewed By: / Date:

Print Signature Reviewed By: / Date:

Print Signature Reviewed By: / Date:

Print Signature PROCEDURE APPROVAL BY QUALI FIEDREVIIEWER Approved By / Date:

Print Signature

EAL BASES DOCUMENT EBD-H Rev. 11 next HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY PAGE 2 of 30 HUI Natural or destructive phenomena affecting the PROTECTED AREA EVENT TYPE: Natural Disasters and Destructive Phenomena OPERATING MODE APPLICABILITY: All EAL THRESHOLD VALUE:

HU1.1 Seismic event identified by ANY 2 of the following:

" Seismic event confirmed per AOP 901, Earthquake

" Report of an earthquake felt on-site

• National Earthquake Information Center (1-303-273-8500)

HU1.2 Tornado striking within the PROTECTED AREA, or within the switchyard, with NO confirmed damage to a Safe Shutdown/Vital Area or Control Room indication of degraded performance of a System of Concern.

HU1.3 High winds greater than 95 mph on-site with NO confirmed damage to a Safe Shutdown/Vital Area or Control Room indication of degraded performance of a System of Concern.

HU1.4 Internal flooding that has the potential to affect safety related equipment required by Technical Specifications for the current operating mode in ANY Safe Shutdown/Vital Area.

HU1.5 Turbine failure resulting in casing penetration or damage to turbine or generator seals.

HU1.6 River level above 757 feet.

DAEC EAL INFORMATION:

The Protected Area is the area within the security fence. This includes ISFSI and the Intake Structure.

EAL #1 This EAL addresses damage that may be caused to some portions of the site, but should not affect ability of safety functions to operate. In accordance with AOP 901 (Earthquake), at DAEC, a minimum detectable earthquake that is indicated on panel 1C35 is an acceleration greater than +/- 0.01 Gravity.

As defined in the EPRI-sponsored Guidelines for Nuclear Plant Response to an Earthquake, dated October 1989, a "felt earthquake" is: An earthquake of sufficient intensity such that: (a) the vibratory ground motion is felt at the nuclear plant site and recognized as an earthquake HU1

EAL BASES DOCUMENT EBD-H Rev. 11 next HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY PAGE 3 of 30 based on a consensus of control room operators on duty at the time, and (b) for plants with operable seismic instrumentation, the seismic switches of the plant are activated.

The National Earthquake Information Center (1-303-273-8500) can confirm if an earthquake has occurred in the area of the plant.

EAL #2 This EAL addresses report of a tornado striking (touching down) within the Protected Area.

Escalation of this emergency classification level, if appropriate, would be based on VISIBLE DAMAGE, or by other in plant conditions, via HAI.

EAL #3 This EAL is based on the assumption that high winds within the PROTECTED AREA may have potentially damaged plant structures containing functions or systems required for safe shutdown of the plant. The EAL addresses high wind speeds as measured by the 33-foot or 156-foot elevations on the Meteorological Tower. The design basis wind speed is 105 miles per hour.

However, the meteorological instrumentation is only capable of measuring wind speeds up to 100 miles per hour. Thus the value of 95 miles per hour is selected to be on scale for the meteorological instrumentation and to conservatively account for potential measurement errors.

EAL #4 This EAL addresses the effect of internal flooding caused by events such as component failures, equipment misalignment, or outage activity mishaps.

The site specific areas include those areas that contain systems required for safe shutdown of the plant, which are not designed to be partially or fully submerged. Escalation of this emergency classification level, if appropriate, would be based on VISIBLE DAMAGE via HA1, or by other plant conditions.

HU1

EAL BASES DOCUMENT EBD-H Rev. 11 next HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY PAGE 4 of 30 Safe Shutdown/Vital Areasis,,,

Category Area 1G31 DG and Day Tank Rooms, 1G21 DG and Day Tank Rooms, Battery Rooms, Essential Switchgear Rooms, Cable Spreading Room Heat Colant Sink /

Suy Torus Room, Intake Structure, Pumphouse Coolant Supply Containment Drywell, Torus Emergency NE, NW, SE Corner Rooms, HPCI Room, RCIC Room, RHR Valve Room, Systems North CRD Area, South CRD Area, CSTs Other Control Building, Remote Shutdown Panel 1C388 Area, Panel 1C55/56 Area, SBGT Room

. ystems of Concern S

• Reactivity Control

" Containment (Drywell/Torus)

• RHR/Core Spray/SRVs

" HPCI/RCIC

" RHRSW/River Water/ESW

" Onsite AC Power/EDGs

" Offsite AC Power

" Instrument AC

" DC Power

" Remote Shutdown Capability EAL #5 This EAL addresses main turbine rotating component failures of sufficient magnitude to cause observable damage to the turbine casing or to the seals of the turbine generator. Generator seal damage observed after generator purge does not meet the intent of this EAL because it did not impact normal operation of the plant.

Of major concern is the potential for leakage of combustible fluids (lubricating oils) and gases (hydrogen cooling) to the plant environs. Actual FIRES and flammable gas build up are appropriately classified via HU2 and HU3.

This EAL is consistent with the definition of a NOUE while maintaining the anticipatory nature desired and recognizing the risk to non-safety related equipment.

HU1

EAL BASES DOCUMENT EBD-H Rev. 11 next HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY PAGE 5 of 30 Escalation of this emergency classification level, if appropriate, would be to HA1 based on damage done by PROJECTILES generated by the failure or by a radiological release. This latter event would be classified by the radiological ICs or Fission Product Barrier ICs.

EAL #6 This EAL addresses the observed effects of flooding in accordance with AOP 902 (Flood). Plant site finished grade is at elevation 757.0 ft. Personnel doors and railroad and truck openings at or near grade would require protection in the event of a flood above elevation 757.0 ft.

Therefore, EAL 6 uses a threshold of flood water levels above 757.0 ft.

REFERENCES:

1. Abnormal Operating Procedure (AOP) 901, Earthquake

2. Abnormal Operating Procedure (AOP) 902, Flood

3. Abnormal Operating Procedure (AOP) 903, Tornado

4. Emergency Operating Procedure (EOP)-3, Secondary Containment Control

5. EOP Basis Document, EOP-3, Secondary Containment Control

6. UFSAR Chapter 3, Design of Structures, Components, Equipment, and Systems

7. Bechtel Drawing BECH-M017, Equipment Location - Intake Structure Plans at Elevations, Rev. 6

8. NEI 99-01 Rev. 5, NEI Methodology for Development of Emergency Action Levels HU1

EAL BASES DOCUMENT EBD-H Rev. 11 next PLANT SAFETY HAZARDS & OTHER CONDITIONS AFFECTING PAGE 6 of 30 HU2 FIRE within the PROTECTED AREA NOT extinguished within 15 minutes of detection or EXPLOSION within the PROTECTED AREA EVENT TYPE: Fire OPERATING MODE APPLICABILITY: All EAL THRESHOLD VALUE:

Note: 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 duration has exceeded, or will likely exceed, the applicable time.

HU2.1. FIRE NOT extinguished within 15 minutes of Control Room notification or verification of a Control Room FIRE alarm in ANY Safe ShutdownNital Area.

HU2.2 EXPLOSION within the PROTECTED AREA.

DAEC EAL INFORMATION:

This EAL addresses the magnitude and extent of FIRES or EXPLOSIONS that may be potentially significant precursors of damage to safety systems. It addresses the FIRE /

EXPLOSION, and not the degradation in performance of affected systems that may result.

As used here, detection is visual observation and report by plant personnel or sensor alarm indication.

EAL #1 The 15 minute time period begins with a credible notification that a FIRE is occurring, or indication of a fire detection system alarm/actuation. Verification of a fire detection system alarm/actuation includes actions that can be taken within the control room or other nearby site specific location to ensure that it is not spurious. An alarm is assumed to be an indication of a FIRE unless it is disproved within the 15 minute period by personnel dispatched to the scene. In other words, a personnel report from the scene may be used to disprove a sensor alarm if received within 15 minutes of the alarm, but shall not be required to verify the alarm.

The intent of this EAL is to exclude buildings (i.e., Warehouse, Construction Support Center, Maintenance Fab Shop, etc.) or areas that are not VITAL AREAS. This excludes FIRES such as waste-basket FIRES, and other small FIRES of no safety consequence HU2

EAL BASES DOCUMENT, EBD-H Rev. 11 next HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY PAGE 7 of 30 The intent of this 15 minute duration is to size the FIRE and to discriminate against small FIRES that are readily extinguished (e.g., smoldering waste paper basket).

Per AOP 913, the location of a FIRE can be determined by observing 1C40B alarm messages, Zone Indicating Unit (ZIU) alarms, or FIRE annunciators on panels 1C40 and 1C40A. The location of a FIRE can also be determined by verbal report of the person discovering the FIRE.

...... _____Safe .. Shutdow..i. eas".. ..

Category Area Electrical Power 1G31 DG and Day Tank Rooms, 1G21 DG and Day Tank Rooms, Battery Rooms, Essential Switchgear Rooms, Cable Spreading Room Heat Colant Sink!/

Suy Torus Room, Intake Structure, Pumphouse Coolant Supply Containment Drywell, Torus Emergency NE, NW, SE Corner Rooms, HPCI Room, RCIC Room, RHR Valve Room, Systems North CRD Area, South CRD Area, CSTs Control Building, Remote Shutdown Panel 1C388 Area, Panel 1C55/56 Area, SBGT Room EAL #2 This EAL addresses only those EXPLOSIONS of sufficient force to damage permanent structures or equipment within the PROTECTED AREA.

No attempt is made to assess the actual magnitude of the damage. The occurrence of the EXPLOSION is sufficient for declaration.

The Emergency Director also needs to consider any security aspects of the EXPLOSION, if applicable.

Escalation of this emergency classification level, if appropriate, would be based on HA2.

REFERENCES:

1. Abnormal Operating Procedure (AOP) 913, Fire

2. Abnormal Operating Procedure (AOP) 914, Security

3. NEI 99-01 Rev. 5, NEI Methodology for Development of Emergency Action Levels HU2

EAL BASES DOCUMENT .. .... B.D-H Rev. 11 next PLANT SAFETY HAZARDS & OTHER CONDITIONS AFFECTING PAGE 8 of 30 HU3 Release of toxic, corrosive, asphyxiant, or flammable gases deemed detrimental to NORMAL PLANT OPERATIONS EVENT TYPE: Other Hazards and Failures OPERATING MODE APPLICABILITY: All EAL THRESHOLD VALUE:

HU3.1 Toxic, corrosive, asphyxiant or flammable gases in amounts that have or could adversely affect NORMAL PLANT OPERATIONS.

HU3.2 Report by Local, County or State Officials for evacuation or sheltering of site personnel based on an off-site event.

DAEC EAL INFORMATION:

This EAL is based on the release of toxic, corrosive, asphyxiant or flammable gases of sufficient quantity to affect Normal Plant Operations.

The fact that SCBA may be worn does not eliminate the need to declare the event.

This EAL is not intended to require significant assessment or quantification. It assumes an uncontrolled process that has the potential to affect plant operations. This would preclude small or incidental releases, or releases that do not impact structures needed for plant operation.

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.

Escalation of this emergency classification level, if appropriate, would be based on HA3.

REFERENCES:

1. UFSAR Section 2.2, Nearby Industrial, Transportation, and Military Facilities

2. UFSAR Section 6.4, Habitability Systems

3. NEI 99-01 Rev. 5, NEI Methodology for Development of Emergency Action Levels HU3

EAL BASES DOCUMENT. ..... EBD-H Rev. 11 next HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY PAGE 9 of 30 HU4 Confirmed SECURITY CONDITION or threat which indicates a potential degradation in the level of safety of the plant EVENT TYPE: Security OPERATING MODE APPLICABILITY: All EAL THRESHOLD VALUE:

HU4.1 A SECURITY CONDITION that does NOT involve a HOSTILE ACTION as reported by DAEC Security Shift Supervision.

HU4.2 A credible site specific security threat notification.

HU4.3 A validated notification from NRC providing information of an aircraft threat DAEC EAL INFORMATION:

Note: Timely and accurate communication between Security Shift Supervision and the Control Room is crucial for the determination and implementation of effective Security EALs.

A SECURITY CONDITION is defined as: Any Security Event as listed in the approved security contingency plan that constitutes a threat/compromiseto site security, threat/riskto site personnel, or a potential degradationto the level of safety of the plant. A SECURITY CONDITION does not involve a HOSTILE ACTION.

Security events which do not represent at least a potential degradation in the level of safety of the plant are reported under 10 CFR 73.71 or in some cases under 10 CFR 50.72. Security events assessed as HOSTILE ACTIONS are classifiable under HA4, HS4 and HG1.

A higher initial classification could be made based upon the nature and timing of the security threat and potential consequences. The licensee shall consider upgrading the emergency response status and emergency classification level in accordance with the site's Safeguards Contingency Plan and Emergency Plan.

EAL #1 Reference is made to site specific security shift supervision because these individuals are the designated personnel on-site qualified and trained to confirm that a security event is occurring or has occurred. Training on security event classification confirmation is closely controlled due to the strict secrecy controls placed on the plant Safeguards Contingency Plan.

This threshold is based on site specific security plans. Site specific Safeguards Contingency Plans are based on guidance provided by NEI 03-12.

HU4

EAL BASES DOCUMENT- EBD-H Rev. 11 next HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY PAGE 10 of 30 EAL #2 This threshold is included to ensure that appropriate notifications for the security threat are made in a timely manner. This includes information of a credible threat. The emergency response to a Credible Security Threat is initiated through AOP 914, "Security Events".

The determination of "credible" is made through use of information found in the site specific Safeguards Contingency Plan.

EAL #3 The intent of this EAL is to ensure that notifications for the aircraft threat are made in a timely manner and that Offsite Response Organizations and plant personnel are at a state of heightened awareness regarding the credible threat. It is not the intent of this EAL to replace existing non-hostile related EALs involving aircraft.

This EAL is met when a plant receives information regarding an aircraft threat from NRC.

Validation is performed by calling the NRC or by other approved methods of authentication.

The NRC Headquarters Operations Officer (HOO) will communicate to the licensee if the threat involves an airliner (airliner is meant to be a large aircraft with the potential for causing significant damage to the plant). The status and size of the plane may be provided by NORAD through the NRC.

Escalation to Alert emergency classification level via HA4 would be appropriate if the threat involves an airliner within 30 minutes of the plant.

REFERENCES:

1. Abnormal Operating Procedure (AOP) 914, Security Events

2. NEI 99-01 Rev. 5, NEI Methodology for Development of Emergency Action Levels

3. NRC Bulletin 2005-02: Emergency Preparedness and Response Actions For Security-Based Events HU4

EAL BASES DOCUMENT EBD-H Rev. 11 next PLANT SAFETY HAZARDS & OTHER CONDITIONS AFFECTING PAGE 11 of 30 HU5 Other conditions exist which in the judgment of the Emergency Director (OSM, EC or ER&RD) warrant declaration of a NOUE EVENT TYPE: Emergency Director Judgment OPERATING MODE APPLICABILITY: All EAL THRESHOLD VALUE:

HU5.1 Other conditions exist which in the judgment of the Emergency Director (OSM, EC or ER&RD) 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 off-site response or monitoring are expected unless further degradation of safety systems occurs.

DAEC EAL INFORMATION:

This EAL 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 NOUE emergency classification level.

REFERENCES:

1. NEI 99-01 Rev. 5, NEI Methodology for Development of Emergency Action Levels HU5

EAL BASES DOCUMENT EBD-H Rev. 11 next HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY PAGE 12 of 30 HA1 Natural and destructive phenomena affecting VITAL AREAS EVENT TYPE: Natural Disasters and Destructive Phenomena OPERATING MODE APPLICABILITY: All EAL THRESHOLD VALUE:

HA1.1 Receipt of the Amber Operating Basis Earthquake Light and the wailing seismic alarm on 1C35 (+/- 0.06 gravity).

AND Earthquake confirmed by ANY of the following:

" Report of an earthquake felt on-site

" National Earthquake Information Center (1-303-273-8500)

• Control Room indication of degraded performance of systems required for the safe shutdown of the plant.

HA1.2 Tornado strike, high winds greater than 95MPH or a vehicle crash resulting in:

• VISIBLE DAMAGE to ANY of the following structures:

o Emergency Diesel Generator Rooms o Control Building o Reactor Building o Pumphouse o Intake Structure o Condensate Storage Tank Area OR 0 Control Room indication of degraded performance of a System of Concern.

HA1.3 Turbine failure-generated PROJECTILES resulting in:

• VISIBLE DAMAGE to or penetration of any of the following structures:

o Emergency Diesel Generator Rooms o Control Building o Reactor Building o Condensate Storage Tank Area HA1

EAL BASES DOCUMENT EBD-H Rev. 11 next HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY PAGE 13 of 30 OR 0 Control Room indication of degraded performance of a System of Concern HA1.4 Internal flooding in ANY Safe Shutdown/Vital Area that results in:

• an electrical shock hazard that precludes access to operate or monitor safety equipment.

OR

• Control Room indication of degraded performance of a System of Concern.

HA1.5 Report to Control Room of VISIBLE DAMAGE affecting a Safe Shutdown/Vital Area.

HA1.6 River level above 767 feet.

HA1.7 River Water Supply Pit low level.

DAEC EAL INFORMATION:

These EALs escalate from HU1 in that the occurrence of the event has resulted in VISIBLE DAMAGE to plant structures or areas containing equipment necessary for a safe shutdown, or has caused damage to the safety systems in those structures evidenced by control room indications of degraded system response or performance. The occurrence of VISIBLE DAMAGE and/or degraded system response is intended to discriminate against lesser events. The initial report should not be interpreted as mandating a lengthy damage assessment prior to classification. No attempt is made in this EAL to assess the actual magnitude of the damage.

The significance here is not that a particular system or structure was damaged, but rather, that the event was of sufficient magnitude to cause this degradation.

In this EAL, "Vital Area" is defined as plant structures or areas containing equipment necessary for a safe shutdown, i.e., synonymous with Safe Shutdown Area.

EAL #1 This EAL addresses seismic events of a magnitude that can result in a VITAL AREA being subjected to forces beyond design limits, and thus damage may be assumed to have occurred to plant safety systems.

OBE events that are detected in accordance with AOP 901. For DAEC, the OBE is associated with a peak horizontal acceleration of +/- 0.06 Gravity.

As defined in the EPRI-sponsored Guidelines for Nuclear Plant Response to an Earthquake, dated October 1989, a "felt earthquake" is: An earthquake of sufficient intensity such that: (a) the vibratory ground motion is felt at the nuclear plant site and recognized as an earthquake HA1

EAL BASES DOCUMENT EBD-H Rev. 11 next HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY PAGE 14 of 30 based on a consensus of control room operators on duty at the time, and (b) for plants with operable seismic instrumentation, the seismic switches of the plant are activated.

The National Earthquake Information Center (1-303-273-8500) can confirm if an earthquake has occurred in the area of the plant.

EAL #2 This EAL addresses 3 potential natural or destructive phenomena that have caused VISIBLE DAMAGE to or results in indication of damage to safety structures, systems, or components containing functions and systems required for safe shutdown of the plant. For the purposes of this EAL, this includes the Emergency Diesel Generator Rooms, Control Building, Reactor Building, Pumphouse, Intake Structure and Condensate Storage Tank Area. Potential natural or destructive phenomena include a tornado strike (touching down), high winds greater than 95 mph or a vehicle crash.

This EAL addresses high wind speeds as measured by the 33-Foot or 156-Foot elevations on the Meteorological Tower. The high winds have caused VISIBLE DAMAGE to or results in indication of damage to safety structures, systems, or components containing functions and systems required for safe shutdown of the plant. The design basis wind speed is 105 miles per hour. However, the meteorological instrumentation is only capable of measuring wind speeds up to 100 miles per hour. Thus the alert level for sustained high wind speed, 95 miles per hour, is selected to be on-scale for the meteorological instrumentation and to conservatively account for potential measurement errors.

This EAL also addresses any vehicle crashes (automobile, aircraft, forklift, train) that results in VISIBLE DAMAGE to or results in indication of damage to safety structures, systems, or components containing functions and systems required for safe shutdown of the plant.

EAL #3 This EAL addresses Turbine failure-generated PROJECTILES affecting safety structures, systems, or components containing functions and systems required for safe shutdown of the plant. For purposes of this EAL, this applies only to those structures where the potential for damage exists from a turbine failure-generated PROJECTILE. This threshold addresses the threat to safety related equipment imposed by PROJECTILES generated by main turbine rotating component failures. Therefore, this EAL is consistent with the definition of an ALERT in that the potential exists for actual or substantial potential degradation of the level of safety of the plant.

EAL #4 HA1

EAL BASES DOCUMENT EBD-H Rev. 11 next HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY PAGE 15 of 30 This EAL addresses the effect of internal flooding that has resulted in degraded performance of systems affected by the flooding, or has created industrial safety hazards (e.g., electrical shock) that preclude necessary access to operate or monitor safety equipment. The inability to operate or monitor safety equipment represents a potential for substantial degradation of the level of safety of the plant. This flooding may have been caused by internal events such as component failures, equipment misalignment, or outage activity mishaps. The site-specific areas include those areas that contain systems required for safe shutdown of the plant, that are not designed to be wetted or submerged.

Safe Shutdown/Vital Areas Category Area 1G31 DG and Day Tank Rooms, 1G21 DG and Day Tank Rooms, Battery Rooms, Essential Switchgear Rooms, Cable Spreading Room Heat Colant Sink!/

Suy Torus Room, Intake Structure, Pumphouse

-Coolant Supply Containment Drywell, Torus Emergency NE, NW, SE Corner Rooms, HPCI Room, RCIC Room, RHR Valve Room, Systems North CRD Area, South CRD Area, CSTs Control Building, Remote Shutdown Panel 1C388 Area, Panel 1C55/56 OArea, SBGT Room Systems of Concern

" Reactivity Control

" Containment (Drywell/Torus)

• RHR/Core Spray/SRVs

• HPCI/RCIC

• RHRSW/River Water/ESW

• Onsite AC Power/EDGs

• Offsite AC Power

" Instrument AC

" DC Power

" Remote Shutdown Capability EAL #5 This EAL addresses other phenomena that result in VISIBLE DAMAGE to a Safe Shutdown/Vital Area or results in indication of damage to safety structures, systems, or components containing functions and systems required for safe shutdown of the plant that can also be precursors of more serious events.

EAL #6 HA1

.AL BASES DOCUMENT EBD-H Rev. 11 next HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY PAGE 16 of 30 This EAL addresses river water levels exceeding design flood water levels. All Seismic Category I structures and non-seismic structures housing Seismic Category I equipment are designed to withstand the hydraulic head resulting from the "maximum probable flood" to which the site could be subjected. The design flood water is at elevation 767.0 ft. Major equipment penetrations in the exterior walls are located above elevation 767.0 ft. Openings below the flood level are either watertight or are provided with means to control the inflow of water in order to ensure that a safe shutdown can be achieved and maintained.

EAL #7 This EAL addresses the effects of loss of river water make-up capability. The intake structure for the safety-related water supply systems (river water, RHR service water, and emergency service water) is located on the west bank of the Cedar River. River levels below the intake structure inlet or a blockage of the intake would result in a loss of the ability to provide make-up water for safety-related systems. The overflow weir is at elevation 724 feet 6 inches. River level at or below this elevation will result in all river flow being diverted to the safety related water supply systems. The top of the intake structure around the pump wells is at elevation 724 feet. If the river water level dropped to this level, the pump suction would have no continuous supply. Blockages of the intake structure may result from debris, ice, or aquatic life. A loss of flow into the intake structure, due to a blockage or low river level, will result in the pit level lowering to the alarm setpoint (723.0 feet) and a resulting alarm in the Control Room.

Therefore, this EAL uses a threshold of low pit level as a potential substantial degradation of the ultimate heat sink capability.

REFERENCES:

1. Abnormal Operating Procedure (AOP) 901, Earthquake

2. Abnormal Operating Procedure (AOP) 902, Flood

3. Abnormal Operating Procedure (AOP) 903, Tornado

4. Abnormal Operating Procedure (AOP) 913, Fire

5. Abnormal Operating Procedure (AOP) 914, Security Events

6. UFSAR Chapter 3, Design of Structures, Components, Equipment, and Systems

7. Bechtel Drawing BECH-M017, Equipment Location - Intake Structure Plans at Elevations, Rev. 6

8. EOP Basis Document, EOP 3 - Secondary Containment Control

9. NEI 99-01 Rev. 5, NEI Methodology for Development of Emergency Action Levels HA1

EAL BASES"DOCUMENT EBD-H Rev. 11 next HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY PAGE 17 of 30 HA2 FIRE or EXPLOSION affecting the operability of plant safety systems required to establish or maintain safe shutdown EVENT TYPE: Fire OPERATING MODE APPLICABILITY: All EAL THRESHOLD VALUE:

HA2.1 FIRE or EXPLOSION resulting in VISIBLE DAMAGE to any Safe ShutdownNital Area or Control Room indication of degraded performance of a System of Concern.

DAEC EAL INFORMATION:

Of particular concern for this EAL are fires that may be detected in any Safe Shutdown/Vital Area.

Damage from fire or explosion can be indicated by physical observation, or by Control Room/local control station instrumentation.

Safe ShutdownVital Areas Category Area Electrical Power 1G31 DG and Day Tank Rooms, 1G21 DG and Day Tank Rooms, Battery Rooms, Essential Switchgear Rooms, Cable Spreading Room Heat Colant Sink /

Suy Torus Room, Intake Structure, Pumphouse Coolant Supply Containment Drywell, Torus Emergency NE, NW, SE Corner Rooms, HPCI Room, RCIC Room, RHR Valve Room, Systems North CRD Area, South CRD Area, CSTs Other Control Building, Remote Shutdown Panel 1C388 Area, Panel 1C55/56

_Area, SBGT Room Per AOP 913, the location of a fire can be determined by observing 1C40B alarm messages, Zone Indicating Unit (ZIU) alarms, or fire annunciators on panels 1C40 and 1C40A.

NOTE Scope of Systems and Equipment of concern was established by review of Appendix R Safe Shutdown credited systems. Only those systems directly affecting safe shutdown or heat removal are listed for consideration, due to fire damage. Support Systems and equipment such as HVAC and specific instrumentation, while included in Appendix R analysis is not considered an immediate threat to the ability to shutdown the plant and remove decay heat.

HA2

EAL BASES DOCUMENT EBD-H Rev. 11 next HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY PAGE 18 of 30 Systemns of Concern

" Reactivity Control

" Containment (Drywell/Torus)

" RHR/Core Spray/SRVs

" HPCI/RCIC

" RHRSW/River Water/ESW

" Onsite AC Power/EDGs

• Offsite AC Power

• Instrument AC

• DC Power

" Remote Shutdown Capability The designation of a single train is intentional and is appropriate when the FIRE / EXPLOSION is large enough to affect more than one component. Lagging fires, fires in waste containers or any miscellaneous fires that may be in the vicinity of safety systems, but do not cause damage to these systems, should NOT be considered for this EAL.

VISIBLE DAMAGE is used to identify the magnitude of the FIRE or EXPLOSION and to discriminate against minor FIRES and EXPLOSIONS.

The reference to structures containing safety systems or components is included to discriminate against FIRES or EXPLOSIONS in areas having a low probability of affecting safe operation.

The significance here is not that a safety system was degraded but the fact that the FIRE or EXPLOSION was large enough to cause damage to these systems.

The use of VISIBLE DAMAGE should not be interpreted as mandating a lengthy damage assessment prior to classification. The declaration of an Alert and the activation of the Technical Support Center will provide the Emergency Director with the resources needed to perform detailed damage assessments.

The Emergency Director also needs to consider any security aspects of the EXPLOSION.

Escalation of this emergency classification level, if appropriate, will be based on System Malfunctions, Fission Product Barrier Degradation or Abnormal Rad Levels / Radiological Effluent ICs.

REFERENCES:

1. Abnormal Operating Procedure (AOP) 913, Fire

2. Abnormal Operating Procedure (AOP) 914, Security Events

3. Abnormal Operating Procedure (AOP) 915, Shutdown Outside Control Room

4. UFSAR Section 6.4, Habitability Systems

5. NEI 99-01 Rev. 5, NEI Methodology for Development of Emergency Action Levels HA2

EAL BASES DOCUMENT EBD-H Rev. 11 next HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY PAGE 19 of 30 HA4 HOSTILE ACTION within the OWNER CONTROLLED AREA or airborne attack threat.

EVENT TYPE: Security OPERATING MODE APPLICABILITY: All EAL THRESHOLD VALUE HA4.1 A HOSTILE ACTION is occurring or has occurred within the OWNER CONTROLLED AREA as reported by DAEC Security Shift Supervision.

HA4.2 A validated notification from NRC of an airliner attack threat within 30 minutes of the site.

DAEC EAL INFORMATION:

Note: Timely and accurate communication between Security Shift Supervision and the Control Room is crucial for the implementation of effective Security EALs.

These EALs address the contingency for a very rapid progression of events, such as that experienced on September 11, 2001. They are not premised solely on the potential for a radiological release. Rather the issue includes the need for rapid assistance due to the possibility for significant and indeterminate damage from additional air, land or water attack elements.

The fact that the site is under serious attack or is an identified attack target with minimal time available for further preparation or additional assistance to arrive requires a heightened state of readiness and implementation of protective measures that can be effective (such as on-site evacuation, dispersal or sheltering).

EAL #1 This EAL addresses the potential for a very rapid progression of events due to a HOSTILE ACTION. It is not intended to address incidents that are accidental events or acts of civil disobedience, such as small aircraft impact, hunters, or physical disputes between employees within the OCA. Those events are adequately addressed by other EALs.

Note that this EAL is applicable for any HOSTILE ACTION occurring, or that has occurred, in the OWNER CONTROLLED AREA. This includes ISFSI's that may be outside the PROTECTED AREA but still within the OWNER CONTROLLED AREA.

HA4

EAL BASES-DOCUMENT EBD-H Rev. 11 next HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY PAGE 20 of 30 Although nuclear plant security officers are well trained and prepared to protect against HOSTILE ACTION, it is appropriate for Offsite Response Organizations to be notified and encouraged to begin activation (if they do not normally) to be better prepared should it be necessary to consider further actions.

If not previously notified by the NRC that the airborne HOSTILE ACTION was intentional, then it would be expected, although not certain, that notification by an appropriate Federal agency would follow. In this case, appropriate federal agency is intended to be NORAD, FBI, FAA or NRC. However, the declaration should not be unduly delayed awaiting Federal notification.

EAL #2 This EAL addresses the immediacy of an expected threat arrival or impact on the site within a relatively short time.

The intent of this EAL is to ensure that notifications for the airliner attack threat are made in a timely manner and that Offsite Response Organizations and plant personnel are at a state of heightened awareness regarding the credible threat. Airliner is meant to be a large aircraft with the potential for causing significant damage to the plant.

This EAL is met when a plant receives information regarding an airliner attack threat from NRC and the airliner is within 30 minutes of the plant.

The NRC Headquarters Operations Officer (HOO) will communicate to the licensee if the threat involves an airliner (airliner is meant to be a large aircraft with the potential for causing significant damage to the plant). The status and size of the plane may be provided by NORAD through the NRC.

REFERENCES:

1. Abnormal Operating Procedure (AOP) 914, Security Events

2. NEI 99-01 Rev. 5, NEI Methodology for Development of Emergency Action Levels

3. NRC Bulletin 2005-02: Emergency Preparedness and Response Actions for Security-Based Events HA4

EAL BASES DOCUMENT EBD-H Rev. 11 next HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY PAGE 21 of 30 HA5 Control room evacuation has been initiated EVENT TYPE: Control Room Evacuation OPERATING MODE APPLICABILITY: All EAL THRESHOLD VALUE:

HA5.1 AOP 915 entered requiring Control Room evacuation.

DAEC EAL INFORMATION:

The applicable procedure for control room evacuation at DAEC is AOP 915.

Evacuation of the Control Room represents a potential for substantial degradation of the level of safety of the plant and therefore requires an ALERT declaration. Additional support, monitoring and direction is required and accomplished by activation of the Technical Support Center at the ALERT classification level. Inability to establish plant control from outside the Control Room will escalate this event to a Site Area Emergency.

REFERENCES:

1. Abnormal Operating Procedure (AOP) 915, Shutdown Outside Control Room

2. UFSAR Section 6.4, Habitability Systems

3. NEI 99-01 Rev. 5, NEI Methodology for Development of Emergency Action Levels HA5

EAL BASES DOCUMENT EBD-H Rev. 11 next HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY PAGE 22 of 30 HA6 Other conditions exist which in the judgment of the Emergency Director (OSM, EC or ER&RD) warrant declaration of an Alert EVENT TYPE: Emergency Director Judgment OPERATING MODE APPLICABILITY: All EAL THRESHOLD VALUE:

HA6.1 Other conditions exist which in the judgment of the Emergency Director (OSM, EC or ER&RD) 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.

DAEC EAL INFORMATION:

This EAL is intended to address 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 Alert emergency class.

REFERENCES:

1. Emergency Plan Implementing Procedure (EPIP) 2.5, Control Room Emergency Response Operations

2. NEI 99-01 Rev. 5, NEI Methodology for Development of Emergency Action Levels HA6

EAL BASES DOCUMENT EBD-H Rev. 11 next PLANT SAFETY HAZARDS & OTHER CONDITIONS AFFECTING PAGE 23 of 30 HS2 Control Room evacuation has been initiated and plant control cannot be established EVENT TYPE: Control Room Evacuation OPERATING MODE APPLICABILITY: All EAL THRESHOLD VALUE:

HS2.1 Control Room evacuation has been initiated.

AND Control of the plant cannot be established per AOP 915 within 20 minutes.

DAEC EAL INFORMATION:

The Emergency Directoris expected to make a reasonable,informed judgment within the 20 minute time limit that control of the plant from the remote shutdown panel has been established.

The intent of this IC is to capture those events where control of the plant cannot be reestablished in a timely manner. In this case, expeditious transfer of control of safety systems has not occurred (although fission product barrier damage may not yet be indicated).

The applicable procedure for control room evacuation at DAEC is AOP 915. Based on the results of the analysis described below, DAEC uses 20 minutes as the site-specific time limit for establishing control of the plant. DAEC has satellite panels associated with the remote shutdown panel at various locations through out the plant. Control of the plant from outside the control room is assumed when the controls are transferred to remote shutdown panel 1C388 in accordance with AOP 915.

The intent of the EAL is to establish control of important plant equipment and knowledge of important plant parameters in a timely manner. Primary emphasis should be placed on those components and instruments that supply protection for and information about safety functions. At a minimum, consistent with the Appendix R safe shutdown analysis described above, these safety functions include reactivity control, maintaining reactor water level, and decay heat removal.

General Electric performed analyses to demonstrate compliance with the requirements of 10 CFR 50 Appendix R for DAEC. The evaluation of Reactor Coolant Inventory was performed using the GE evaluation model (SAFE). The SAFE code determines if the reactor coolant inventory is HS2

EAL BASES DOCUMENT EBD-H Rev. 11 next PLANT SAFETY HAZARDS & OTHER CONDITIONS AFFECTING PAGE 24 of 30 above the TAF during the safe shutdown operation. If core uncovery occurs, the fuel clad integrity evaluation is performed by determining the duration of the core uncovery and the resulting peak cladding temperature (PCT). The PCT calculations were performed by incorporating the SAFE output into the Core Heatup Analysis code (CHASTE). The details of these calculations are provided in Section 4 of the final report for DAEC Appendix R analyses ("Safe Shutdown Appendix R Analyses for Duane Arnold Energy Center", MDE-44-036).

The required analyses include evaluation of the safe shutdown capability of the remote shutdown system for various control room fire events assuming: (1) no spurious operation of equipment, (2) spurious operation of a safety-relief valve (SRV) for 20 minutes, (3) spurious operation of a SRV for 10 minutes, and (4) spurious leakage from a one-inch line. The analyses show that the worst case spurious operation of SRV or isolation valves on a one-inch liquid line (high-low pressure interface) will not affect the safe shutdown ability of the remote shutdown system for DAEC in case of a fire requiring control room evacuation before the identified time limit for the necessary operator actions at the auxiliary shutdown panels. For the limiting cases of worst case spurious leakage from a one-inch line and spurious operation of a SRV, operator control within 20 minutes would not impact the integrity of the fuel clad, the reactor pressure vessel, and the primary containment.

Escalation of this emergency classification level, if appropriate, would be by Fission Product Barrier Degradation or Abnormal Rad Levels/Radiological Effluent EALs.

REFERENCES:

1. Abnormal Operating Procedure (AOP) 915, Shutdown Outside Control Room

2. General Electric Report MDE-44-0386, Safe Shutdown Appendix R Analysis for DAEC, March 1986

3. UFSAR Section 6.4, Habitability Systems

4. NEI 99-01 Rev. 5, NEI Methodology for Development of Emergency Action Levels HS2

EAL BASES DOCUMENT EBD-H Rev. 11 next HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY PAGE 25 of 30 HS3 Other conditions exist which in the judgment of the Emergency Director (OSM, EC, ER&RD) warrant declaration of a Site Area Emergency EVENT TYPE: Emergency Director Judgment OPERATING MODE APPLICABILITY: All EAL THRESHOLD VALUE:

HS3.1 Other conditions exist which in the judgment of the Emergency Director (OSM, EC or ER&RD) 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.

DAEC EAL INFORMATION:

This EAL is intended to address 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 class description for Site Area Emergency.

REFERENCES:

1. Emergency Plan Implementing Procedure (EPIP) 2.5, Control Room Emergency Response Operation

2. NEI 99-01 Rev. 5, NEI Methodology for Development of Emergency Action Levels HS3

- EALBASES DOCUMENT, EBD-H Rev. 11 next HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY PAGE 26 of 30 HS4 HOSTILE ACTION within the PROTECTED AREA EVENT TYPE: Security OPERATING MODE APPLICABILITY: All EAL THRESHOLD VALUE:

HS4.1 A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by DAEC Security Shift Supervision.

DAEC EAL INFORMATION:

This EAL for Site Area Emergency applies to a HOSTILE ACTION within the Plant PROTECTED AREA and Intake PROTECTED AREA. This EAL does not apply to a HOSTILE ACTION occurring within the ISFSI PROTECTED AREA or Switchyard. A HOSTILE ACTION at these locations would be classified under HA4.1 This condition represents an escalated threat to plant safety above that contained in the Alert IC in that a HOSTILE FORCE has progressed from the OWNER CONTROLLED AREA to the PROTECTED AREA.

This EAL is intended to address the contingency for a very rapid progression of events, such as that experienced on September 11, 2001. It is not premised solely on the potential for a radiological release. Rather the issue includes the need for rapid assistance due to the possibility for significant and indeterminate damage from additional air, land or water attack elements.

The fact that the site is under serious attack with minimal time available for further preparation or additional assistance to arrive requires Offsite Response Organizations (OROs) readiness and preparation for the implementation of protective measures.

This EAL is intended to address the potential for a very rapid progression of events due to a HOSTILE ACTION. It is not intended to address incidents that are accidental or acts of civil disobedience, such as small aircraft impact, hunters or physical disputes between employees within the PROTECTED AREA. Those events are adequately addressed by other EALs.

Although vulnerability analyses show nuclear power plants to be robust, it is appropriate for OROs to be notified and to activate in order to be better prepared to respond should protective actions become necessary.

If not previously notified by NRC that the airborne HOSTILE ACTION was intentional, then it would be expected, although not certain, that notification by an appropriate Federal Agency would follow. In this case, appropriate federal agency is intended to be NORAD, FBI, FAA or NRC. However, the declaration should not be unduly delayed awaiting Federal notification.

HS4

EAL BASES DOCUMENT EBD-H Rev. 11 next HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY PAGE 27 of 30 Escalation of this emergency classification level, if appropriate, would be based on actual plant status after impact or progression of attack.

REFERENCES:

1. NEI 99-01 Rev. 5, NEI Methodology for Development of Emergency Action Levels

2. NRC Bulletin 2005-02: Emergency Preparedness and Response Actions for Security-Based Events HS4

EAL BASES DOCUMENT. EBD-H Rev. 11 next HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY PAGE 28 of 30 HG1 HOSTILE ACTION resulting in loss of physical control of the facility EVENT TYPE: Security OPERATING MODE APPLICABILITY: All EAL THRESHOLD VALUE:

HG1.1 A HOSTILE ACTION has occurred such that plant personnel are unable to operate equipment required to maintain safety functions as indicated by loss of physical control of ANY Safe ShutdownNital Area such that operation of equipment required for safe shutdown is lost.

HG1.2 A HOSTILE ACTION has caused failure of Spent Fuel Pool Cooling systems and IMMINENT fuel damage is likely for a freshly off-loaded reactor core in the pool.

DAEC EAL INFORMATION:

EAL #1 This EAL encompasses conditions under which a HOSTILE ACTION has resulted in a loss of physical control of a Safe Shutdownfital Area (containing vital equipment or controls of vital equipment) required to maintain safety functions and control of that equipment cannot be transferred to and operated from another location. Typically, these safety functions are reactivity control (ability to shut down the reactor and keep it shutdown) reactor water level (ability to cool the core), and decay heat removal (ability to maintain a heat sink) for a BWR.

Loss of physical control of the control room or remote shutdown capability alone may not prevent the ability to maintain safety functions per se. Design of the remote shutdown capability and the location of the transfer switches should be taken into account. Primary emphasis should be placed on those components and instruments that supply protection for and information about safety functions.

If control of the plant equipment necessary to maintain safety functions can be transferred to another location, then the above initiating condition is not met.

EAL #2 This EAL addresses failure of spent fuel pool cooling systems as a result of HOSTILE ACTION if IMMINENT fuel damage is likely, such as when a freshly off-loaded reactor core is in the spent fuel pool. IMMINENT fuel damage is based on the Time to Boil (TTB) calculations. During refueling outages, this calculation is done at least daily making it easily obtainable to make a timely classification. During non-refueling times, the TTB calculation is much longer (e.g.

several days) therefore, while the TTB calculation would have to be done, it would not have to be completed immediately.

HG1

EAL BASES DOCUMENT EBD-H Rev. 11 next HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY PAGE 29 of 30

REFERENCES:

1. UFSAR Section 6.4, Habitability Systems

2. NEI 99-01 Rev. 5, NEI Methodology for Development of Emergency Action Levels HG1

EAL BASES DOCUMENT EBD-H Rev. 11 next PLANT SAFETY HAZARDS & OTHER CONDITIONS AFFECTING PAGE 30 of 30 HG2 Other conditions exist which in the judgment of the Emergency Director (OSM, EC, ER&RD) warrant declaration of a General Emergency EVENT TYPE: Emergency Director Judgment OPERATING MODE APPLICABILITY: All EAL THRESHOLD VALUE:

HG2.1 Other conditions exist which in the judgment of the Emergency Director (OSM, EC or ER&RD) 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 off-site for more than the immediate site area.

DAEC EAL INFORMATION:

This EAL is intended to address 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 General Emergency class.

REFERENCES:

1. NUREG-0654/FEMA-REP-1, Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants, Revision 1, October 1980, Appendix 1

2. NEI 99-01 Rev. 5, NEI Methodology for Development of Emergency Action Levels HG2

EAL BASES DOCUMENT.,' EBD-R Rev. 11 next ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT II Usage Level INFORMATION USE PAGE 1 of 28 Effective Date:

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(on designated printers)

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EAL BASES DOCUMENT EBD-R Rev. 11 next ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT PAGE 2 of 28 RU1 Any release of gaseous or liquid radioactivity to the environment GREATER THAN 2 times the Offsite Dose Assessment Manual (ODAM) limit and is expected to continue for 60 minutes or longer EVENT TYPE: Offsite Rad Conditions OPERATING MODE APPLICABILITY: All EAL THRESHOLD VALUE:

Note: The Emergency Director (OSM, EC or ER&RD) should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the release duration has exceeded, or will likely exceed, the applicable time. In the absence of data to the contrary, assume that the release duration has exceeded the applicable time if an ongoing release is detected and the release start time is unknown.

RU1.1 VALID Reactor Building ventilation rad monitor (Kaman 3/4, 5/6, 7/8) or Turbine Building ventilation rad monitor (Kaman 1/2) GREATER THAN 1.0 E-3 pCi/cc and is expected to continue for 60 minutes or longer.

OR RUI1.2 VALID Offgas Stack rad monitor (Kaman 9/10) reading GREATER THAN 2.0 E-1 [iCi/cc and is expected to continue for 60 minutes or longer.

OR RU1.3 VALID LLRPSF rad monitor (Kaman 12) reading GREATER THAN 1.0 E-3 pCi/cc and is expected to continue for 60 minutes or longer.

OR RU1.4 VALID GSW rad monitor (RIS-4767) reading GREATER THAN 3000 (3.0 X 103) CPS and is expected to continue for 60 minutes or longer.

OR RU1.5 VALID RHRSW & ESW rad monitor (RM-1 997) reading GREATER THAN 800 (8.0 X 102)

CPS and is expected to continue for 60 minutes or longer.

OR RU1.6 VALID RHRSW & ESW Rupture Disc rad monitor (RM-4268) reading GREATER THAN 1000 (1.0 X 103) CPS and is expected to continue for 60 minutes or longer.

OR RU1.7 Confirmed sample analyses for gaseous or liquid releases indicates concentrations or release rates GREATER THAN 2 times ODAM limit and is expected to continue for 60 minutes or longer.

RU1

EAL BASES DOCUMENT EBD-R Rev. 11 next ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT PAGE 3 of 28 DAEC EAL INFORMATION:

Valid means that the reading is from instrumentation determined to be operable in accordance with the Technical Specifications or has been verified by other independent methods such as indications displayed on the control panels, reports from plant personnel, or radiological survey results.

This EAL includes any release for which a radioactivity discharge permit was not prepared, or a release that exceeds the conditions (e.g., minimum dilution flow, maximum discharge flow, alarm setpoints, etc.) on the applicable permit. The Emergency Director should not wait until 60 minutes has elapsed, but should declare the event as soon as it is determined that the release duration has met or will likely exceed 60 minutes. Also, if an ongoing release is detected and the starting time for that release is unknown, the Emergency Director should, in the absence of data to the contrary, assume that the release has exceeded 60 minutes.

The approach taken for calculation of gaseous radioactive effluent EAL setpoints includes use of the ODAM Table 3-2 source term computed by BWR-GALE for the DAEC Base Case. The release is assumed to be from a single release point. Multiple release points would be difficult to present as explicit EAL threshold values and in any case, are addressed by off-site dose assessment by MIDAS, which is the preferred method for determining this condition. The calculation methods for setpoint determination are from ODAM Section 3.4 and are based on Regulatory Guide 1.109 methodology. The table below lists the results of the gaseous effluent EAL calculations. The Kaman extended range capability is used because the General Electric Offgas Stack monitor has a limited range.

RU1

EAL BASES DOCUMENT EBD-R Rev. 11 next ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT PAGE 4 of 28 GASEOUS EFFLUENT, EALS Offgas Stack Kaman Turbine Bldg (Kaman Reactor Bldg

• ,*i 9110 i;; *;; .. . " !:*.* 12)

.... Ka m ........ 5/6, 7/8) , ....

• Assumed flow 10,000 72,000 93,000 (CFM)

Release Concentration Release Concentration Release Concentration Release Rate Rate Rate Limits (jCi/cc) (pCi/sec) (gtCi/cc) (j.Ci/sec) (tCi/cc) (iCi/sec)

Tech Sec 1.1E-1 5.2E+5 6.2E-4 2.1E+4 4.8E-4 2.1E+4 Spec Unusual Event 2.OE-1 1.OE+6 1.OE-3 4.2E+4 1.OE-3 4.2E+4 (2 x TS)

Alert (60 6.OE+0 3.OE+7 3.OE-2 1.3E+6 3.OE-2 1.3E+6 (60 x TS)

Assumed flow (CFM) 75,000 Concentration Release Rate (ftCi/cc) ([tCi/sec)

Tech Spec 5.9E-4 2.1E+4 Unusual Event (2 x TS) 1.OE-3 4.2E+4 Alert (200 x TS) 1.OE-1 4.2E+6 RU1

EAL BASES DOCUMENT EBD-R Rev. 11 next ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT PAGE 5 of 28 The off-gas stack is treated as an elevated release and the turbine building and reactor building vents are treated as mixed-mode releases. The ground level setpoints are taken from the default setpoint calculations from the quarterly surveillance tests performed by DAEC Chemistry technicians. Reactor Building, Turbine Building, LLRPSF (Low Level Radwaste Processing and Storage Facility) and Offgas Stack Noble Gas Monitor alarm setpoints are calculated based on achieving the Tech Spec/ODAM instantaneous release limit, assuming annual average meteorology as defined in the ODAM. The monitor alarm setpoint can be periodically adjusted but typically does not vary by much. The DAEC EAL therefore addresses valid radiation levels exceeding 2 times the alarm setpoint for greater than 60 minutes.

For the Turbine Building, 2X the TS setpoint is 1.2E-3 ptCi/cc. Rounded off, this corresponds to 1E-3 .iCi/cc.

For the Reactor Building, 2X the TS setpoint is 9.6E-4 jLCi/cc. Rounded off, this also corresponds to 1 E-3 [tCi/cc.

For the Offgas Stack, 2X the TS setpoint is 2.2E-1 pCi/cc. Rounded off, this also corresponds to 2E-1 p.Ci/cc.

For the LLRPSF Building, 2X the TS setpoint is 1.2E-3 jiCi/cc. Rounded off, this corresponds to 1E-3 [tCi/cc.

Technical specification setpoints for radioactive liquid radiation monitors are 10 times the 10 CFR 20 Appendix B, Table 2, Water Effluent Concentration (WEC) limits. It is the policy of DAEC to process all liquid radwaste so that no release of radioactive liquid to the environment is allowed. The radwaste effluent line which could be used as a batch release mechanism has a trip function that prevents exceeding the DAEC release limit, however, an EAL has been provided. The other pathways to the environment (RHRSW - to cooling tower, RHRSW - to discharge canal) have radiation monitors with readouts going to the Control Room. These systems could become contaminated if heat exchanger leaks develop; however, historically this has not occurred in the service water systems at DAEC. These monitors are displayed on panels 1C02 and 1C10.

Reactor water is the likely source of contamination through the service water systems as opposed to floor drain, detergent drain, and chemical waste discharge. The floor drain and detergent drains go to Radwaste Processing and would be batch released to the Radwaste effluent discharge line (if such a release were to occur). The chemical discharge sump is normally a radioactivity clean system and is tested by Chemistry to ensure no contamination prior to discharging to the canal.

The setpoints for the three service water radiation effluent monitors vary because of differences in detector efficiencies and background. Setpoints based on the same reactor water sample are listed below to show the differences. The rounded off readings will be used for the EALs for ease of reading the monitor scales.

RU1

EAL BASES DOCUMENT EBD-R Rev. 11 next EFFLUENT ABNORMAL RAD LEVELS/RADIOLOGICAL PAGE 6 of 28 Monitor' TS Limit Reading Unusual Alert Level

_..._ .. _.-

• Event Level 1,555 1.5X 103 3.0 X 103 3.0 X 105 GSW CPS CPS CPS CPS 413 4.0X 102 8.0X 102 8.0X RHRSW & ESW to cooling tower CPS CPS CPS CPS 507 5.0 X 102 1.0 X 103 1.0 X 105 RHRSW & ESW to Discharge Canal CPS CPS CPS CPS DAEC does not have a telemetered radiation monitoring system or an automatic real-time dose assessment system.

REFERENCES:

1. Offsite Dose Assessment Manual

2. Emergency Plan Implementing Procedure (EPIP) 3.3, Dose Assessment and Protective Action

3. Radiation Protection Calculation No. 95-001-C, Emergency Actions Levels Based on Effluent Radiation Monitors, January 24, 1995

4. UFSAR Section 11.5, Process and Effluent Radiation Monitoring and Sampling Systems

5. NEI Methodology for Development of Emergency Action Levels NEI 99-01 Revision 5, February 2008

6. Radiological Engineering Calculation No. 07-002C, Calculated Default Setpoint Value for the Reactor Building KAMAN with a Realistic Building Flow Rate, November 29, 2007 RU1

EAL BASES DOCUMENT EBD-R Rev. 11 next ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT PAGE 7 of 28 RU2 UNPLANNED rise in plant radiation levels EVENT TYPE: Onsite Rad Conditions OPERATING MODE APPLICABILITY: All EAL THRESHOLD VALUE:

RU2.1 UNPLANNED VALID Refuel Floor ARM reading rise with an UNPLANNED water level drop of reactor cavity, fuel pool, or fuel transfer canal as indicated by ANY of the following:

" Report to control room

• VALID fuel pool level indication (LI-3413) LESS THAN 36 feet and lowering

• VALID WR GEMAC Floodup indication (LI-4541) coming on scale.

OR RU2.2 Any UNPLANNED VALID ARM reading GREATER THAN 1000 times normal*.

RU2.3 Any UNPLANNED VALID radiation survey results GREATER THAN 1000 times normal* levels.

• Normal levels can be considered as the highest readingin the past twenty-four hours excluding the currentpeak value.

DAEC EAL INFORMATION:

Unplannedmeans that the condition is not the result of planned actions by the plant staff in accordance with procedures. Valid means that the reading is from instrumentation determined to be operable in accordance with the Technical Specifications or has been verified by other independent methods such as indications displayed on the control panels, reports from plant personnel, or radiological survey results.

There are three methods to determine water level decreases of concern. The first method is by report to the control room. The other methods include use of the Floodup level indicator and the spent fuel pool level indicator. These are further described below.

RU2

EAL BASES DOCUMENT. EBD-R Rev. 11 next ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT PAGE 8 of 28 During preparation for reactor cavity flood up prior to entry into refuel mode, reactor vessel level instrument LI-4541 (WR GEMAC, FLOODUP) on control room panel 1C04 is placed in service by I&C personnel connecting a compensating air signal after the reference leg is disconnected from the reactor head. Normal refuel water level is above the top of the span of this flood up level indicator. A valid indication (e.g., not due to loss of compensating air signal or other instrument channel failure) of reactor cavity level coming on span for this instrument is used at DAEC as an indicator of uncontrolled reactor cavity level decrease.

DAEC Technical Specifications require a minimum of 36 feet of water in the spent fuel pool when moving irradiated fuel into the secondary containment. During refueling, the gates between the reactor cavity and the refueling cavity are removed and the spent fuel pool level indicator LI 3413 is used to monitor refueling water level. Procedures require that a normal refueling water level be maintained at 37 feet 5 inches. A low level alarm actuates when spent fuel pool level drops below 37 feet 1 inch. Symptoms of inventory loss at DAEC include visual observation of decreasing water levels in reactor cavity or spent fuel storage pool, Reactor Building (RB) fuel storage pool radiation monitor or refueling area radiation monitor alarms, observation of a decreasing trend on the spent fuel pool water level indicator, and actuation of the spent fuel pool low water level alarm.

To eliminate minor level perturbations from concern, DAEC uses L13413 indicated water level below 36 feet and lowering.

Increased radiation levels can be detected by the local refueling floor area radiation monitors, the refueling floor Continuous Air Monitor (CAM) alarm, refueling areas radiation monitors, fuel pool ventilation exhaust monitors, and by Standby Gas Treatment (SGBT)

System automatic start. Applicable area radiation monitors include those that are displayed on Panel 1C02 and alarmed on Panel 1C04B. The DAEC EAL has also been written to reflect the case where an ARM may go offscale high prior to reaching 1,000 times the normal reading.

EALs 2 and 3 address increases in plant radiation levels that represent a loss of control of radioactive material resulting in a potential degradation in the level of safety of the plant.

NOTE: On Annunciator Panel 1C04B, the indicators listed below are expected alarms during pre-planned transfers of highly radioactive material through the affected area. If an HP Technician is present, sending an Operator is not required. Radiation levels other than those expected should be promptly investigated. The indicators are high radiation alarms from the Hot Laboratory or Administrative Building, the new fuel storage area, and the radwaste building.

RU2

EAL BASES DOCUMENT EBD-R Rev. 11 next ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT PAGE 9 of 28

REFERENCES:

1. Alarm Response Procedure (ARP) 1C04B, Reactor Water Cleanup and Isolation

2. Technical Specification 3.7.8, Spent Fuel Pool Water Level

3. Emergency Plan Implementing Procedure (EPIP) Form TSC-40, ARM Locations

4. Emergency Operating Procedures (EOP) Basis Document, Breakpoints for RC/L & L

5. Surveillance Test Procedure (STP) 3.0.0.0-01 PA, Daily and Shift Instrument Checks

6. Integrated Plant Operating Instruction (IPOI) 8, Outage and Refueling Operations

7. Core Alterations, RFP403, Procedure for Moving Core Components Between Reactor Core and Spent Fuel Pool, Within the Reactor Core, or Within the Spent Fuel Pool

8. NEI Methodology for Development of Emergency Action Levels NEI 99-01 Revision 5, February 2008.

RU2

EAL BASES DOCUMENT EBD-R Rev. 11 next ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT PAGE 10 of 28 RA1 Any release of gaseous or liquid radioactivity to the environment GREATER THAN 200 times the Offsite Dose Assessment Manual (ODAM) limit and is expected to continue for 15 minutes or longer EVENT TYPE: Offsite Rad Conditions OPERATING MODE APPLICABILITY: All EAL THRESHOLD VALUE:

Note: The Emergency Director (OSM, EC or ER&RD) should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the release duration has exceeded, or will likely exceed, the applicable time. In the absence of data to the contrary, assume that the release duration has exceeded the applicable time if an ongoing release is detected and the release start time is unknown.

RA1.1 VALID Reactor Building ventilation rad monitor (Kaman 3/4, 5/6, 7/8) or Turbine Building ventilation rad monitor (Kaman 1/2) reading GREATER THAN 3.0 E-2 [tCi/cc and is expected to continue for 15 minutes or longer.

OR RA1.2 VALID Offgas Stack rad monitor (Kaman 9/10) reading GREATER THAN 6.0 E+0 [tCi/cc and is expected to continue for 15 minutes or longer.

OR RA1.3 VALID LLRPSF rad monitor (Kaman 12) reading GREATER THAN 1.0 E-1 pCi/cc and is expected to continue for 15 minutes or longer.

OR RA1.4 VALID GSW rad monitor (RIS-4767) reading GREATER THAN 300,000 (3.0 X 105)

CPS and is expected to continue for 15 minutes or longer.

OR RA1.5 VALID RHRSW & ESW rad monitor (RM-1997) reading GREATER THAN 80,000 (8.0 X 104) CPS and is expected to continue for 15 minutes or longer.

OR RA1.6 VALID RHRSW & ESW Rupture Disc rad monitor (RM-4268) reading GREATER THAN 100,000 (1.0 X 105) CPS and is expected to continue for 15 minutes or longer.

OR RA1.7 Confirmed sample analyses for gaseous or liquid releases indicates concentrations or release rates GREATER THAN 200 times ODAM limit and is expected to continue for 15 minutes or longer.

RA1

L BASES DOCUMENT EBD-R Rev. 11 next ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT PAGE 11 of 28 DAEC EAL INFORMATION:

Valid means that the reading is from instrumentation determined to be operable in accordance with the Technical Specifications or has been verified by other independent methods such as indications displayed on the control panels, reports from plant personnel, or radiological survey results. In a case where data from Kaman readings is being used to determine whether an EAL threshold value has been exceeded, Valid means that flow through the associated Kaman Monitor has been verified and does exist as indicated in pCi/sec on SPRAD (Safety Parameter Display System (SPDS) screen).

This EAL includes any release for which a radioactivity discharge permit was not prepared, or a release that exceeds the conditions (e.g., minimum dilution flow, maximum discharge flow, alarm setpoints, etc.) on the applicable permit. The Emergency Director should not wait until 15 minutes has elapsed, but should declare the event as soon as it is determined that the release duration has met or will likely exceed 15 minutes. Also, if an ongoing release is detected and the starting time for that release is unknown, the Emergency Director should, in the absence of data to the contrary, assume that the release has exceeded 15 minutes.

RA1

EAL BASES DOCUMENT EBD-R Rev. 11 next ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT PAGE 12 of 28 GASEOUS EFFLUENT EALS Offgas Stack Kaman Turbed( a 2 Reactor Bldg,',,,

9110 (Kaman 3/4,45/6, 7/8)

Assumed 10,000 72,000 93,000 Release Release Release Release Concentratio Rate Concentration Rate Concentration Rate Limits n (jiCi/cc) (Ci/sec) ([ICi/cc) (Ci/sec) (pci/cc) (aCi/sec)

Tech Spec 1.1E-1 5.2E+5 6.2E-4 2.1E+4 4.8E-4 2.1E+4 Unusual Event 2.OE-1 1.OE+6 1.OE-3 4.2E+4 1.OE-3 4.2E+4 (2 x TS) I _III Alert (60 6.OE+0 3.OE+7 3.OE-2 1.3E+6 3.OE-2 1.3E+6 (60 x TS)

LLRPSF Kaman 12 .

Assumed flow (CFM) 75,000 Concentration Release Rate (pCi/cc) (pCi/sec)

Tech Spec 5.9E-4 2.1E+4 Unusual Event (2 x TS) 1.OE-3 4.2E+4 Alert (200 x TS) 1.OE-1 4.2E+6 The off-gas stack is treated as an elevated release and the turbine building and reactor building vents are treated as mixed-mode releases. The ground level setpoints are taken from the default setpoint calculations from the quarterly surveillance tests performed by DAEC Chemistry technicians. Reactor Building, Turbine Building, LLRPSF (Low Level Radwaste Processing and Storage Facility) and Offgas Stack Noble Gas Monitor alarm setpoints are calculated based on achieving the Tech Spec/ODAM instantaneous release limit, assuming annual average meteorology as defined in the ODAM. The monitor alarm setpoint can be periodically adjusted but typically does not vary by much. For the Offgas Stack, Reactor Building and Turbine building KAMAN monitor readings, DAEC chose to multiply the technical specification concentration by a factor of 60 (instead of 200) in order to allow for a logical step progression in monitor setpoints from the RU1 through RA1 to RS1. For the LLRPSF monitor, the DAEC EAL addresses valid radiation levels exceeding 200 times the alarm setpoint for 15 minutes or longer.

RAI

EAL BASES DOCUMENT EBD-R Rev. 11 next ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT PAGE 13 of 28 For the Turbine Building, 60X the TS setpoint is 3.7E-2 ýtCi/cc. Rounded off, this corresponds to 3E-2 p.Ci/cc.

For the Reactor Building, 60X the TS setpoint is 2.9E-2 pCi/cc. Rounded off, this also corresponds to 3E-2 jtCi/cc.

For the Offgas Stack, 60X the TS setpoint is 6.6E+0 [tCi/cc. Rounded off, this also corresponds to 6E+O [tCi/cc.

For the LLRPSF Building, 200X the TS setpoint is 1.2E-1 itCi/cc. Rounded off, this corresponds to 1E- 1 pCi/cc.

Technical specification setpoints for radioactive liquid radiation monitors are 10 times the 10 CFR 20 Appendix B, Table 2, Water Effluent Concentration (WEC) limits. It is the policy of DAEC to process all liquid radwaste so that no release of radioactive liquid to the environment is allowed.

The radwaste effluent line which could be used as a batch release mechanism has a trip function that prevents exceeding the DAEC release limit, and therefore no EAL limits are provided. The other pathways to the environment (RHRSW - to cooling tower, RHRSW - to discharge canal) have radiation monitors with readouts going to the Control Room. These systems could become contaminated if heat exchanger leaks develop; however, historically this has not occurred in the service water systems at DAEC. These monitors are displayed on panels 1C02 and 1C10.

Reactor water is the likely source of contamination through the service water systems as opposed to floor drain, detergent drain, and chemical waste discharge. The floor drain and detergent drains go to Radwaste Processing and would be batch released to the Radwaste effluent discharge line (if such a release were to occur). The chemical discharge sump is normally a radioactivity clean system and is tested by Chemistry to ensure no contamination prior to discharging to the canal.

The setpoints for the three service water radiation effluent monitors vary because of differences in detector efficiencies and background. Setpoints based on the same reactor water sample are listed below to show the differences. The rounded off readings will be used for the EALs for ease of reading the monitor scales.

RA1

EAL BASES DOCUMENT EBD-R Rev. 11 next ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT PAGE 14 of 28 Monitor TS Limmit Reading Unusual Alert Level Event Level 1,555 1.5 X 103 3.0 X 103 3.0 X 105 GSW CPS CPS CPS CPS 413 4.0X 102 8.0X 102 8.0 X 104 RHRSW & ESW to cooling tower CPS CPS CPS CPS 507 5.0X 102 1.0X 103 1..X 105 RHRSW & ESW to Discharge Canal CPS CPS CPS CPS DAEC does not have a telemetered radiation monitoring system or an automatic real-time dose assessment system.

REFERENCES:

1. Offsite Dose Assessment Manual Section 6.0, 6.1.2 and 7.1.2 Bases

2. Emergency Plan Implementing Procedure (EPIP) 3.3, Dose Assessment and Protective Action

3. Radiation Protection Calculation No. 95-001-C, Emergency Actions Levels Based on Effluent Radiation Monitors, January 24, 1995

4. UFSAR Section 11.5, Process and Effluent Radiation Monitoring and Sampling Systems

5. EPA 400-R-92-001, Manual of Protective Action Guides and Protective Actions for Nuclear Incidents

6. NEI Methodology for Development of Emergency Action Levels NEI 99-01 Revision 5, February 2008

7. Radiological Engineering Calculation No. 07-002C, Calculated Default Setpoint Value for the Reactor Building KAMAN with a Realistic Building Flow Rate, November 29, 2007 RA1

EAL BASES DOCUMENT EBD-R Rev. 11 next ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT PAGE 15 of 28 RA2 Damage to spent fuel or loss of water level that has resulted or will result in the uncovering of spent fuel outside the reactor vessel EVENT TYPE: Onsite Rad Conditions OPERATING MODE APPLICABILITY: All EAL THRESHOLD VALUE:

RA2.1 Report of ANY of the following due to damage to spent fuel or loss of water level:

a. VALID Hi Rad alarm for ANY of the following ARMs:

" RM-9163 (Refueling Floor North End)

• RM-9164 (Refueling Floor South End)

" RM-9153 (New Fuel Storage)

" RM-9178 (Spent Fuel Storage Area).

OR

b. VALID reading GREATER THAN 10 millirem/hr for ANY of the following ARMs:

" RM-9163 (Refueling Floor North End)

• RM-9164 (Refueling Floor South End)

• RM-9153 (New Fuel Storage Area)

OR

c. VALID reading GREATER THAN 100 millirem/hr for ARM RM-9178 (Spent Fuel Storage Area)

OR RA2.2 VALID WR GEMAC Floodup indication (LI-4541) LESS THAN 450 inches that will result in spent fuel becoming uncovered.

DAEC EAL INFORMATION:

Valid means that the reading is from instrumentation determined to be operable in accordance with the Technical Specifications or has been verified by other independent methods such as indications displayed on the control panels, reports from plant personnel, RA2

EAL BASES DOCUMENT EBD-R Rev. 11 next ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT PAGE 16 of 28 or radiological survey results. Valid alarms are solely due to damage to irradiated fuel or loss of water level that has or will result in the uncovering of irradiated fuel.

There are no significant deviations from the generic EALs. Increased radiation levels can be detected by the local radiation monitors, in-plant radiological surveys, new fuel and spent fuel storage area radiation monitor alarms displayed on panel 1C04B, fuel pool ventilation exhaust monitors, and by Standby Gas Treatment (SBGT) System automatic start. Applicable area radiation monitors include RM-9163, RM-9164, RM-9153, and RM-9178. These monitors are located in the north end of the refuel floor, the south end of the refuel floor, the new fuel vault area, and near the spent fuel pool, respectively.

Per ARP 1C04B, the applicable area radiation monitor alarms actuate when radiation levels increase above 100 millirem/hr in the spent fuel pool area or above 10 millirem/hr in the other three areas of concern. If a valid actuation of these alarms were to occur, the refueling floor would be immediately evacuated. Thus, a report of a fuel handling accident with either valid actuation of the fuel area alarms on panel 1C04B or with measured radiation levels in the spent fuel pool or north fuel area are used to address the generic concern consistent with DAEC design and procedures.

During preparation for reactor cavity flood up prior to entry into refuel mode, reactor vessel level instrument LI-4541 (VAR GEMAC, FLOODUP) on control room panel 1C04 is placed in service by I&C personnel connecting a compensating air signal after the reference leg is disconnected from the reactor head. Normal refuel water level is above the top of the span of this flood up level indicator. A valid on-scale indication (e.g., not due to loss of compensating air signal or other instrument channel failure) from this instrument can be used to determine uncontrolled loss of water level in the reactor cavity.

During refueling, the gates between the reactor cavity and the refueling cavity are removed and the spent fuel pool level indicator LI 3413 is used to monitor refueling water level.

This measures the common water level in the reactor cavity and the fuel pool. The bottom of the fuel transfer canal between the spent fuel pool and the reactor cavity is 16 feet above the bottom of the spent fuel pool. The top of the active fuel in the spent fuel storage racks is slightly less than 13 feet 9 inches above the bottom of the spent fuel pool.

Therefore, postulated failures which drain the reactor cavity through the reactor vessel cannot uncover fuel in the spent fuel storage racks. However, valid indication of spent fuel pool level less than 16 feet would indicate that spent fuel in the storage racks may potentially become uncovered. Radiological Engineering Calculation #10-002A determined that a spent fuel pool level of just over 16 feet, with 144 bundles in the pool, would result in a dose rate of 100 millirem/hr on RE-9178. Therefore, there is no specific EAL related to fuel pool level indicator LI-3413 since the ARM threshold of this EAL would be met, prior to reaching 16 feet.

RA2

EA-LEBASES DOCUMENT. EBD-R Rev. 11 next ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT PAGE 17 of 28 Any loss of level resulting in the uncovering of irradiated fuel, regardless if in the Fuel Transfer Canal, Fuel Pool, or Reactor Refueling Cavity, will cause the ARMs to alarm.

RFP403 requires that upon a loss of water level situation, that the refueling crew on the refueling floor shall discharge any fuel assembly on the fuel grapple as follows:

• If a fuel assembly is currently being withdrawn from a slot in the core or spent fuel pool, immediately reinsert it into that slot.

" If a fuel assembly is being transferred and is still over or near the core, insert it into the closest available slot in the core.

• If a fuel assembly is being transferred and is over or near the spent fuel pool, insert it into the closest available slot in the spent fuel racks.

Following these actions, the refueling floor is to be evacuated of all personnel. The DAEC EAL is written to address the generic concern that a spent fuel assembly was not fully covered by water. This can either be by visual observation of an uncovered spent fuel assembly or by trending fuel pool level in the control room if a spent fuel assembly could not be placed in a safe storage location specified by RFP 403 as described above.

REFERENCES:

1. Alarm Response Procedure (ARP) 1C04B, Reactor Water Cleanup and Recirculation

2. Technical Specification 3.7.8, Spent Fuel Pool Water Level

3. Emergency Operating Procedures (EOP) Basis Document, Breakpoints for RC/L &

L

4. Emergency Plan Implementing Procedure (EPIP) Form TSC-40 ARM Locations

5. Surveillance Test Procedure (STP) 3.0.0.0-01, Daily and Shift Instrument Checks

6. Integrated Plant Operating Instruction (IPOI) 8, Outage and Refueling Operations

7. Core Alterations, RFP403, Procedure for Moving Core Components Between Reactor Core and Spent Fuel Pool, Within the Reactor Core, or Within the Spent Fuel Pool

8. Bechtel Drawing C-492, Reactor Building - Reactor Well, Spent Fuel & Dryer-Separator Pool General Arrangement, Rev. 6

9. Bechtel Drawing C-493, Reactor Building - Spent Fuel Liner Plan Elevations and Details, Sheet 1, Rev. 6

10. Holtec International Drawing No. 1045, Rack Construction - Spent Fuel Storage Racks, Rev. 3

11. NEI Methodology for Development of Emergency Action Levels NEI 99-01 Revision 5, February 2008 RA2

EAL BASES DOCUMENT EBD-R Rev. 11 next ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT PAGE 18 of 28

12. Radiological Engineering Calculation No. 10-002A, "Determine radiation dose rate at the Spent Fuel Pool ARM with the water in the Fuel Pool drained to a pool depth of 16 feet and only 144 bundles in the pool", February 17, 2010.

RA2

EAL BASES DOCUMENT EBD-R Rev. 11 next ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT PAGE 19 of 28 RA3 Rise in radiation levels within the facility that impedes operation of systems required to maintain plant safety functions EVENT TYPE: Onsite Rad Conditions OPERATING MODE APPLICABILITY: All EAL THRESHOLD VALUE:

RA3.1 VALID area radiation levels GREATER THAN 15 millirem/hr in ANY of the following areas:

• Control Room ARM (RM-9162)

• Central Alarm Station (by survey)

DAEC EAL INFORMATION:

Valid means that the reading is from instrumentation determined to be operable in accordance with the Technical Specifications or has been verified by other independent methods such as indications displayed on the control panels, reports from plant personnel, or radiological survey results.

There are no significant deviations from the generic EALs. The control room and the Central Alarm Station (CAS) are the only areas that are required to be continuously occupied to achieve and maintain safe operation or safe shutdown.

Expected increases in monitor readings due to controlled evolutions (such as lifting the steam dryer during refueling) do not result in emergency declaration. Nor should momentary increases due to events such as resin transfers or controlled movement of radioactive sources result in emergency declaration. In-plant radiation level increases that would result in emergency declaration, are also unplanned,e.g., outside the limits established by an existing radioactive discharge permit.

REFERENCES:

1. Alarm Response Procedure (ARP) 1C04B, Reactor Water Cleanup and Isolation

2. Surveillance Test Procedure (STP) 3.0.0.0-01, Daily and Shift Instrument Checks

3. Integrated Plant Operating Instruction (IPOI) 8, Outage and Refueling Operations

4. Emergency Plan Implementing Procedure (EPIP) 3.1, Inplant Radiological Monitoring

5. UFSAR Section 6.4, Habitability Systems RA3

EAL BASES DOCUMENT EBD-R Rev. 11 next ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT PAGE 20 of 28

6. Bechtel Calculation DA-4, Project Number 265-002, Control Room Habitability, 9/3/80

7. NEI Methodology for Development of Emergency Action Levels NEI 99-01 Revision 5, February 2008 RA3

EAL BASES DOCUMENT EBD-R Rev. 11 next ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT PAGE 21 of 28 RSI Offsite dose resulting from an actual or IMMINENT release of gaseous radioactivity GREATER THAN 100 millirem TEDE or 500 millirem CDE thyroid for the actual or projected duration of the release EVENT TYPE: Offsite Rad Conditions OPERATING MODE APPLICABILITY: All EAL THRESHOLD VALUE:

Note: The Emergency Director (OSM, EC or ER&RD) should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the condition will likely exceed the applicable time. If dose assessment results are available, declaration should be based on dose assessment instead of radiation monitor values. Do not delay declaration awaiting dose assessment results.

RS1.1 Dose assessment using actual meteorology indicates doses GREATER THAN 100 millirem TEDE or 500 millirem thyroid CDE at or beyond the site boundary. (Preferred method)

OR RS1.2 If Dose Assessment is unavailable, either of the following:

• Valid Reactor Building ventilation rad monitor (Kaman 3/4, 5/6, 7/8) or Turbine Building ventilation rad monitor (Kaman 1/2) reading GREATER THAN 6.0 E-2 [tCi/cc and is expected to continue for 15 minutes or longer.

• Valid Offgas Stack rad monitor (Kaman 9/10) reading GREATER THAN 4.0 E+1 pCi/cc and is expected to continue for 15 minutes or longer.

OR RS1.3 Field survey results indicate closed window dose rates GREATER THAN 100 millirem/hr and is expected to continue for 60 minutes or longer at or beyond the site boundary.

OR RS1.4 Analyses of field survey samples indicate thyroid CDE GREATER THAN 500 millirem for one hour of inhalation at or beyond the site boundary.

RS1

EAL BASES DOCUMENT EBD-R Rev. 11 next ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT PAGE 22 of 28 DAEC EAL INFORMATION:

Valid means that the reading is from instrumentation determined to be operable in accordance with the Technical Specifications or has been verified by other independent methods such as indications displayed on the control panels, reports from plant personnel, or radiological survey results. In a case where data from Kaman readings is being used to determine whether an EAL threshold value has been exceeded, Valid means that flow through the associated Kaman Monitor has been verified and does exist as indicated in pCi/sec on SPRAD.

The preferred method for declaration of RS1 is by means of Dose Assessment using the MIDAS computer model. If dose assessment results are available at the time of declaration, the classification should be based on RS1.1. However, if Kaman monitor readings are sustained for 15 minutes or longer and the required MIDAS dose assessments cannot be completed within this period, then the declaration can be made using Kaman readings PROVIDED the readings are not from an isolated flow path. If Kaman readings are not valid, field survey results may be utilized.

DAEC's Meteorological Information and Dose Assessment System (MIDAS) was utilized to determine the Kaman monitor limits. Eight separate combinations of release point, source term, meteorological conditions and equipment status were analyzed. Pathways considered were the offgas stack, the turbine building exhaust vent and a single reactor building exhaust vent. Multiple release points were not considered. In this same vein, it was assumed that only one of the three reactor building vents is on during the release.

The source terms used have been pre-loaded into MIDAS and are the default mixes associated with a loss of coolant accident (LOCA) and a control rod drop (CRD). The LOCA mix was used in conjunction with a release via the offgas stack while the CRD mix was used for releases via the turbine or reactor building vents. The source term for a release via the offgas stack is further impacted by the status of the standby gas treatment system. The status of that system was also taken into consideration.

Based on 1995 data (NG-96-0987), the atmospheric stability was classified as Pascal E 33% of the time. Consequently, both classifications were evaluated. Based on the same report, the most common wind speeds were:

Pascal Class Altitude Speed (mph)

D 156 feet 8-12 D 33 feet 8-12 E 156 feet 8-12 E 33 feet 4-7 RS1

EAL BASES DOCUMENT. EBD-R Rev. 11 next ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT PAGE 23 of 28 Though the temperature setting has no impact on the MIDAS calculations, a value must be entered in order for the program to run. Consequently, the temperature was arbitrarily set at 50 F.

The rain estimate was set at zero, to eliminate any on site washout of radioactive material.

For the first MIDAS runs a 1Ci/cc concentration was assumed. The results of these runs were then normalized to the limits, thus generating a theoretical Kaman limit. Additional MIDAS runs were made with these theoretical limits as input to verify the normalization process.

In addition to the total integrated dose, MIDAS calculates a peak whole body DDE rate resulting from the plume and a peak thyroid CDE rate resulting from inhalation. Because the RS1 and RG1 KAMAN limits are to be based on a one-hour exposure, establishing concentration limits so these peak values match the NUMARC limits is acceptable.

iltiating". Condti on ste Area Emergency General Emergency RSI.RG Valid Turbine or Reactor Building ventilation rad monitor (KAMAN) reading, 0.06 !iCi/cc 0.6 gtCi/cc for 15 minutes or longer, above:

Valid Offgas Stack ventilation rad monitor (Kaman) reading, for 15 minutes or 40 1 iCi/cc 400 1.Ci/cc longer, above:

DAEC does not have a telemetered radiation monitoring system.

Dose assessment using MIDAS is based on the EPA-400 methodology, e.g., use of Total Effective Dose Equivalent (TEDE) and Committed Dose Equivalent (CDE) Thyroid. TEDE is somewhat different from whole body dose from gaseous effluents determined by ODAM methodology which forms the basis for the radiation monitor readings calculated in RU1.

These factors can introduce differences that are at least as large as those introduced by using TEDE versus whole body dose. The gaseous effluent radiation monitors can only detect noble gases. The contribution of iodines to TEDE and CDE Thyroid could therefore only be determined either by: (1) utilizing the source term mixture in MIDAS, or (2) gaseous effluent sampling. Therefore, DAEC EAL Threshold Value 1 is written in terms of TEDE and CDE Thyroid.

RS1

EAL BASES DOCUMENT EBD-R Rev. 11 next ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT PAGE 24 of 28

REFERENCES:

1. Offsite Dose Assessment Manual, Section 6.0, 6.1.2 and 7.1.2, Bases

2. Emergency Plan Implementing Procedure (EPIP) 3.3, Dose Assessment and Protective Action

3. Radiation Protection Calculation No. 95-001-C, Emergency Actions Levels Based on Effluent Radiation Monitors, January 24, 1995

4. Radiation Engineering Calculation No. 96-007-A, Determination of DAEC Radioactive Release Initiating Conditions for AS1, & AGI Emergency Classifications, July 3, 1996

5. UFSAR Section 11.5, Process and Effluent Radiation Monitoring and Sampling Systems

6. EPA 400-R-92-001, Manual of ProtectiveAction Guides and Protective Actions for Nuclear Incidents

7. NEI Methodology for Development of Emergency Action Levels NEI 99-01 Revision 5, February 2008 RS1

EAL BASES DOCUMENT. .: EBD-R Rev. 11 next EFFLUENT ABNORMAL RAD LEVELS/RADIOLOGICAL PAGE 25 of 28 RG10 ffsite dose resulting from an actual or IMMINENT release of gaseous radioactivity GREATER THAN 1000 millirem TEDE or 5000 millirem thyroid CDE for the actual or projected duration of the release EVENT TYPE: Offsite Rad Conditions OPERATING MODE APPLICABILITY: All EAL THRESHOLD VALUE:

Note: The Emergency Director (OSM, EC or ER&RD) should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the condition will likely exceed the applicable time. If dose assessment results are available, declaration should be based on dose assessment instead of radiation monitor values. Do not delay declaration awaiting dose assessment results.

RG1.1 Dose assessment using actual meteorology indicates doses GREATER THAN 1000 millirem TEDE or 5000 millirem thyroid CDE at or beyond the site boundary. (Preferred method)

OR RG1.2 If Dose Assessment is unavailable, either of the following:

• VALID Reactor Building ventilation rad monitor (Kaman 3/4, 5/6, 7/8) or Turbine Building ventilation rad monitor (Kaman 1/2) reading GREATER THAN 6.0 E-1 1.Ci/cc and is expected to continue for 15 minutes or longer.

• VALID Offgas Stack rad monitor (Kaman 9/10) reading GREATER THAN 4.0 E+2 [tCi/cc and is expected to continue for 15 minutes or longer.

OR RG1.3 Field survey results indicate closed window dose rates GREATER THAN 1000 millirem/hr and is expected to continue for 60 minutes or longer at or beyond the site boundary.

OR RG1.4 Analyses of field survey samples indicate thyroid CDE GREATER THAN 5000 millirem for one hour of inhalation at or beyond the site boundary.

RG1

EAL BASES DOCUMENT EBD-R Rev. 11 next ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT PAGE 26 of 28 DAEC EAL INFORMATION:

Valid means that the reading is from instrumentation determined to be operable in accordance with the Technical Specifications or has been verified by other independent methods such as indications displayed on the control panels, reports from plant personnel, or radiological survey results. In a case where data from Kaman readings is being used to determine whether an EAL threshold value has been exceeded, Valid means that flow through the associated Kaman Monitor has been verified and does exist as indicated in pCi/sec on SPRAD.

The preferred method for declaration of RG1 is by means of Dose Assessment using the MIDAS computer model. If dose assessment results are available at the time of declaration, the classification should be based on RG1.1. However, if Kaman monitor readings are sustained for 15 minutes or longer and the required MIDAS dose assessments cannot be completed within this period, then the declaration can be made using Kaman readings PROVIDED the readings are not from an isolated flow path. If Kaman readings are not valid, field survey results may be utilized.

DAEC's Meteorological Information and Dose Assessment System .(MIDAS) was utilized to determine the Kaman monitor limits. Eight separate combinations of release point, source term, meteorological conditions and equipment status were analyzed. Pathways considered were the offgas stack, the turbine building exhaust vent and a single reactor building exhaust vent. Multiple release points were not considered. In this same vein, it was assumed that only one of the three reactor building vents is on during the release.

The source terms used have been pre-loaded into MIDAS and are the default mixes associated with a loss of coolant accident (LOCA) and a control rod drop (CRD). The LOCA mix was used in conjunction with a release via the offgas stack while the CRD mix was used for releases via the turbine or reactor building vents.

The source term for a release via the offgas stack is further impacted by the status of the standby gas treatment system. The status of that system was also taken into consideration.

Based on 1995 data (NG-96-0987), the atmospheric stability was classified as Pascal E 33% of the time. Consequently, both classifications were evaluated.

Based on the same report, the most common wind speeds were:

RG1

EAL BASES., DOCUMENT EBD-R Rev. 11 next ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT PAGE 27 of 28 Pascal Class Altitude Speed (mph)

D 156 feet 8-12 D 33 feet 8-12 E 156 feet 8-12 E 33 feet 4-7 Though the temperature setting has no impact on the MIDAS calculations, a value must be entered in order for the program to run. Consequently, the temperature was arbitrarily set at 50 F.

The rain estimate was set at zero, to eliminate any on site washout of radioactive material.

For the first MIDAS runs a 1Ci/cc concentration was assumed. The results of these runs were then normalized to the limits, thus generating a theoretical Kaman limit.

Additional MIDAS runs were made with these theoretical limits as input to verify the normalization process.

In addition to the total integrated dose, MIDAS calculates a peak whole body DDE rate resulting from the plume and a peak thyroid CDE rate resulting from inhalation.

Because the RS1 and RG1 Kaman limits are to be based on a one-hour exposure, establishing concentration limits so these peak values match the NUMARC limits is acceptable.

Initiating Condition - '.Initiating

.... C.ndition Site

... *- Area"

... Emergency General Emergency RS1 RG1 Valid Turbine or Reactor Building ventilation rad monitor (KAMAN) reading, 0.06 jiCi/cc 0.6 pCi/cc for 15 minutes or longer, above:

Valid Offgas Stack ventilation rad monitor (Kaman) reading, for 15 minutes or 40 ýtCi/cc 400 pCi/cc longer, above: I II DAEC does not have a telemetered radiation monitoring system.

Dose assessment using MIDAS is based on the EPA-400 methodology, e.g., use of Total Effective Dose Equivalent (TEDE) and Committed Dose Equivalent (CDE)

Thyroid. TEDE is somewhat different from whole body dose from gaseous effluents determined by ODAM methodology which forms the basis for the radiation monitor readings calculated in RU1. These factors can introduce differences that are at least as large as those introduced by using TEDE versus whole body dose.

RG1

EAL BASES DOCUMENT EBD-R Rev. 11 next ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT PAGE 28 of 28 The gaseous effluent radiation monitors can only detect noble gases. The contribution of iodines to TEDE and CDE Thyroid could therefore only be determined either by: (1) utilizing the source term mixture in MIDAS, or (2) gaseous effluent sampling. Therefore, DAEC EAL Threshold Value 1 is written in terms of TEDE and CDE Thyroid.

REFERENCES:

1. Offsite Dose Assessment Manual, Section 6.1.2 and 7.1.2, Bases

2. Emergency Plan Implementing Procedure (EPIP) 3.3, Dose Assessment and Protective Action

3. Radiation Protection Calculation No. 95-001-C, Emergency Actions Levels Based on Effluent Radiation Monitors, January 24, 1995

4. Radiation Engineering Calculation No. 96-007-A, Determination of DAEC Radioactive Release Initiating Conditions for AS1 & AG1 Emergency Classifications, July 3, 1996

5. UFSAR Section 11.5, Process and Effluent Radiation Monitoring and Sampling Systems

6. EPA 400-R-92-001, Manual of Protective Action Guides and ProtectiveActions for Nuclear Incidents

7. NEI Methodology for Development of Emergency Action Levels NEI 99-01 Revision 5, February 2008 RG1

EAL BASES DOCUMENT EBD-REF Rev. 2 next EAL SUPPORTING REFERENCE INFORMATION Page 1 of 19 II Usage Level Information Use Effective Date:

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Record the following: Date / Time: Initials:

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EAL BASES DOCUMENT EBD-REF Rev. 2 next INFORMATION EAL SUPPORTING REFERENCE Page 2 of 19 Table of Contents Section Page EXECUTIVE SUM MARY ..................................................................................................... 3 ACRO NYM S ............................................................................................................................... 3 1.0 PURPO SE ............................................................................................................................ 3

2.0 REFERENCES

..................................................................................................................... 3 3.0 DISC USSIO N ....................................................................................................................... 3 4.0 TECHNICAL BASES INFO RMATIO N .............................................................................. 3 5.0 DEFINITIO NS ....................................................................................................................... 3

EAL BASES DOCUMENT EBD-REF Rev. 2 next INFORMATION EAL SUPPORTING REFERENCE Page 3 of 19 EXECUTIVE

SUMMARY

DAEC implemented NEI EALs via NESP-007 in January 1998. A subsequent revision to the DAEC EAL scheme occurred in January 2006 based upon NEI 99-01 Revision 4. Since that time, NEI has developed Revision 5 to the NEI 99-01 document and DAEC has again upgraded its EAL scheme as a result of the revision.

Revision 5 recognized implementation difficulties, interpretations and errors of Revision 4 and was developed through use of a Frequently Asked Questions (FAQ) format where stakeholders submitted concerns to the NEI Task Force and technical solutions were found to better transition the classification process. The vast majority of the EALs are functionally similar to what was already in place at DAEC.

However, many EALs were revised to better align with the intent and wording of Revision 5. It is important to note that DAEC does not deviate from the intent of Revision 5 guidance, however the following conditions should be noted:

• DAEC is a BWR and therefore cannot implement PWR specific guidance.

• DAEC does not have a perimeter rad monitoring system or an automatic dose assessment system and therefore cannot implement guidance for these systems.

On July 18, 2005 the NRC issued Bulletin 2005-02. This document provided new security-related EALs based on evaluation and review of EP planning basis and licensee performance during security-based EP drills. The NEI developed an industry white paper, "Enhancements to Emergency Preparedness Programs for Hostile Action", that provided an acceptable method of implementing requirements in Bulletin 2005-02. The NRC endorsement of the methods in the white paper is documented in the letter from N. Mamish to A. Nelson dated December 8, 2005. The DAEC EALS and Bases were revised in accordance with Bulletin 2005-02 and the methods outlined in the NEI white paper.

Using NEI 99-01 Revision 5, DAEC implemented new security-based EALs in March 2010 in support of implementation of the new Security Contingency Plan contained in NEI 03-12, Revision 6. The remaining EALS for NEI 99-01 Revision 5 will be implemented at a later date.

-EAL BASES DOCUMENT- EBD-REF Rev. 2 next EAL SUPPORTING REFERENCE INFORMATION Page 4 of 19 ACRONYMS AC Alternating Current ATWS Anticipated Transient Without Scram BWR Boiling Water Reactor CCW Component Cooling Water CDE Committed Dose Equivalent CFR Code of Federal Regulations CMT Containment DC Direct Current DHR Decay Heat Removal DOT Department of Transportation EAL Emergency Action Level ECCS Emergency Core Cooling System ECL Emergency Classification Level EOF Emergency Operations Facility EOP Emergency Operating Procedure EPA Environmental Protection Agency EPG Emergency Procedure Guideline EPIP Emergency Plan Implementing Procedure EPRI Electric Power Research Institute ESF Engineered Safeguards Feature ESW Emergency Service Water GE General Emergency HPCI High Pressure Coolant Injection IC Initiating Condition IDLH Immediately Dangerous to Life and Health IPEEE Individual Plant Examination of External Events (Generic Letter 88-20)

ISFSI Independent Spent Fuel Storage Installation LCO Limiting Condition of Operation LER Licensee Event Report LFL Lower Flammability Limit LOCA Loss of Coolant Accident MSIV Main Steam Isolation Valve

EAL BASES DOCUMENT! EBD-REF Rev. 2 next EAL SUPPORTING REFERENCE INFORMATION Page 5 of 19 Mw Megawatt NEI Nuclear Energy Institute NESP National Environmental Studies Project NOUE Notification Of Unusual Event NRC Nuclear Regulatory Commission NSSS Nuclear Steam Supply System NUMARC Nuclear Management and Resources Council OBE Operating Basis Earthquake ODCM Offsite Dose Calculation Manual PRA/PSA Probabilistic Risk Assessment / Probabilistic Safety Assessment PSIG Pounds per Square Inch Gauge RCIC Reactor Core Isolation Cooling RCS Reactor Coolant System RPS Reactor Protection System RPV Reactor Pressure Vessel SAE Site Area Emergency SBGT Stand-By Gas Treatment SPDS Safety Parameter Display System SRO Senior Reactor Operator SSE Safe Shutdown Earthquake TEDE Total Effective Dose Equivalent TOF Top of Active Fuel TSC Technical Support Center USAR Updated Safety Analysis Report

EAL BASES DOCUMENT EBD-REF Rev. 2 next EAL SUPPORTING REFERENCE INFORMATION Page 6 of 19 1.0 PURPOSE This document provides the detailed set of Emergency Action Levels (EALs) applicable to the Duane Arnold Energy Center (DAEC) and the associated Technical Bases using the EAL development methodology found in NEI 99-01 Revision 5 [Ref. 2.1]. Personnel responsible for the classification of emergencies may use this document as a technical reference and an aid in EAL interpretation.

The primary tool for determining the emergency classification level is the Emergency Action Level Matrix. The user of the Emergency Action Level Matrix may (but is not required) to consult the EAL Technical Basis Document in order to obtain additional information concerning the EALs under classification consideration.

2.0 REFERENCES

2.1 NEI 99-01 Revision 5, Methodology for Development of Emergency Action Levels 2.2 Emergency Action Level Matrix 2.3 DAEC Technical Specifications 2.4 DAEC Emergency Plan & Implementing Procedures 2.5 NEI 03-12, Rev. 6, "Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan; [and Independent Spent Fuel Storage Installation Security Program]"

EAL'BASES'DOCUMENT EBD-REF Rev. 2 next EAL SUPPORTING REFERENCE INFORMATION Page 7 of 19 3.0 DISCUSSION 3.1 Background EALs are the plant-specific indications, conditions or instrument readings that are utilized to classify emergency conditions defined in the DAEC 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.

Using NEI 99-01 Rev. 4, DAEC conducted an EAL implementation upgrade project that produced the EALs discussed herein. While the upgraded EALs are site-specific, an objective of the project was to ensure to the extent possible EAL conformity and consistency between the NMC plant sites.

Using NEI 99-01 Revision 5, DAEC revised the EAL scheme to be in accordance with the most current industry approved EAL guidance document.

NEI 99-01 (NUMARC/NESP-007) Revision 5 represents the most recently NRC endorsed methodology per RG 1.101 Revision 4, "Emergency Planning and Preparedness for Nuclear Power Reactors."

Revision 5 enhancements over earlier revisions included:

• Security EALs with the Hostile Action changes endorsed by the NRC in RIS 2006-12 on July 19, 2006;

• Enhanced guidance related to Security EALs to ensure consistency with NEI 03-12;

" Clarification of several EALs to resolve any potential misunderstanding as to the intent of the EAL.

" Words that could be confused with similar sounding words were replaced in EALs, e.g., "rise and drop" replaced "increase and decrease."

• Mathematical symbols were replaced with text, e.g., "greater than or equal to" replaced ">".

• Information previously contained in the EAL Basis section that could reasonably alter how, or when, an EAL is declared have been incorporated into the example EALs.

EAL BASESDOCUMENT EBD-REF Rev. 2 next EAL SUPPORTING REFERENCE INFORMATION Page 8 of 19 3.2 Key Definitions in EAL Methodology The following definitions apply to the generic EAL methodology:

EMERGENCY CLASSIFICATION LEVEL: One of a minimum set of names or titles, established by the Nuclear Regulatory Commission (NRC), for grouping of normal nuclear power plant conditions according to (1) their relative radiological seriousness, and (2) the time sensitive onsite and off site radiological emergency preparedness actions necessary to respond to such conditions. The existing radiological emergency classes, in ascending order of seriousness, are called:

• Notification of Unusual Event (NOUE or UE)

" Alert

• Site Area Emergency (SAE)

" General Emergency (GE)

Section 3.4 provides further discussion of the emergency classes.

INITIATING CONDITION (IC): One of a predetermined subset of nuclear power plant conditions when either the potential exists for a radiological emergency, or such an emergency has occurred.

" An IC is an emergency condition which sets it apart from the broad class of conditions that may or may not have the potential to escalate into a radiological emergency.

• It can be a continuous, measurable function that is outside technical specifications, such as elevated RCS temperature or falling reactor coolant level (a symptom).

• It also encompasses occurrences such as FIRE (an event) or reactor coolant pipe failure (an event or a barrier breach).

EMERGENCY ACTION LEVEL (EAL): A pre determined, site-specific, observable threshold for a plant Initiating Condition that places the plant in a given emergency classification level. An EAL can be: an instrument reading; an equipment status indicator; a measurable parameter (onsite or offsite); a discrete, observable event; results of analyses; entry into specific emergency operating procedures; or another phenomenon which, if it occurs, indicates entry into a particular emergency classification level.

" There are times when an EAL will be a threshold point on a measurable continuous function, such as a primary system coolant leak that has exceeded technical specifications.

• At other times, the EAL and the IC will coincide, both identified by a discrete event that places the plant in a particular emergency classification level.

• EAL BASESDOCUMENT. EBD-REF Rev. 2 next INFORMATION EAL SUPPORTING REFERENCE Page 9 of 19 3.3 Recognition Categories ICs and EALs are grouped in one of several categories. This classification scheme incorporates symptom-based, event-based, and barrier-based ICs and EALs.

• R - Abnormal Rad Levels/Radiological Effluent

• C - Cold Shutdown / Refueling System Malfunction

• E - Independent Spent Fuel Storage Installation (ISFSI)

• F - Fission Product Barrier Degradation

• H - Hazards

• S - System Malfunction Some recognition categories are further divided into one or more subcategories depending on the types and number of plant conditions that dictate emergency classifications. An EAL may or may not exist for each subcategory at all four classification levels. Similarly, more than one EAL may exist for a subcategory in a given emergency classification when appropriate (i.e., no EAL at the General Emergency level but three EALs at the Unusual Event level).

3.4 Emergency Class Descriptions There are three considerations related to the emergency classification levels. These are:

• The potential impact on radiological safety, either as now known or as can be reasonably projected.

• How far the plant is beyond its predefined design, safety and operating envelopes.

• Whether or not conditions that threaten health are expected to be confined to within the site boundary.

The ICs deal explicitly with radiological safety affect by escalating from levels corresponding to releases within regulatory limits to releases beyond EPA Protective Action Guideline (PAG) plume exposure levels.

NOTIFICATION OF UNUSUAL EVENT: 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.

" Potential degradation of the level of safety of the plant is indicated primarily by exceeding plant technical specification Limiting Condition of Operation (LCO) allowable action statement time for achieving required mode change.

• Precursors of more serious events should also be included because precursors represent a potential degradation in the level of safety of the plant.

" Minor releases of radioactive materials are included. In this emergency classification level, however, releases do not require monitoring or offsite response.

EAL BASES DOCUMENT EBD-REF Rev. 2 next EAL SUPPORTING REFERENCE INFORMATION Page 10 of 19 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.

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 Protective Action Guideline (PAG) exposure levels beyond the site boundary.

• The discriminator (threshold) between Site Area Emergency and General Emergency is whether or not the EPA PAG plume exposure levels are expected to be exceeded outside the site boundary.

• This threshold, in addition to dynamic dose assessment considerations discussed in the EAL guidelines, clearly addresses NRC and offsite emergency response agency concerns as to timely declaration of a General Emergency.

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.

" The bottom line for the General Emergency is whether evacuation or sheltering of the general public is indicated based on EPA PAGs and, therefore, should be interpreted to include radionuclide release regardless of cause.

" To better assure timely notification, EALs in this category are primarily expressed in terms of plant function status, with secondary reliance on dose projection. In terms of fission product barriers, loss of two barriers with loss or potential loss of the third barrier constitutes a General Emergency.

EAL BASES DOCUMENT"4< EBD-REF Rev. 2 next EAL SUPPORTING REFERENCE INFORMATION Page 11 of 19 3.5 Operating Mode Applicability The following definitions of operating modes are used in this document:

1 Power Operation Reactor mode switch in "Run."

2 Startup Reactor mode switch in "Start to Hot Stby" or "Refuel" (with all reactor vessel head closure bolts fully tensioned).

3 Hot Shutdown Reactor mode switch in "Shutdown" with average reactor coolant temperature GREATER THAN 212°F.

4 Cold Shutdown Reactor mode switch in "Shutdown" with average reactor coolant temperature LESS THAN OR EQUAL TO 212°F.

5 Refuelinq Reactor Mode Switch in "Shutdown" or "Refuel" with one or more reactor vessel head closure bolts less than fully tensioned.

In addition to these operating modes, NEI 99-01 [Ref. 1] defines the following additional mode:

D Defueled All reactor fuel removed from Reactor Vessel (full core off load during refueling or extended outage)

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

Recognition categories are associated with the operating modes listed in the following matrix:

Recognition Category Mode R C E F H S 1 - Power Operation X X X X 2 - Startup X X X X 3 - Hot Shutdown X X X X 4 - Cold Shutdown X X X 5 - Refueling X X X

EAL BASES DOCUMENT EBD-REF Rev. 2 next EAL SUPPORTING REFERENCE INFORMATION Page 12 of 19 3.6 Fission Product Barriers Many of the EALs derived from the NEI methodology are fission product barrier based. That is, the conditions that define the EALs are based upon loss of or potential loss to one or more of the three fission product barriers. "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 and "potential loss" means IMMINENT loss of the barrier.

The primary fission product barriers are:

• Fuel Cladding (FC): Zirconium tubes which house the ceramic uranium oxide pellets along with the end plugs which are welded into each end of the fuel rods comprise the FC barrier.

" Reactor Coolant System (RCS): The reactor vessel shell, vessel head, vessel nozzles and penetrations and all primary systems directly connected to the reactor vessel up to the first containment isolation valve comprise the RCS barrier.

• Containment (CMT): The primary containment structure and all isolation valves required to maintain primary containment integrity under accident conditions comprise the CMT barrier.

3.7 Emergency Classification Based on Fission Product Barrier Degradation The following criteria are the bases for event classification related to fission product barrier loss or challenge:

" Notification of Unusual Event:

Any loss or any potential loss of Containment

• Alert:

Any loss or any potential loss of either Fuel Cladding or RCS

• Site Area Emergency:

Loss or potential loss of any two barriers

• General Emergency:

Loss of any two barriers and loss or potential loss of third barrier

EAL BASES DOCUMENT EBD-REF Rev. 2 next INFORMATION EAL SUPPORTING REFERENCE Page 13 of 19 3.8 EAL Relationship to EOPs Where possible, the EALs have been made consistent with and utilize the conditions defined in the DAEC Emergency Operating Procedures (EOPs). While the symptoms that drive operator actions specified in the EOPs are not indicative of all possible conditions which warrant emergency classification, they do define the symptoms, independent of initiating events, for which reactor plant safety and/or fission product barrier integrity are threatened. Where these symptoms are clearly representative of one of the NEI Initiating Conditions, they have been utilized as an EAL. This permits rapid classification of emergency situations based on plant conditions without the need for additional evaluation or event diagnosis. Although some of the EALs presented here are based on conditions defined in the EOPs, classification of emergencies using these EALs is not dependent upon EOP entry or execution. The EALs can be utilized independently or in conjunction with the EOPs.

3.9 Symptom Based vs. Event Based Approach To the extent possible, the EALs are symptom based. That is, the action level is defined by values of key plant operating parameters that identify emergency or potential emergency conditions. This approach is appropriate because it allows the full scope of variations in the types of events to be classified as emergencies. But, a purely symptom based approach is not sufficient to address all events for which emergency classification is appropriate. Particular events to which no predetermined symptoms can be ascribed have also been utilized as EALs since they may be indicative of potentially more serious conditions not yet fully realized.

Category R - Abnormal Rad Levels/Radiological Effluent and Category F - Fission Product Barrier Degradation are primarily symptom-based. The symptoms are indicative of actual or potential degradation of either fission product barriers or personnel safety.

Other categories tend to be event-based. For example, System Malfunctions are abnormal and emergency events associated with vital plant system failures, while Hazards are those non-plant system related events that have affected or may affect plant safety.

3.10 Treatment of Emergency Class Upgrading When multiple simultaneous events occur, the emergency classification level is based on the highest EAL reached. For example, two Alerts remain in the Alert category, or an Alert and a Site Area Emergency is a Site Area Emergency.

3.11 Treatment of Emergency Class Downgrading Another important aspect of usable EAL guidance is the consideration of what to do when the risk posed by an emergency is clearly decreasing. A combination approach involving recovery from General Emergencies and Site Area Emergencies and termination from Unusual Events and Alerts is used.

Downgrading to lower emergency classes adds notifications but may have merit under certain circumstances.

EAL BASES, DOCUMENT EBD-REF Rev. 2 next EAL SUPPORTING REFERENCE INFORMATION Page 14 of 19 3.12 Classifying Transient Events For some events, the condition may be corrected before a declaration has been made. For example, an emergency classification is warranted when automatic and manual actions taken within the control room do not result in a required reactor scram. However, it is likely that actions taken outside of the control room will be successful, probably before the Emergency Director classifies the event. The key consideration in this situation is to determine whether or not further plant damage occurred while the corrective actions were being taken. In some situations, this can be readily determined, in other situations, further analyses (e.g., coolant sampling) may be necessary.

In general, observe the following guidance: Classify the event as indicated and terminate the emergency once assessment shows that there were no consequences from the event and other termination criteria are met. For example, a momentary event, such as an ATWS or an earthquake, requires declaration even though the condition may have been resolved by the time the declaration is made.

• An ATWS represents a failure of a front line safety system (RPS) designed to protect the health and safety of the public.

• The affect of an earthquake on plant equipment and structures may not be readily apparent until investigations are conducted.

There may be cases in which a plant condition that exceeded an EAL threshold was not recognized at the time of occurrence, but is identified well after the condition has occurred (e.g., as a result of routine log or record review) and the condition no longer exists. In these cases, an emergency should not be declared. Reporting requirements of 10 CFR 50.72 are applicable and the guidance of NUREG-1 022, Event Reporting Guidelines 10 CFR 50.72 and 50.73, should be applied.

3.13 IMMINENT EAL Thresholds Although the majority of the EALs provide very specific thresholds, the Emergency Director must remain alert to events or conditions that lead to the conclusion that exceeding the EAL threshold is IMMINENT.

If, in the judgment of the Emergency Director, an IMMINENT situation is at hand, the classification should be made as if the threshold has been exceeded. While this is particularly prudent at the higher emergency classes (as the early classification may provide for more effective implementation of protective measures), it is nonetheless applicable to all emergency classes. Explicit EALs, specifying use of Emergency Director judgment, are given in the Hazards, ISFSI and Fission Product Barrier Degradation categories.

EA'L BASES DOCUMENT EBD-REF Rev. 2 next EAL SUPPORTING REFERENCE INFORMATION Page 15 of 19 4.0 TECHNICAL BASES INFORMATION 4.1 Recognition Category Organization The technical bases of the EALs are provided under Recognition Categories R, C, E, F, H and S of this document. A table summarizing the Initiating Conditions introduces each category. The tables provide an overview of how the ICs are related under each emergency class. ICs within each category are listed according to classification (as applicable) in the following order: Notification of Unusual Event, Alert, Site Area Emergency, and General Emergency.

The basis information for the fission barrier table indicators is organized similarly to the other basis information described above. For each barrier - fuel clad, RCS, and primary containment - basis information is organized by "Indicator." The indicator is the name for the row on the fission barrier table and is used for convenient grouping of similar symptoms, similar to the "Event Type" used for the R, C, E, H, and S EALs described above. Indicators include Radiation/Core Damage, RPV Level, Leakage, Primary Containment Atmosphere, and Emergency Director Judgment.

After the DAEC Indicator, the applicable generic BWR fission product barrier indicators are then displayed, showing both the generic loss and potential loss conditions, as applicable. Next displayed is the appropriate DAEC information and references. These are displayed in the same manner as the R, C, E, H, and S recognition category basis information described above.

EAL BASES DOCUMENT, EBD-REF Rev. 2 next EAL SUPPORTING REFERENCE INFORMATION Page 16 of 19 4.2 Initiating Condition Structure ICs in Recognition Categories R, C, E, H and S are structured in the following manner:

• Recognition Category Title

• IC Identifier:

o First character identifies the category by letter (R, C, E, H and S) o Second character identifies the emergency classification level (U for Notification of Unusual Event, A for Alert, S for Site Area Emergency, and G for General Emergency) o Third character is the numerical sequence as given in Revision 5 of NEI 99-01 [Ref. 1]

(e.g., SA2). Due to document revisions, certain NEI ICs have been deleted, leaving gaps in the numerical sequence.

• Emergency Class: Notification of Unusual Event, Alert, Site Area Emergency, or General Emergency

" IC Description

• Operating Mode Applicability: Refers to the operating mode during which the IC/EAL is applicable

• Emergency Action Level(s): EALs are the conditions applicable to the criteria of the IC and are used to determine the need to classify an event/condition. If more than one EAL is applicable to an IC, emergency classification is required when any EAL within the IC reaches the EAL threshold. To clarify this intent, ICs with multiple EALs include a parenthetical phrase in the EAL title line, indicating that each constitutes an emergency classification. For example, the phrase

"(RAI. 1 or RA1.2)" indicates that either EAL is a Notification of Unusual Event.

• Basis: Provides information that explains the IC and EAL(s). Plant source document references are provided as needed to substantiate site-specific information included in the EALs and bases.

4.3 EAL Identification The EAL identifier is the IC identifier followed by a period and sequence number (e.g., RU1.1, RU1.2, etc.). If only one EAL is assigned to an IC, the EAL is given the number one.

The primary purpose of the EAL identifier is to uniquely distinguish each classifiable condition.

Secondary purposes are to assist location of an EAL within the EAL classification scheme and to announce the emergency classification level.

EAL BASES DOCUMENT EBD-REF Rev. 2 next EAL SUPPORTING REFERENCE INFORMATION Page 17 of 19 5.0 DEFINITIONS In the ICs and EALs, selected words are in uppercase print. These words are defined terms. Definitions are provided below.

AFFECTING SAFE SHUTDOWN: Event in progress has adversely affected functions that are necessary to bring the plant to and maintain it in the applicable HOT or COLD SHUTDOWN condition. Plant condition applicability is determined by Technical Specification LCOs in effect.

Example 1: Event causes damage that results in entry into an LCO that requires the plant to be placed in HOT SHUTDOWN. HOT SHUTDOWN is achievable, but COLD SHUTDOWN is not.

This event is not "AFFECTING SAFE SHUTDOWN."

Example 2: Event causes damage that results in entry into an LCO that requires the plant to be placed in COLD SHUTDOWN. HOT SHUTDOWN is achievable, but COLD SHUTDOWN is not.

This event is "AFFECTING SAFE SHUTDOWN."

BOMB: refers to an explosive device suspected of having sufficient force to damage plant systems or structures.

CIVIL DISTURBANCE: is a group of persons violently protesting station operations or activities at the site.

COLD SHUTDOWN: As defined in Technical Specification Table 1.1-1, the reactor is in the shutdown mode, the reactor coolant temperature is less than or equal to 212 0F, and all reactor vessel head closure bolts fully tensioned.

COMPENSATORY NON-ALARMING INDICATIONS: Information displayed in the main control room including analog and digital parameter displays, trend recorders, the Safety Parameter Display System (SPDS), and the plant process computer.

CONFINEMENT BOUNDARY: is the barrier(s) between areas containing radioactive substances and the environment.

CONTAINMENT CLOSURE: Site specific procedurally defined actions taken to secure containment and its associated structures, systems, and components as a functional barrier to fission product release under existing plant conditions. For BWRs, this is considered to be Secondary Containment as required by Technical Specifications.

EMERGENCY DIRECTOR: Individual responsible for overall direction and control of the Emergency Response Organization (ERO). In the Control Room this would be the Operations Shift Manager (OSM).

In the Technical Support Center (TSC) this would be the Emergency Coordinator (EC). In the Emergency Operations Facility (EOF) this would be the Emergency Response and Recovery Director (ER&RD).

EXPLOSION: is a rapid, violent, unconfined combustion, or catastrophic failure of pressurized/energized equipment that imparts energy of sufficient force to potentially damage permanent structures, systems, or components.

EXTORTION: is an attempt to cause an action at the station by threat of force.

FIRE: is 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.

EAL BASES DOCUMENT EBD-REF Rev. 2 next EAL SUPPORTING REFERENCE INFORMATION Page 18 of 19 HOSTAGE: is a person(s) held as leverage against the station to ensure that demands will be met by the station.

HOSTILE ACTION: An.act toward an NPP or its personnel that includes the use of violent force to destroy equipment, take hostages, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILES, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the NPP. Non-terrorism-based EALs should be used to address such activities, (e.g., this may include violent acts between individuals in the owner controlled area).

HOSTILE FORCE: one or more individuals who are engaged in a determined assault, overtly or by stealth and deception, equipped with suitable weapons capable of killing, maiming, or causing destruction.

HOT SHUTDOWN: As defined in Technical Specifications Table 1.1-1, the reactor mode switch is in the shutdown position and the reactor coolant temperature is greater than 2121F and all reactor head closure bolts fully tensioned.

IMMINENT: Mitigation actions have been ineffective, additional actions are not expected to be successful, and trended information indicates that the event or condition will occur. Where IMMINENT time frames are specified, they shall apply.

INTRUSION: is a person(s) present in a specified area without authorization. Discovery of a BOMB in a specified area is indication of INTRUSION into that area by a HOSTILE FORCE.

INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI): The on site facility where the loaded Dry Shielded Canisters (DSCs) will be stored in Horizontal Storage Modules (HSMs). The installation is intended for interim storage until the spent fuel is removed from the plant site.

NORMAL PLANT OPERATIONS: activities at the plant site associated with routine testing, maintenance, or equipment operations, in accordance with normal operating or administrative procedures. Entry into abnormal or emergency operating procedures, or deviation from normal security or radiological controls posture, is a departure from NORMAL PLANT OPERATIONS.

PROJECTILE: is an object directed toward a NPP that could cause concern for its continued operability, reliability or personnel safety.

PROTECTED AREA: is typically the site-specific area which normally encompasses all controlled areas within the security PROTECTED AREA fence.

SABOTAGE: is deliberate damage, mis-alignment, or mis-operation of plant equipment with the intent to render the equipment inoperable. Equipment found tampered with or damaged due to malicious mischief may NOT meet the definition of SABOTAGE until this determination is made by security supervision.

SAFE SHUTDOWN AREA: Any area containing equipment, systems, or components that are necessary to bring the plant to, and maintain it in a shutdown condition. In the EAL Bases Documents and Tables, Safe Shutdown Area is synonymous with Vital Area.

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.

SIGNIFICANT TRANSIENT: is an UNPLANNED event involving one or more of the following: (1) automatic turbine runback greater than 25% thermal reactor power, (2) electrical load rejection greater

EAL BASES DOCUMENT EBD-REF Rev. 2 next EAL SUPPORTING REFERENCE INFORMATION Page 19 of 19 than 25% full electrical load, (3) Reactor Trip, (4) Safety Injection Activation, or (5) thermal power oscillations greater than 10%

STRIKE ACTION: is a work stoppage within the PROTECTED AREA by a body of workers to enforce compliance with demands. The STRIKE ACTION must threaten to interrupt NORMAL PLANT OPERATIONS.

UNISOLABLE: a breach or leak that cannot be promptly isolated.

UNPLANNED: a parameter change or an event that is not the result of an intended evolution and requires corrective or mitigative actions.

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

VISIBLE DAMAGE: is damage to equipment or structure that is readily observable without measurements, testing, or analysis. Damage is sufficient to cause concern regarding the continued operability or reliability of affected safety structure, system, or component. Example damage includes:

deformation due to heat or impact, denting, penetration, rupture, cracking, paint blistering. Surface blemishes (e.g., paint chipping, scratches) should not be included.

VITAL AREA: is any area, normally within the PLANT PROTECTED AREA, which contains equipment, systems, components, or material, the failure, destruction, or release of which could directly or indirectly endanger the public health and safety by exposure to radiation. In the EAL Bases Documents and Tables, Safe Shutdown Area is synonymous with Vital Area. NOTE: Vital areas for EAL purposes are not the same as vital areas for Security purposes.

EAL BASES DOCUMENT Rev. 1 next REGULATORY CONTEXT Page 1 of 3 Effective Date:

Title 10, Code of Federal Regulations, Part 50 provides the regulations that govern emergency preparedness at nuclear power plants. Nuclear power reactor licensees are required to have NRC-approved "emergency response plans" for dealing with "radiological emergencies." The requirements call for both onsite and offsite emergency response plans, with the offsite plans being those approved by FEMA and used by the State and local authorities. This document deals with the utilities' approved onsite plans and procedures for response to radiological emergencies at nuclear power plants, and the links they provide to the offsite plans.

Section 50.47 of Title 10 of the Code of Federal Regulations (10 CFR 50.47),

entitled "Emergency Plans," states the requirement for such plans. Part (a)(1) of this regulation states that "no operating license will be issued unless a finding is made by NRC that there is reasonable assurance that adequate protective measures can and will be taken in the event of a radiological emergency."

The major portion of 10 CFR 50.47 lists "standards" that emergency response plans must meet. The standards constitute a detailed list of items to be addressed in the plans. Of particular importance to this project is the fourth standard, which addresses "emergency classification" and "action levels." These terms, however, are not defined in the regulation.

10 CFR 50.54, "Conditions of licenses," emphasizes that power reactor licensees must "follow, and maintain in effect, emergency plans which meet the standards in Part 50.47(b) and the requirements in Appendix E to this part-" The remainder of this part deals primarily with required implementation dates.

10 CFR 50.54(q) allows licensees to make changes to emergency plans without prior Commission approval only if: (a) the changes do not decrease the effectiveness of the plans and (b) the plans, as changed, continue to meet 10 CFR 50.47(b) standards and 10 CFR 50 Appendix E requirements. The licensee must keep a record of any such changes. Proposed changes that decrease the effectiveness of the approved emergency plans may not be implemented without application to and approval by the Commission.

10 CFR 50.72 deals with "Immediate notification requirements for operating nuclear power reactors." The "immediate" notification section actually includes three types of reports: (1) immediately after notification of State or local agencies

(for emergency classification events); (2) one-hour reports; and, (3) four-hour reports.

Although 10 CFR 50.72 contains significant detail, it does not define either "Emergency Class" or "Emergency Action Level." But one-hour and four-hour reports are listed as "non-emergency events," namely, those which are "not reported as a declaration of an Emergency Class." Certain 10CFR50.72 events can also meet the Notification of Unusual Event emergency classification if they are precursors of more serious events. These situations also warrant anticipatory notification of state and local officials.

By footnote, the reader is directed from 10 CFR 50.72 to 10 CFR 50 Appendix E, for information concerning "Emergency Classes."

10 CFR 50.73 describes the "Licensee event report system," which requires submittal of follow-up written reports within thirty days of required notification of NRC.

10 CFR 50 Appendix E, Section B, "Assessment Actions," mandates that emergency plans must contain "emergency action levels." EALs are to be described for: (1) determining the need for notification and participation of various agencies, and (2) determining when and what type of protective measures should be considered. Appendix E continues by stating that the EALs are to be based on: (1) in-plant conditions; (2) in-plant instrumentation; (3) onsite monitoring; and (4) offsite monitoring.

10 CFR 50 Appendix E, Section C, "Activation of Emergency Organization," also addresses "emergency classes" and "emergency action levels." This section states that EALs are to be based on: (1) onsite radiation monitoring information; (2) offsite radiation monitoring information; and, (3) readings from a number of plant sensors that indicate a potential emergency, such as containment pressure and the response of the Emergency Core Cooling System. This section also states that "emergency classes" shall include: (1) Notification of Unusual Events (NOUEs), (2) Alert, (3) Site Area Emergency, and (4) General Emergency.

These regulations are supplemented by various regulatory guidance documents.

A significant document that has dealt specifically with EALs is NUREG-0654/FEMA-REP-1, "Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants," October 1980.

Recognition Category E (Events Related to ISFSI) is applicable to licensees using their 10 CFR 50 emergency plan to fulfill the requirements of 10 CFR 72.32. Recognition Category E is not applicable to stand alone ISFSls, Monitored

EAL BASES DOCUMENT Rev. 1 next REGULATORY CONTEXT Page 3 of 3 Retrievable Storage Facilities (MRS), or ISFSIs that may process and/or repackage spent fuel. The emergency classifications for Recognition Category E are those provided by NUREG 0654/FEMA Rep.1 in accordance with 10 CFR 50.47. The classification of an ISFSI event under provisions of a 10 CFR 50.47 emergency plan should be consistent with the definitions of the emergency classes as used by that plan. A site-specific analysis would make this determination, but in most cases it is expected that classification of an NOUE would be appropriate. It is expected that the initiating conditions germane to a 10 CFR 72.32 emergency plan (described in NUREG-1567) are subsumed within 10 CFR 50.47 emergency plan's classification scheme.

EAL BASES DOCUMENT <7. EBD-S Rev. 8 next SYSTEM MALFUNCTION CATEGORY Page 1 of 33 EfceUsage Levelt Information Use Effective Date:

TECHNICAL. REVIEW  :

Prepared by: Date:

Reviewed by: Date:

Independent Reviewer I am responsible for the technical content of this procedure.

Approved by. Date:

Manager, Emergency Planning

EL BASES DOCUMENT EBD-S Rev. 8 next SYSTEM MALFUNCTION CATEGORY Page 2 of 33 SUI Loss of all offsite AC power to Essential Busses for 15 Minutes or longer EVENT TYPE: Loss of Power OPERATING MODE APPLICABILITY: Power Operation, Startup, Hot S/D EAL THRESHOLD VALUE:

Note: The Emergency Director (OSM, EC or ER&RD) 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.

SU1.1 Loss of all offsite AC power to 1A3 and 1A4 for 15 minutes or longer.

DAEC EAL INFORMATION:

This event is a precursor of a more serious Station Blackout condition and is thus considered as a potential degradation of the level of safety of the plant. It is possible to be operating within Technical Specification LCO Action Statement time limits and make a declaration of an Unusual Event in accordance with this EAL.

The intent of this EAL is to declare an UNUSUAL EVENT when offsite power has been lost and both of the emergency diesel generators have successfully started and energized their respective 4kv emergency bus.

REFERENCES:

1. Abnormal Operating Procedure (AOP) 301, Loss of Essential Electrical Power

2. UFSAR Section 8.2, Offsite Power System

3. NEI 99-01 Rev. 5, NEI Methodology for Development of Emergency Action Levels Sul

EAL BASES DOCUMENT EBD-S Rev. 8 next SYSTEM MALFUNCTION CATEGORY Page 3 of 33 SU2 Inability to reach required shutdown within Technical Specification limits EVENT TYPE: Inability to Reach or Maintain Shutdown Conditions OPERATING MODE APPLICABILITY: Power Operation, Startup, Hot S/D EAL THRESHOLD VALUE:

SU2.1 Plant is not brought to required operating mode within applicable Technical Specifications LCO Action Statement Time.

DAEC EAL INFORMATION:

Limiting Conditions for Operations (LCO) require the plant to be brought to a required operating mode when the Technical Specification required configuration cannot be restored (e.g. LCO has been entered). Depending on the circumstances this may or may not be an emergency or a precursor to a more severe condition. In any case the initiation of plant shutdown required by Technical Specifications requires a four-hour report under 10CFR50.72(b) non-emergency events. The plant is within its safety envelope when being shutdown within the allowable action statement time in Technical Specifications. An immediate classification of UNUSUAL EVENT is required when the plant is NOT brought to the required operating mode within the allowable action statement time in the Technical Specifications LCO. Declaration of the NOUE is based on the time at which the LCO Action Statement specified time period elapses and is NOT related to how long a condition may have existed.

REFERENCES:

1. DAEC Technical Specifications

2. NEI 99-01 Rev. 5, NEI Methodology for Development of Emergency Action Levels SU2

EAL BASES DOCUMENT EBD-S Rev. 8 next SYSTEM MALFUNCTION CATEGORY Page 4 of 33 SU3 UNPLANNED loss of safety system annunciation or indication in the control room for 15 minutes or longer EVENT TYPE: Instrumentation/Communication OPERATING MODE APPLICABILITY: Power Operation, Startup, Hot S/D EAL THRESHOLD VALUE:

Note: The Emergency Director (OSM, EC or ER&RD) 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.

SU3.1 UNPLANNED loss of approximately 75% or more of any of the following for 15 minutes or longer:

• 1C03, 1C04 and 1C05 annunciators

• 1C03, 1C04 and 1C05 indications 0 radiation monitor indications DAEC EAL INFORMATION:

Control room panels 1C03, 1C04, and 1C05 contain the annunciators associated with safety systems at DAEC. Therefore, the DAEC EAL addresses UNPLANNED loss of most annunciators on these panels. UNPLANNED loss of annunciators or indicators excludes scheduled maintenance and testing activities.

Under the conditions of concern, entry into AOP 302.2, Loss of Alarm Panel Power, would be made. The procedure requires alerting operators on shift to the nature of the lost annunciation. It further requires that operators be attendant and responsive to abnormal indications that relate to those systems and components that have lost annunciation.

This IC and its associated EAL are intended to recognize the difficulty associated with monitoring changing plant conditions without the use of a major portion of the annunciation or indication equipment.

SU3

EAL BASES DO CUMENT EBD-S Rev. 8 next SYSTEM MALFUNCTION CATEGORY Page 5 of 33 Recognition of the availability of computer based indication equipment is considered (e.g., SPDS, plant computer, etc.).

Quantification is arbitrary, however, it is estimated that if approximately 75% of the safety system annunciators or indicators are lost, there is an increased risk that a degraded plant condition could go undetected. It is not intended that plant personnel perform a detailed count of the instrumentation lost but use the value as a judgment threshold for determining the severity of the plant conditions.

The initiation of a Technical Specification imposed plant shutdown related to the instrument loss will be reported via 10 CFR 50.72. Ifthe shutdown is not in compliance with the Technical Specification action, the NOUE is based on SU2 "Inability to reach required shutdown within Technical Specification limits."

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

Due to the limited number of safety systems in operation during cold shutdown, refueling, and defueled modes, no IC is indicated during these modes of operation.

This NOUE will be escalated to an Alert based on a concurrent loss of compensatory indicators or if a SIGNIFICANT TRANSIENT is in progress during the loss of annunciation or indication.

UNPLANNED loss of critical safety function indicators (i.e., EOP/EAL parameters) for 15 minutes or longer may preclude operators from taking actions to mitigate a transient.

Annunciators on 1C03, 1C04, and 1C05 share a common power supply from 125 VDC Division I that is fed through circuit breaker 1D13.

SU3

EAL BASES DOCUMENT EBD-S Rev. 8 next SYSTEM MALFUNCTION CATEGORY Page 6 of 33 Indications of loss of annunciators associated with safety systems include:

• 125 VDC charger, battery, or system annunciators on control room panel 1C08

• Loss of "sealed in" annunciators at affected panels

• Failure of affected annunciator panels shiftily testing by plant operators

• Expected alarms are not received

• Computer point ID B350 indicates "NSS ANN DC LOSS TRBL." (Loss of DC power to panels 1C03, 1C04, and 1CO5)

REFERENCES:

1. Operating Instruction (01) No. 317.2 Annunciator System

2. Abnormal Operating Procedure (AOP) 302.1, Loss of 125 VDC Power

3. Abnormal Operating Procedure (AOP) 302.2, Loss of Alarm Panel Power

4. NEI 99-01 Rev. 5, NEI Methodology for Development of Emergency Action Levels SU3

EAL BASES DOCUMENT EBD-S Rev. 8 next SYSTEM MALFUNCTION CATEGORY Page 7 of 33 SU4 Fuel Clad Degradation EVENT TYPE: Fuel Clad Degradation OPERATING MODE APPLICABILITY: Power Operation, Startup, Hot S/D EAL THRESHOLD VALUE:

SU4.1 Pretreatment Offgas System (RM-4104) Hi-Hi Radiation Alarm OR SU4.2 Reactor Coolant sample activity value GREATER THAN 2.0 jiCi/gm dose equivalent 1-131.

DAEC EAL INFORMATION:

This IC is included as a NOUE because it is considered to be a potential degradation in the level of safety of the plant and a potential precursor of more serious problems.

EAL 1 - RM-4104 Hi-Hi Radiation Alarm has been chosen because it is operationally significant, is readily recognizable by the Control Room Operations Staff, and is set at a level corresponding to noble gas release rate, after 30-minute delay and decay of 1 Ci/sec. A Notification of Unusual Event is classified because a valid Offgas Pretreatment Hi-Hi radiation alarm is considered to be an indication of a potential degradation in the level of safety of the plant and a potential precursor of more serious problems.

EAL 2 - Coolant samples exceeding the short-term concentration permitted by Technical Specifications are representative of minor fuel cladding degradation. A Notification of Unusual Event is classified because Reactor Coolant Activity levels exceeding the maximum concentration in Technical Specification is considered to be a potential degradation in the level of safety of the plant and a potential precursor of more serious problems. Iodine equivalence is per Technical Specifications definition for Dose Equivalent 1-131.

Escalation of this IC to the Alert level is via the Fission Product Barrier Degradation Monitoring ICs.

SU4

EAL BASES DOCUMENT EBD-S Rev. 8 next SYSTEM MALFUNCTION CATEGORY Page 8 of 33 Though the referenced Technical Specification limits are mode dependent, it is appropriate that the EALs be applicable in all modes, as they indicate a potential degradation in the level of safety of the plant.

REFERENCES:

1. Abnormal Operating Procedure (AOP) 672.2, Offgas Radiation/Reactor Coolant High Activity

2. Technical Specification 3.4.6, Coolant Chemistry

3. Annunciator Response Procedure (ARP) 1C03A, Reactor and Containment Cooling and Isolation

4. PCP 8.6, Alarm Setpoints and Efficiency for OG Pretreatment

5. NEI 99-01 Rev. 5, NEI Methodology for Development of Emergency Action Levels SU4

EAL BASES DOCUMENT. - EBD-S Rev. 8 next SYSTEM MALFUNCTION CATEGORY Page 9 of 33 SU5 RCS Leakage EVENT TYPE: Coolant Leakage OPERATING MODE APPLICABILITY: Power Operation, Startup, Hot S/D EAL THRESHOLD VALUE:

SU5.1 Unidentified drywell leakage GREATER THAN 10 gpm.

OR SU5.2 Identified drywell leakage GREATER THAN 25 gpm.

DAEC EAL INFORMATION:

EAL Threshold Values 1 and 2 are precursors of more serious RCS barrier challenges and are thus considered as a potential degradation of the level of safety of the plant.

Thus, it is possible to be operating within Technical Specification LCO Action Statement time limits and make a declaration of an Unusual Event in accordance with these EALs.

Credit for the action statement time limit should only be given when leakage exceeds technical specification limits but has not yet exceeded the Unusual Event EAL thresholds described above.

The DAEC Tech Spec Section 3.4.4 coolant system leakage LCO limits are: (1) _<5 gpm unidentified leakage, (2) < 25 gpm total leakage averaged over the previous 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period, and (3) < 2 gpm increase in unidentified leakage within the previous 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period in Mode 1. Total leakage is defined as the sum of identified and unidentified leakage.

DAEC EAL Threshold Value 1 uses the generic value of 10 GPM for unidentified leakage.

The 10 gpm value for the unidentified was selected as it is observable with normal control room indications.

Relief valve normal operation should be excluded from this IC. However, a relief valve that operates and fails to close per design should be considered applicable to this IC if the relief valve cannot be isolated.

SU5

EAL BASES DOCUMENT, EBD-S Rev. 8 next SYSTEM MALFUNCTION CATEGORY Page 10 of 33 DAEC EAL Threshold Value 2 uses identified leakage set at a higher value due to the lesser significance of identified leakage in comparison to unidentified or pressure boundary leakage.

REFERENCES:

1. Technical Specification 3.4.4, Coolant Leakage

2. Surveillance Test Procedure No. (STP) 3.0.0.0-01, Reactor Coolant System Leak Rate Calculation

3. Operating Instruction No. (01) 920, Drywell Sump System

4. Alarm Response Procedure (ARP) 1C04B, Reactor Water Cleanup and Recirculation

5. Alarm Response Procedure (ARP) 1C04C, Reactor Water Cleanup and Recirculation

6. UFSAR Section 5.2.5, Detection of Leakage through Reactor Coolant Pressure Boundary

7. UFSAR Section 15.6.6, Loss-of-Coolant-Accident

8. NEI 99-01 Rev. 5, NEI Methodology for Development of Emergency Action Levels SU5

EAL BASES DOCUMENT EBD-S Rev. 8 next SYSTEM MALFUNCTION CATEGORY Page 11 of 33 SU6 Loss of all onsite or offsite communications capabilities EVENT TYPE: Instrumentation/Communication OPERATING MODE APPLICABILITY: Power Operation, Startup, Hot S/D EAL THRESHOLD VALUE:

SU6.1 Loss of ALL of the following onsite communication methods affecting the ability to perform routine operations:

" Plant Operations Radio System

" In-Plant Telephones

• Plant Paging System OR SU6.2 Loss of ALL of the following offsite communication methods affecting the ability to perform offsite notifications:

" All telephone lines (commercial)

" Cell phones (including fixed cell phone system)

• Control Room fixed satellite phones

" Microwave Phone System

• Emergency Notification System (ENS)

" FTS Phone System DAEC EAL INFORMATION:

The purpose of this IC and its associated EALs is to recognize a loss of communications capability that either defeats the plant operations staff ability to perform routine tasks necessary for plant operations or the ability to communicate problems with offsite authorities. The loss of offsite communications ability is expected to be significantly more comprehensive than the condition addressed by 10 CFR 50.72.

The availability of one method of ordinary offsite communications is sufficient to inform state and local authorities of plant problems. This EAL is intended to be used only when SU6

BASES DOCUMENT EBD-S Rev. 8 next SYSTEM MALFUNCTION CATEGORY Page 12 of 33 extraordinary means (e.g., use of personal cell phones, relaying of information from radio transmissions, individuals being sent to offsite locations, etc.) are being utilized to make communications possible.

REFERENCES:

1. Emergency Plan, Section F, Emergency Communications

2. Abnormal Operating Procedure (AOP) 399, Loss of Communication

3. NEI 99-01 Rev. 5, NEI Methodology for Development of Emergency Action Levels SU6

EAL BASES DOCUMENT EBD-S Rev. 8 next SYSTEM MALFUNCTION CATEGORY Page 13 of 33 SU8 Inadvertent Criticality EVENT TYPE: Inadvertent Criticality OPERATING MODE APPLICABILITY: Hot S/D EAL THRESHOLD VALUE:

SU8.1 An UNPLANNED sustained positive period observed on nuclear instrumentation.

DAEC EAL INFORMATION:

This IC addresses inadvertent criticality events. This IC indicates a potential degradation of the level of safety of the plant, warranting a NOUE classification. This IC excludes inadvertent criticalities that occur during planned reactivity changes associated with reactor startups (e.g., criticality earlier than estimated).

This condition can be identified using period monitors. The term "sustained" is used in order to allow exclusion of expected short term positive periods from planned control rod movements for BWRs. These short term positive periods are the result of the increase in neutron population due to subcritical multiplication.

Escalation would be by the Fission Product Barrier Matrix, as appropriate to the operating mode at the time of the event.

REFERENCES:

1. NEI 99-01 Rev. 5, NEI Methodology for Development of Emergency Action Levels SU8

EAL',BASES DOCUMENT EBD-S Rev. 8 next SYSTEM MALFUNCTION CATEGORY Page 14 of 33 SA2 Automatic Scram fails to shutdown the reactor and the manual actions taken from 1C05 are successful in shutting down the reactor.

EVENT TYPE: RPS Failure OPERATING MODE APPLICABILITY: Power Operation, Startup EAL THRESHOLD VALUE:

SA2.1 An automatic scram failed to shutdown the reactor AND ANY of the following manual actions taken at 1C05 are successful in lowering reactor power below 5% power

" Manual Scram Pushbuttons

" Mode Switch to Shutdown

• Alternate Rod Insertion (ARI)

DAEC EAL INFORMATION:

NOTE If the mode switch is in Startup and the rods are fully inserted (i.e., the reactor is shutdown) prior to the automatic signal failure, then declaration of an Alert would not be required.

In this case, the event would be reported under 10 CFR 50.72 (b) (2) as a four hour report.

The reactor should be considered shutdown when it is producing less heat than the maximum decay heat load for which the safety systems are designed (typically 3 to 5%

power).

SA2

EAL BASES DOCUMENT EBD-S Rev. 8 next SYSTEM MALFUNCTION CATEGORY Page 15 of 33 This condition indicates failure of the automatic protection system to scram the reactor.

This condition is more than a potential degradation of a safety system in that a front line automatic protection system did not function in response to a plant transient. Thus the plant safety has been compromised because design limits of the fuel may have been exceeded. An Alert is indicated because conditions may exist that lead to potential loss of fuel clad or RCS and because of the failure of the Reactor Protection System to automatically shut down the plant.

Manual scram actions taken at 1C05 are any set of actions by the reactor operator(s) which causes or should cause control rods to be rapidly inserted into the core and shuts down the reactor. For the purpose of this EAL, a successful manual scram results when operators shut down the reactor using either the manual scram pushbuttons, the mode switch to Shutdown, or ARI.

If manual actions taken at 1C05 fail to shutdown the reactor, the event would escalate to a Site Area Emergency.

REFERENCES:

1. Integrated Plant Operating Instruction (IPOI) No. 5, Reactor Scram

2. ATWS Emergency Operating Procedure (EOP) - RPV Control

3. Emergency Operating Procedure (EOP) 1 - RPV Control

4. NEI 99-01 Rev. 5, NEI Methodology for Development of Emergency Action Levels SA2

,EAL BASES DOCUMENT EBD-S Rev. 8 next SYSTEM MALFUNCTION CATEGORY Page 16 of 33 SA4 UNPLANNED loss of safety system annunciation or indication in the control room with EITHER (1) a SIGNIFICANT TRANSIENT in progress, OR (2) compensatory indicators unavailable EVENT TYPE: Instrumentation/Communication OPERATING MODE APPLICABILITY: Power Operation, Startup, Hot S/D EAL THRESHOLD VALUE:

Note: The Emergency Director (OSM, EC or ER&RD) 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.

SA4.1 UNPLANNED loss of approximately 75% or more of any of the following for 15 minutes or longer:

0 1C03, 1C04 and 1C05 annunciators

• 1C03, 1C04 and 1C05 indications

• radiation monitor indications AND Either of the following conditions exist:

• A SIGNIFICANT TRANSIENT is in progress.

• Compensatory indications are unavailable.

DAEC EAL INFORMATION:

Control room panels 1C03, 1C04, and 1C05 contain the annunciators associated with safety systems at DAEC. Therefore, the DAEC EAL addresses UNPLANNED loss of annunciators on these panels. Compensatoryindicationsinclude the plant process computer, SPDS, plant recorders, or plant instrument displays in the control room.

UNPLANNED loss of annunciators or indicators excludes scheduled maintenance and testing activities. SIGNIFICANT TRANSIENT includes response to automatic or manually initiated functions such as scrams, runbacks involving greater than 25% thermal power change, ECCS injections, or thermal power oscillations of 10% or greater.

SA4

EAL BASES DOCUMENT EBD-S Rev. 8 next SYSTEM MALFUNCTION CATEGORY Page 17 of 33 Radiation monitor indications are included because these are identified in AOPs, EOPs and other EALs.

Under the conditions of concern, entry into AOP 302.2, Loss of Alarm Panel Power, would be made. The procedure requires alerting operators on shift to the nature of the lost annunciation. It further requires that operators be attendant and responsive to abnormal indications that relate to those systems and components that have lost annunciation.

Therefore, the generic criterion related to specific opinion of the Operations Shift Manager that additional operating personnel will be required to safely operate the unit is not included in the DAEC EAL because the concern is addressed by the AOP.

Quantification is arbitrary, however, it is estimated that if approximately 75% of the safety system annunciators or indicators are lost, there is an increased risk that a degraded plant condition could go undetected. It is not intended that plant personnel perform a detailed count of the instrumentation lost but use the value as a judgment threshold for determining the severity of the plant conditions. It is also not intended that the Shift Supervisor be tasked with making a judgment decision as to whether additional personnel are required to provide increased monitoring of system operation.

UNPLANNED loss of critical safety function indicators (i.e., EOP/EAL parameters) for 15 minutes or longer may preclude operators from taking actions to mitigate a transient.

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

Annunciators on 1C03, 1C04, and 1C05 share a common power supply from 125 VDC Division I that is fed through circuit breaker 1D1 3.

Indications of loss of annunciators associated with safety systems include:

• 125 VDC charger, battery, or system annunciators on control room panel 1C08

• Loss of "sealed in" annunciators at affected panels 0 Failure of affected annunciator panels shiftly testing by plant operators

• Expected alarms are not received

• Computer point ID B350 indicates "NSS ANN DC LOSS TRBL." (Loss of DC power to panels 1C03, 1C04, and 1C05)

SA4

EAL BASES DOCUMENT EBD-S Rev. 8 next SYSTEM MALFUNCTION CATEGORY Page 18 of 33 This Alert will be escalated to a Site Area Emergency if the operating crew cannot monitor the transient in progress due to a concurrent loss of compensatory indications with a SIGNIFICANT TRANSIENT in progress during the loss of annunciation or indication.

REFERENCES:

1. Operating Instruction (01) No. 317.2 Annunciator System

2. Abnormal Operating Procedure (AOP) 302.1, Loss of 125 VDC Power

3. Abnormal Operating Procedure (AOP) 302.2, Loss of Alarm Panel Power

4. NEI 99-01 Rev. 5, NEI Methodology for Development of Emergency Action Levels SA4

EAL,.BASES DOCUMENT EBD-S Rev. 8 next SYSTEM MALFUNCTION CATEGORY Page 19 of 33 SA5 AC power capability to essential busses reduced to a single power source for 15 minutes or longer such that any additional single failure would result in station blackout EVENT TYPE: Loss of Power OPERATING MODE APPLICABILITY: Power Operation, Startup, Hot S/D EAL THRESHOLD VALUE:

Note: The Emergency Director (OSM, EC or ER&RD) 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.

SA5.1 AC power capability to 1A3 or 1A4 busses reduced to a single power source for 15 minutes or longer AND Any additional single power source failure will result in station blackout.

DAEC EAL INFORMATION:

The DAEC EAL is written to address the underlying concern, i.e., only one AC power source remains and if it is lost, a Station Blackout will occur. Under the conditions of concern, entry into AOP 301, Loss of Essential Electrical Power, would be made under Tab 1, Loss of One Essential 4160V Bus, and/or under Tab 3, Loss of Offsite Power.

Indications/alarms related to degraded AC power are displayed on control room panel 1C08 and are listed in AOP 301 under "Probable Indications."

At DAEC, the Essential Busses of concern are 4160V Busses 1A3 and 1A4. Each of these busses feed their associated 480V and 120V AC busses through step down transformers. Onsite power sources at DAEC include the A and B Diesel Generators, 1G-31 and 1G-21, respectively.

SA5

EAL BASES DOCUMENT EBD-S Rev. 8 next SYSTEM MALFUNCTION CATEGORY Page 20 of 33

REFERENCES:

1. Abnormal Operating Procedure (AOP) 301, Loss of Essential Electrical Power

2. UFSAR Chapter 8 Electrical Power

3. Technical Specifications Section 3.8. Electrical Power Systems

4. NEI 99-01 Rev. 5, NEI Methodology for Development of Emergency Action Levels SA5

EAL BASES DOCUMENTEDS EBD-S I-Rev. 8 next SYSTEM MALFUNCTION CATEGORY Page 21 of 33 SSl Loss of all offsite and all onsite AC power to essential busses for 15 minutes or longer EVENT TYPE: Loss of Power OPERATING MODE APPLICABILITY: Power Operation, Startup, Hot S/D EAL THRESHOLD VALUE:

Note: The Emergency Director (OSM, EC or ER&RD) 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.

SS1.1 Loss of all offsite and all onsite AC power to 1A3 and 1A4 for 15 minutes or longer.

DAEC EAL INFORMATION:

In accordance with the generic guidance, DAEC is using a threshold of 15 minutes for Station Blackout to exclude transient or momentary power losses.

Under the conditions of concern, entry into AOP 301.1, Station Blackout, would be made under Tab 1. Indications/alarms related to station blackout are displayed on control room panel 1C08 and are listed in the procedure under "Probable Indications."

Consideration should be given to operable loads necessary to remove decay heat or provide Reactor Vessel makeup capability when evaluating loss of AC power to essential busses. Even though an essential bus may be energized, if necessary loads (i.e., loads that if lost would inhibit decay heat removal capability or Reactor Vessel makeup capability) are not operable on the energized bus then the bus should not be considered operable. If this bus was the only energized bus, then a Site Area Emergency per SS1 should be declared.

SS1

EAL BASES DOCUMENT EBD-S Rev. 8 next SYSTEM MALFUNCTION CATEGORY Page 22 of 33

REFERENCES:

1. Abnormal Operating Procedure (AOP) 301.1, Station Blackout

2. Technical Specifications Section 3.8, Electrical Power Systems

3. UFSAR Chapter 8, Electric Power

4. NEI 99-01 Rev. 5, NEI Methodology for Development of Emergency Action Levels SS1

EAL BASES DOCUMENT EBD-S Rev. 8 next SYSTEM MALFUNCTION CATEGORY Page 23 of 33 SS2 Automatic Scram fails to shutdown the reactor and manual actions taken from 1C05 are NOT successful in shutting down the reactor EVENT TYPE: RPS Failure OPERATING MODE APPLICABILITY: Power Operation, Startup EAL THRESHOLD VALUE:

SS2.1 An automatic scram failed to shutdown the reactor AND NONE of the following manual actions taken at 1C05 are successful in lowering reactor below 5% power

" Manual Scram Pushbuttons

• Mode Switch to Shutdown

• Alternate Rod Insertion (ARI)

DAEC EAL INFORMATION:

Under these conditions, the reactor is producing more heat than the maximum decay heat load for which the safety systems are designed and efforts to bring the reactor subcritical are unsuccessful. A Site Area Emergency is warranted because conditions exist that lead to IMMINENT loss or potential loss of both fuel clad and RCS.

The reactor should be considered shutdown when it is producing less heat than the maximum decay heat load for which the safety systems are designed (typically 3 to 5%

power).

Manual scram actions taken at 1C05 are any set of actions by the reactor operator(s) which causes or should cause control rods to be rapidly inserted into the core and shuts down the reactor.

Automatic and manual scram are not considered successful if action away from the reactor control console was required to scram the reactor. This EAL is still applicable even if actions taken away from the reactor control console are successful in shutting SS2

EAL BASES.. *DOCUMENT EBD-S Rev. 8 next SYSTEM MALFUNCTION CATEGORY Page 24 of 33 the reactor down because the design limits of the fuel may have been exceeded or because of the gross failure of the Reactor Protection System to shutdown the plant.

Although this IC may be viewed as redundant to the Fission Product Barrier Degradation IC, its inclusion is necessary to better assure timely recognition and emergency response.

Escalation of this event to a General Emergency would be due to a prolonged condition leading to an extreme challenge to either core-cooling or heat removal..

REFERENCES:

1. Integrated Plant Operating Instruction (IPOI) 5, Reactor Scram

2. ATWS Emergency Operating Procedure (EOP) - RPV Control

3. NEI 99-01 Rev. 5, NEI Methodology for Development of Emergency Action Levels SS2

ELBASES DOCUMENT EBD-S Rev. 8 next SYSTEM MALFUNCTION CATEGORY Page 25 of 33 SS3 Loss of all vital DC power for 15 minutes or longer EVENT TYPE: Loss of Power OPERATING MODE APPLICABILITY: Power Operation, Startup, Hot S/D EAL THRESHOLD VALUE:

Note: The Emergency Director (OSM, EC or ER&RD) 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.

SS3.1 Less than 105 VDC bus voltage on BOTH Div 1 and Div 2 125 VDC busses for 15 minutes or longer.

DAEC EAL INFORMATION:

Under the conditions of concern, AOP 302.1, Loss of 125 VDC Power, would be entered under Tab 3, Complete Loss of 125 VDC. Consequently, the DAEC EAL addresses loss of both divisions of the 125V DC system consistent with AOP. At DAEC, the 125V DC Systems ensure power is available for the reactor to be shutdown safely and maintained in a safe condition. The 125V System is divided into two independent divisions - Division I and Division II - with separate DC power supplies. These power supplies consist of two separate 125V batteries and chargers serving systems such as RCIC, RHR, EDGs, and HPCI. Complete loss of both 125V DC Divisions could compromise the ability to monitor and control the removal of decay heat during cold shutdown or refueling operations.

Escalation to a General Emergency would occur by Abnormal Rad Levels/Radiological Effluent or Fission Product Barrier Degradation.

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

SS3

EAL-BASES DOCUMENT EBD-S Rev. 8 next SYSTEM MALFUNCTION CATEGORY Page 26 of 33

REFERENCES:

1. Abnormal Operating Procedure (AOP) 302.1, Loss of 125 VDC Power

2. Abnormal Operating Procedure (AOP) 388, Loss of 250 VDC Power

3. Technical Specification 3.8, Electrical Power Systems

4. UFSAR Section 8.3, Onsite Power Systems

5. NEI 99-01 Rev. 5, NEI Methodology for Development of Emergency Action Levels SS3

EAL BASES DOCUMENT., EBD-S Rev. 8 next SYSTEM MALFUNCTION CATEGORY Page 27 of 33 SS6 Inability to monitor a SIGNIFICANT TRANSIENT in progress EVENT TYPE: Instrumentation/Communication OPERATING MODE APPLICABILITY: Power Operation, Startup, Hot S/D EAL THRESHOLD VALUE:

Note: The Emergency Director (OSM, EC or ER&RD) 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.

SS6.1 Loss of approximately 75% or more of any of the following for 15 minutes or longer:

• 1C03, 1C04 and 1C05 annunciators

• 1C03, 1C04 and 1C05 indications

• radiation monitor indications

• indicators needed to monitor criticality, or core heat removal, or Fission Product Barrier status AND A SIGNIFICANT TRANSIENT is in progress.

AND Compensatory indications are unavailable.

DAEC EAL INFORMATION:

This IC is intended to recognize the threat to plant safety associated with the complete loss of capability of the control room staff to monitor plant response to a SIGNIFICANT TRANSIENT.

Control room panels 1C03, 1C04, and 1C05 contain the annunciators, and indicators, associated with safety systems at DAEC.

SS6

EAL BASES DOCUMENT EBD-S Rev. 8 next SYSTEM MALFUNCTION CATEGORY Page 28 of 33 Radiation monitor indications are included because these are identified in AOPs, EOPs and other EALs.

Planned and UNPLANNED actions are not differentiated since the loss of instrumentation of this magnitude is of such significance during a transient that the cause of the loss is not a mitigating factor.

Quantification is arbitrary, however, it is estimated that if approximately 75% of the safety system annunciators or indicators are lost, there is an increased risk that a degraded plant condition could go undetected. It is not intended that plant personnel perform a detailed count of the instrumentation lost but use the value as a judgment threshold for determining the severity of the plant conditions. It is also not intended that the Shift Supervisor be tasked with making a judgment decision as to whether additional personnel are required to provide increased monitoring of system operation.

The DAEC EAL is written in terms of a significant transientin progress with loss of both safety system annunciators and loss of compensatory non-alarming instrumentation as well as loss of indications needed to monitor criticality, core heat removal, or fission product barrier status.

Significant transientincludes response to automatic or manually initiated functions such as scrams, runbacks involving greater than 25% thermal power change, ECCS injections, or undamped thermal power oscillations greater than 10%.

Compensatoryindicationsinclude the plant process computer, SPDS, plant recorders, or plant instrument displays in the control room. These indications are needed to monitor safety functions that are of concern in the generic EAL.

A Site Area Emergency is considered to exist if the control room staff cannot monitor safety functions needed for protection of the public while a significant transient is in progress.

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

Annunciators on 1C03, 1C04, and 1C05 share a common power supply from 125 VDC Division I that is fed through circuit breaker 1D1 3.

SS6

EAL BASES DOCUMENT EBD-S Rev. 8 next SYSTEM MALFUNCTION CATEGORY Page 29 of 33 Indications of loss of annunciators associated with safety systems include:

• 125 VDC charger, battery, or system annunciators on control room panel 1C08

" Loss of "sealed in" annunciators at affected panels

• Failure of affected annunciator panels shiftly testing by plant operators

• Expected alarms are not received

" Computer point ID B350 indicates "NSS ANN DC LOSS TRBL." (Loss of DC power to panels 1C03, 1C04, and 1C05)

REFERENCES:

1. Operating Instruction (01) No. 317.2, Annunciator System

2. Abnormal Operating Procedure (AOP) 302.1, Loss of 125 VDC Power

3. Abnormal Operating Procedure (AOP) 302.2, Loss of Alarm Panel Power

4. NEI 99-01 Rev. 5, NEI Methodology for Development of Emergency Action Levels SS6

EAL BASES DOCUMENT. EBD-S Rev. 8 next SYSTEM MALFUNCTION CATEGORY Page 30 of 33 SGI Prolonged loss of all offsite and all onsite AC power to essential busses EVENT TYPE: Loss of Power OPERATING MODE APPLICABILITY: Power Operation, Startup, Hot S/D EAL THRESHOLD VALUE:

SG1.1 Loss of all offsite and all onsite AC power to 1A3 and 1A4 AND ANY of the following:

" Restoration of power to either 1A3 or 1A4 in less than 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is NOT likely.

" RPV water level cannot be determined.

• RPV water level is LESS THAN +15 inches.

DAEC EAL INFORMATION:

Under prolonged Station Blackout (SBO) conditions, fission product barrier monitoring capability may be degraded. Although it may be difficult to predict when power can be restored, it is necessary to give the Emergency Director a reasonable idea of how quickly a General Emergency should be declared based on the following considerations:

• Are there any present indications that core cooling is already degraded to the point where a General Emergency is IMMINENT (i.e., loss of two barriers and a potential loss of the third barrier)?

• If there are presently no indications of degraded core cooling, how likely is it that power can be restored prior to occurrence of a General Emergency?

The first part of this EAL corresponds to the threshold conditions for Initiating Condition SS1 - namely, entry into AOP 301.1, Station Blackout. The second part of the EAL addresses the conditions that will escalate the SBO to General Emergency. Occurrence of any of the following is sufficient for escalation: (1) SBO coping capability exceeded, or (2) loss of drywell cooling that continues to make RPV water level measurements SG1

EAL BASES DOCUMENT EBD-S Rev. 8 next SYSTEM MALFUNCTION CATEGORY Page 31 of 33 unreliable, or (3) indications of inadequate core cooling. Each of these conditions is discussed below:

1. SBO Coping Capability Exceeded DAEC has a SBO coping duration of four hours. The likelihood of restoringat least one emergency bus should be based on a realisticappraisalof the situation since a delay in an upgrade decision based on only a chance of mitigating the event could result in a loss of valuable time in preparingand implementing public protective actions.

2. RPV Water Level Measurements Remaining Unreliable Flashing of the reference leg water will result in erroneously high RPV water level readings giving a false indication of actual water inventory and potentially indicating adequate core cooling when it may not exist. EOP Graph 1, RPV Saturation Temperature, defines the conditions under which RPV level instrument leg boiling may occur.

3. Indications of Inadequate Core Cooling DAEC uses the RPV level that is used for the Fuel Clad "potential loss" condition in the Fission Product Barrier Matrix. This is RPV level below +15 inches.

REFERENCES:

1. Abnormal Operating Procedure (AOP) 301.1, Station Blackout

2. Letter NG-92-0283, John F. Franz, Jr. to Dr. Thomas E. Murley, Response to Safety Evaluation by NRC-NRR "Station Blackout Evaluation Iowa Electric Light and Power Company Duane Arnold Energy Center," February 10, 1992

3. Emergency Operating Procedure (EOP)I - RPV Control

4. Emergency Operating Procedure (EOP) ALC - Alternate Level Control

5. NEI 99-01 Rev. 5, NEI Methodology for Development of Emergency Action Levels SG1

EAL BASES DOCUMENT EBD-S Rev. 8 next SYSTEM MALFUNCTION CATEGORY Page 32 of 33 SG2 Automatic Scram and all manual actions fail to shutdown the reactor and indication of an extreme challenge to the ability to cool the core exists EVENT TYPE: RPS Failure OPERATING MODE APPLICABILITY: Power Operation, Startup EAL THRESHOLD VALUE:

SG2.1 An automatic scram failed to shutdown the reactor AND ALL manual actions to lower reactor power below 5% power are unsuccessful.

AND Either of the following exist or have occurred due to continued power generation:

" RPV water level cannot be restored and maintained GREATER THAN -25 inches.

• Heat Capacity Limit (HCL) Curve (EOP Graph 4) exceeded.

DAEC EAL INFORMATION:

This EAL addresses conditions where failure of an automatic scram has occurred and all manual actions to lower reactor power below 5% have been unsuccessful AND a subsequent loss of adequate core cooling or decay heat removal capability occurs. If either of these challenges exists during an ATWS, a core melt sequence exists. In this situation, core degradation can occur rapidly. For this reason, the General Emergency declaration is intended to be anticipatory of the fission product barrier matrix declaration to permit maximum offsite intervention time.

If injection with all available Preferred and Alternate ATWS Injection Systems fails to provide sufficient injection to restore and maintain level above -25 inches (Minimum Steam Cooling RPV Water Level), adequate core cooling is threatened and submergence of the SG2

EAL BASES DOCUMENT EBD-S Rev. 8 next SYSTEM MALFUNCTION CATEGORY Page 33 of 33 core is attempted by flooding the primary containment. This is accomplished by transfer to and implementation of the DAEC Severe Accident Guidelines (SAGs).

The Heat Capacity Limit (EOP Graph 4) is defined to be the highest torus temperature at which initiation of RPV depressurization will not result in exceeding the Primary Containment Pressure Limit (the PCPL is 53 psig at the DAEC) before the rate of energy transfer from the RPV to the primary containment is within the capacity of the containment vent.

Control of torus temperature relative to the Heat Capacity Limit is directed in the Primary Containment Control Guideline, EOP 2. If the actions being taken in EOP 2 to preserve torus heat capacity are inadequate or not effective, RPV pressure must be reduced in order to remain below the Heat Capacity Limit. Therefore, actions in the RPV pressure control section of the ATWS EOP must accommodate these requirements. Failure to do so may lead to failure of the containment or loss of equipment necessary for the safe shutdown of the plant.

The reactor should be considered shutdown when it is producing less heat than the maximum decay heat load for which the safety systems are designed (typically 3 to 5%

power).

REFERENCES:

1. Emergency Operating Procedure ATWS EOP - RPV Control

2. NEI 99-01 Rev. 5, NEI Methodology for Development of Emergency Action Levels SG2