Text

Response to Request for Additional Information Amendment to Change Emergency Action Levels
	ML12076A236
Person / Time
Site:	Duane Arnold
Issue date:	03/16/2012
From:	Wells P; NextEra Energy Duane Arnold
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	NG-12-0064
	Download: ML12076A236 (135)
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DUANE ARNOLD NG-12-0064 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 Request for Additional Information Re: Amendment to Change Emergency Action Levels

Reference:

1.

2.

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])

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)

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 is provided in Enclosure 1 to this letter.

A Revised Comparison Matrix of the NEI 99-01, Rev. 5 generic guidance to the proposed NextEra Energy Duane Arnold Emergency Classification System is provided in Enclosure 2 to this letter. Revised Emergency Action Level Design Basis Documents are provided in Enclosure 3 to this letter.

NextEra Energy Duane Arnold, LLC requests approval of the amendment request of Reference 1 by June 1, 2012.

NextEra Energy Duane Arnold, LLC, 3277 DAEC Road, Palo, IA 52324

NG-12-0064 March 16, 2012 Page 2 of 2 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 declare under penalty of perjury that the foregoing is true and correct. Executed on h 16,2012.

Peter Wells Vice President, Duane Arnold Energy Center NextEra Energy Duane Arnold, LLC Enclosures

1.

2.

3.

Response to Request for Additional Information Re: Amendment to Change Emergency Action Levels Revised Comparison Matrix of NEI 99-01, Rev. 5 Generic Guidance to Proposed NextEra Energy Duane Arnold Emergency Classification System Revised Emergency Action Level Design Basis Documents for Nextera Energy Duane Arnold cc:

M. Rasmusson (State of Iowa) to NG-12-0064 Page 1 of 8 ENCLOSURE 1 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION RE: AMENDMENT TO CHANGE EMERGENCY ACTION LEVELS to NG-12-0064 Page 2 of 8 Response to Request for Additional Information Re: Amendment to Change Emergency Action Levels NRC Question #1:

Section 3.0 of Enclosure 1, Evaluation of Proposed Change the last sentence in the last paragraph states that any changes to the approved ICs (initiating conditions) and EALs will be made in accordance with 10 CFR 50.54(q). It is the expectation of the staff that changes to the EAL basis information and operating modes will be controlled in the same way as any changes to the approved ICs and EALs. Please explain the rationale for not controlling the EAL Basis Document, as a whole, in accordance with 10 CFR 50.54(q) or revise this statement accordingly.

NextEra Energy Duane Arnold Response:

EAL Basis document information and operating modes are planned to be controlled in the same way as any changes to the approved ICs and EALs.

NRC Question #2:

EAL RU2.2: The use of offscale high readings for EAL thresholds is problematic as it is difficult to differentiate between failed instrumentation and actual plant conditions. Please explain how you will differentiate between the two, in the time allowed, for EAL declaration purposes and how this wording will not result in erroneous declarations, or revise accordingly.

NextEra Energy Duane Arnold Response:

RU2.2 has been revised as follows:

Any UNPLANNED VALID ARM reading offscale high or GREATER THAN 1000 times normal* reading.

NRC Question #3:

EAL RA2: The staff requires further justification for the removal of the level threshold previously approved for EAL RA2.3. The proposed justification for the change is confusing in that it uses a calculation as a basis for the change when in fact there are three unique EAL thresholds expected for this EAL. One is based upon a valid Hi-Rad alarm from RM-9178, a valid reading > 100 millirem/hr from RM-9178, and indication of lowering fuel pool water level from LI-3413. Redundancy is expected for this EAL due to the significance of the concern and the availability of instrumentation. Please provide further justification that supports this revision, or revise accordingly.

to NG-12-0064 Page 3 of 8 NextEra Energy Duane Arnold Response:

In November 2009, during a Licensed Operator Requalification training session on proposed changes to EALs based on NEI 99-01 Rev. 5 guidance, a student asked if the Fuel Pool Level Indicator (LI-3413) could indicate down to 16 feet (as EAL RA2.3 states). This prompted a condition report to be written which resulted in two separate condition evaluations.

Prior to completion of the evaluations, the Assistant Operations Manager implemented immediate interim guidance on how to use alternate indications to determine spent fuel pool level.

The first evaluation was a functionality assessment of the level indicator. This assessment determined that LI-3413 is considered non-functional as an indication for entry into EAL RA2.3 as it will not measure down to 16 feet. Compensatory actions were added to Annunciator Response Procedure (ARP) for Fuel Pool Level (ARP-1C04B). These compensatory actions provide assurance to properly classify the EAL until permanent changes to the EAL are made.

The second evaluation was to determine how to properly disposition or revise the EAL. This evaluation determined that a lack of historical documentation makes it difficult to determine when the LI was added to the EALs. It is believed to have occurred between 1990 and 1999. It was also determined that that there must have been no analysis of the limitations of the instrument when adding to the EALs since the instrument can only read in the range of 35 to 40.

The evaluation provided 4 different options as follows:

Option #1: Since the SFP water low level alarm (per ARP 1C04B) is 371 and the Tech Spec level is at 36, one option could be to change the threshold for RA2.3 to a value lower than 36 yet still within the instrument range of 35.

This option would be significantly more conservative than the current 16 threshold.

Option #2: Install an instrument which can measure SFP water level down to the EAL threshold value of 16.

Option #3:

o Current EAL RA2.1 reads as follows:

Report of any of the following:

o Valid ARM Hi Rad alarm for the Refueling Floor North End (RM 9163),

Refueling Floor South End (RM 9164), New Fuel Storage (RM 9153),

or Spent Fuel Storage Area (RM 9178).

o Valid Refueling Floor North End (RM-9163), Refueling Floor South End (RM-9164), or New Fuel Storage Area (RM-9153) ARM Reading GREATER THAN 10 mRem/hr o Valid Spent Fuel Storage Area ARM (RM-9178) Reading GREATER THAN 100 mRem/hr o Recommend: Radiation Protection to perform calculation to determine area dose rates near Refuel Floor ARMs with to NG-12-0064 Page 4 of 8 SFP water level 2 above the fuel. Results of this calculation will be provided to EP personnel for determination of necessity of EAL RA2.3 due to the hypothesis that RM-9178 would reach 100 milliRem/hr before LI-3413 reaches 16'.

o While there exists no calculation at this time that substantiates this hypothesis, if validated, this would mean that EAL RA2.1 would be declared before SFP level ever reached 16 and would therefore indicate that RA2.3 would never be declared. As a result, this EAL (RA2.3) could be removed.

o Option #4: During the implementation of NEI 99-01 Rev. 5 EALs, use the words from the NEI document to replace DAEC current EAL statement for RA2.3, without including instrument IDs. Language is as follows: A water level drop in the reactor refueling cavity, spent fuel pool or fuel transfer canal that will result in irradiated fuel becoming uncovered.

The site chose Option 3. Radiation Protection performed a Radiological Calculation which is the justification for removing EAL RA2.3 from the DAEC scheme.

According to the very conservative calculation, immediately prior to the fuel pool level reaching 16, RM-9178 will reach 100 millirem/hr, meeting the threshold for EAL RA2.1. The assumptions in the calculation provided conservatism by assuming (1) a limited number (144) of fuel bundles were present in the pool, (2) the bundles were located in the 12X12 rack the furthest distance from RM-9178, (3) the fuel bundles are assumed to have been removed from the core 3 months prior and (4) the bundles were removed from a core with a power of 1690 MWTh.

NRC Question #4:

EAL RA3: Please explain why the Secondary Alarm Station is used as a threshold for this EAL as this EAL only requires either the Control Room, and either the Central Alarm Station or Secondary Alarm Station, not both (typically). Please justify or revise accordingly.

NextEra Energy Duane Arnold Response:

SAS is being removed from EAL 3.1 as follows:

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)

Secondary Alarm Station (by survey) to NG-12-0064 Page 5 of 8 The basis is also being revised as follows:

There are no significant deviations from the generic EALs. Per the UFSAR,The control room and the Central Alarm Station (CAS) are is the only areas identified that is are required to be continuously occupied to achieve and maintain safe operation or safe shutdown following design basis accidents. However, the Central Alarm Station (CAS) and Secondary Alarm Station (SAS) shall be included in this EAL as access to these areas must be maintained and the CAS & SAS are continuously occupied.

NRC Question #5:

EAL CG1: The staff noted an apparent typographical error in the IC. In addition, the staff recommends developing this EAL using the table as provided in the endorsed EAL scheme development guidance as it aids in clarifying the logic for this EAL, however, the staff has no technical opposition to the proposed format and wording.

Also, please move the wording for 30 minutes or longer from the bulleted list of conditions provided in CG1.2 to the threshold for RPV level cannot be monitored.

This is a known error in the guidance document which will be corrected in a future revision.

NextEra Energy Duane Arnold Response:

CG1 has been revised as follows:

The redundant with containment challenged was deleted from the IC The EAL was revised to model the endorsed EAL scheme development guidance from NEI 99-01 Rev. 5. The bullets were placed in a table format.

For CG1.2, for 30 minutes or longer was moved from the bulleted list to the EAL threshold for RPV level cannot be monitored.

NRC Question #6:

EAL EU1: Please provide basis information related to how security events at the ISFSI [independent spent fuel storage installations] are classified.

NextEra Energy Duane Arnold Response:

The following clarification has been added to HS4.1:

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 to NG-12-0064 Page 6 of 8 NRC Question #7:

EAL HU2.1: Note that consideration of adjacent areas was removed as history had shown this to be an area of confusion throughout the nuclear industry and resulted in a delay in classification in some cases.

NextEra Energy Duane Arnold Response:

HU2.1 has been revised as follows:

FIRE NOT extinguished within 15 minutes of Control Room notification or verification of a Control Room FIRE alarm in ANY building or area in direct contact with or immediately adjacent to a Safe Shutdown/Vital Area.

HU2.1 basis information has been revised as follows:

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

Immediately adjacent implies that the area immediately adjacent contains or may contain equipment or cabling that could impact equipment located in VITAL AREAS or the fire could damage equipment inside VITAL AREAS or that precludes access to VITAL AREAS.

NRC Question #8:

EAL HA1: Please be aware that the areas of concern are expected to be areas where equipment necessary for safe operation, shutdown, and/or cooldown are located AND where the equipment is at risk from the given hazard. For example, not all areas are susceptible to vehicle crashes, high winds, tornados, or turbine blade failure. Please ensure to consider this when developing the list.

NextEra Energy Duane Arnold Response:

HA1.2 and HA1.3 have been combined into 1 EAL as follows:

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

o VISIBLE DAMAGE to any of the following structures:

Emergency Diesel Generator Rooms Control Building Reactor Building Pumphouse Intake Structure to NG-12-0064 Page 7 of 8 Condensate Storage Tank Area OR o Control Room indication of degraded performance of a System of Concern.

Also HA1.4 for turbine failure generated projectile has been revised and renumbered as HA1.3 as follows:

HA1.3 Turbine failure-generated PROJECTILES resulting in:

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

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

NRC Question #9:

EAL HA2: Please provide the list of areas applicable to this EAL.

NextEra Energy Duane Arnold Response:

HA2 basis information has been revised as follows:

The Safe Shutdown/Vital Area Table has been added.

NRC Question #10:

EAL HA32: Please revise the list of areas applicable to this EAL to ensure consistent, and accurate, EAL declaration. Only those areas that contain equipment that must be operated locally for safe operation, safe shutdown, or safe cooldown should be listed. If an area does not contain equipment that must be operated locally, then consideration for not including this area should be given. Note that the need to operate the equipment is not a threshold, i.e., the impediment to access an area, developed as per the above guidance, is the expectation for this EAL. Also, consider adding guidance information that allows the decision-maker to use reasonable judgment when considering declaring this EAL. For example, a small CO2 bottle leaking in a large room/area may not warrant declaration.

to NG-12-0064 Page 8 of 8 NextEra Energy Duane Arnold Response:

This RAI refers to HA3.2 which was deleted in NEI 99-01 Rev. 5 (from Rev. 4).

NextEra Energy Duane Arnold is following Rev. 5 guidance and deleted HA3.2 with the submittal. Discussions with the NRC (Don Johnson) regarding this RAI have centered on HA3.1. Therefore, NextEra Energy Duane Arnolds response to this RAI is for addressing HA3.1 Operations in conjunction with Emergency Preparedness reviewed plant procedures associated with shutdown and cooldown. This review consisted of identifying those local actions required by Operators in areas requiring access for normal plant shutdown and cooldown.

No required local Operator actions, outside of the Control Room, were identified assuming no equipment failure, no coincident events and no accidents. Specific to this review, were the consideration of Station Blackout required actions and local action required for initiation of shutdown cooling.

Station Blackout required actions were reviewed and considered, however a Station Blackout is considered a coincident event and therefore does not apply to this EAL. In addition, Station Blackout is covered by the System Malfunction EALs SS1.1 and SG1.1.

Also reviewed local manual operation of the Low Pressure Cooling Injection (LPCI) system crosstie during Mode 3 for initiation of Shutdown Cooling.

Since LPCI is inop/removed from service as a result of this evolution, this also would not apply to this EAL when considering NEI guidance:

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 declared as 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.

As a result, NextEra Energy Duane Arnold has removed this EAL (HA3.1) from the scheme.

NG-12-0064 37 Pages Follow ENCLOSURE 2 REVISED COMPARISON MATRIX OF NEI 99-01, REV. 5 GENERIC GUIDANCE TO PROPOSED NEXTERA ENERGY DUANE ARNOLD EMERGENCY CLASSIFICATION SYSTEM

Last SER Approved EALs NEI 99-01 Rev. 5 EALs Proposed EALs Difference or Deviation Explanation RU1 Any Unplanned Release of Gaseous or Liquid Radioactivity to the Environment That Exceeds Two Times the Offsite Dose Assessment Manual (ODAM) Limit and is Expected to Continue For 60 Minutes or Longer Op. Modes: ALL RU1.1 Valid Reactor Building ventilation rad monitor (Kaman 3/4, 5/6, 7/8) or Turbine Building ventilation rad monitor (Kaman 1/2) reading that exceeds 1 E-3 Ci/cc and is expected to continue for 60 minutes or longer.

OR RU1.2 Valid Offgas Stack rad monitor (Kaman 9/10) reading that exceeds 2.0 E-1 Ci/cc and is expected to continue for 60 minutes or longer.

OR RU1.3 Valid LLRPSF rad monitor (Kaman 12) reading that exceeds 1.0 E-3 Ci/cc and is expected to continue for 60 minutes or longer.

OR RU1.4 Valid GSW rad monitor (RIS-4767) reading that exceeds 3E+3 CPS and is expected to continue for 60 minutes or longer.

OR RU1.5 Valid RHRSW & ESW rad monitor (RM-1997) reading that exceeds 8E+2 CPS and is expected to continue for 60 minutes or longer.

OR RU1.6 Valid RHRSW & ESW Rupture Disc rad monitor (RM-4268) reading that exceeds 1E+3 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 in excess of 2 times ODAM limit and is expected to continue for 60 minutes or longer.

AU1 Any release of gaseous or liquid radioactivity to the environment GREATER THAN 2 times the Radiological Effluent Technical Specifications/ODCM for 60 minutes or longer.

Operating Mode Applicability: All 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 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.

1. VALID reading on ANY of the following radiation monitors GREATER THAN the reading shown for 60 minutes or longer:

(site specific monitor list and threshold values)

OR

2. VALID reading on any effluent monitor reading GREATER THAN 2 times the alarm setpoint established by a current radioactivity discharge permit for 60 minutes or longer.

OR

3. Confirmed sample analyses for gaseous or liquid releases indicates concentrations or release rates GREATER THAN 2 times (site specific RETS values) for 60 minutes or longer.

OR

4. VALID reading on perimeter radiation monitoring system reading GREATER THAN 0.10 mR/hr above normal* background for 60 minutes or longer. [for sites having telemetered perimeter monitors]

OR

5. VALID indication on automatic real-time dose assessment capability indicating GREATER THAN (site specific value) for 60 minutes or longer. [for sites having such capability]

Normal can be considered as the highest reading in the past twenty-four hours excluding the current peak value.

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 Operating Modes: ALL 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) reading GREATER THAN 1.0 E-3 Ci/cc and is expected to continue for 60 minutes or longer.

RU1.2 VALID Offgas Stack rad monitor (Kaman 9/10) reading GREATER THAN 2.0 E-1 Ci/cc and is expected to continue for 60 minutes or longer.

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

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.

RU1.5 VALID RHRSW & ESW rad monitor (RM-1997) reading GREATER THAN 800 (8.0 X 102) CPS and is expected to continue for 60 minutes or longer.

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.

RU1.7 Confirmed sample analyses for gaseous or liquid releases indicates concentrations or DIFFERENCE Changing A to R for IC designator to signify Rad Table.

Provided clarification on ERO positions associated with who could be Emergency Director, depending on which facility is in command and control. This is consistent throughout this recognition category.

and is expected to continue for 60-minutes or longer meets intent of for 60-minutes or longer and is consistent with NEI 99-01 rev 5 bases information and the Note associated with AU1 IC.

NEI EAL RU1 is broken down into 3 DAEC EALs (RU1.1, 1.2 and 1.3).

NEI EAL RU2 is broken down into 3 DAEC EALs (RU1.4, 1.5, 1.6)

DAEC does not have a perimeter radiation monitoring system.

DAEC does not have automatic dose assessment capability.

Listing specific values with instrument IDs and names.

Consistent with current NEI EAL and last SER approved EAL.

At DAEC, ODAM is the same as RETS/ODCM

Last SER Approved EALs NEI 99-01 Rev. 5 EALs Proposed EALs Difference or Deviation Explanation release rates GREATER THAN 2 times ODAM limit and is expected to continue for 60 minutes or longer.

RU2 Unexpected Increase in Plant Radiation Operating Modes: ALL RU2.1 Unplanned valid Refuel Floor ARM reading increase with an uncontrolled loss of reactor cavity, fuel pool, or fuel transfer canal water level with all irradiated fuel assemblies remaining covered by water 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 ARM reading offscale high or GREATER THAN 1000 times normal*

reading.

Normal levels can be considered as the highest reading in the past twenty-four hours excluding the current peak value.

AU2 UNPLANNED rise in plant radiation levels Operating Modes: ALL

1.

a.

UNPLANNED water level drop in a reactor refueling pathway as indicated by (site specific level or indication).

AND

b.

VALID Area Radiation Monitor reading rise on (site specific list).

OR

2.

UNPLANNED VALID Area Radiation Monitor readings or survey results indicate a rise by a factor of 1000 over normal* levels.

Normal can be considered as the highest reading in the past twenty-four hours excluding the current peak value.

RU2 UNPLANNED rise in plant radiation levels Operating Modes: ALL 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.

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 reading in the past twenty-four hours excluding the current peak value.

DIFFERENCE Changing A to R for IC designator to signify Rad Table.

For clarity purposes, reworded and changed the order of 1st EAL statement. Provided site-specific indications.

Added UNPLANNED prior to VALID to match intent in the IC.

VALID area radiation monitor is equivalent to Refuel Floor ARM.

refueling pathway is equivalent to reactor cavity, fuel pool, or fuel transfer canal which is consistent with basis discussion for this EAL.

Added any to the ARM threshold to ensure all ARMs are considered.

Split EAL #2 into 2 separate EALs to differentiate between surveys and ARM readings. This was done for human performance factors and will help to prevent the survey part of the EAL getting buried and missed.

Last SER Approved EALs NEI 99-01 Rev. 5 EALs Proposed EALs Difference or Deviation Explanation RA1 Any Unplanned Release of Gaseous or Liquid Radioactivity to the Environment that Exceeds 200X the Offsite Dose Assessment Manual (ODAM) Limit and is Expected to Continue for 15 Minutes or Longer Operating Modes: ALL 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 that exceeds 3 E-2 Ci/cc and is expected to continue for 15 minutes or longer.

OR RA1.2 Valid Offgas Stack rad monitor (Kaman 9/10) reading that exceeds 6 E+0 Ci/cc and is expected to continue for 15 minutes or longer.

OR RA1.3 Valid LLRPSF rad monitor (Kaman 12) reading that exceeds 1 E-1 Ci/cc and is expected to continue for 15 minutes or longer.

OR RA1.4 Valid GSW rad monitor (RIS-4767) reading that exceeds 3E+5 CPS and is expected to continue for 15 minutes or longer.

OR RA1.5 Valid RHRSW & ESW rad monitor (RM-1997) reading that exceeds 8E+4 CPS and is expected to continue for 15 minutes or longer.

OR RA1.6 Valid RHRSW & ESW Rupture Disc rad monitor (RM-4268) reading that exceeds 1E+5 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 with a release duration expected to continue for 15 minutes or longer in excess of 200 times ODAM limit.

AA1 Any release of gaseous or liquid radioactivity to the environment GREATER THAN 200 times the Radiological Effluent Technical Specifications/ODCM for 15 minutes or longer.

Operating Modes: ALL 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 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.

1. VALID reading on ANY of the following radiation monitors GREATER THAN the reading shown for 15 minutes or longer:

(site specific monitor list and threshold values)

OR

2. VALID reading on any effluent monitor reading GREATER THAN 200 times the alarm setpoint established by a current radioactivity discharge permit for 15 minutes or longer.

OR

3. Confirmed sample analyses for gaseous or liquid releases indicates concentrations or release rates GREATER THAN 200 times (site specific RETS values) for 15 minutes or longer.

OR

4. VALID reading on perimeter radiation monitoring system reading GREATER THAN 10.0 mR/hr above normal* background for 15 minutes or longer. [for sites having telemetered perimeter monitors]

OR

5. VALID indication on automatic real-time dose assessment capability indicating GREATER THAN (site specific value) for 15 minutes or longer. [for sites having such capability]

Normal can be considered as the highest reading in the past twenty-four hours excluding the current peak value.

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 Operating Modes: ALL 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 Ci/cc and is expected to continue for 15 minutes or longer.

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

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

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.

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.

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.

RA1.7 Confirmed sample analyses for gaseous or DIFFERENCE Changing A to R for IC designator to signify Rad Table.

Provided clarification on ERO positions associated with who could be Emergency Director, depending on which facility is in command and control. This is consistent throughout this recognition category.

Do not have a perimeter radiation monitoring system.

and is expected to continue for 15-minutes or longer meets intent of for 15-minutes or longer and is consistent with NEI 99-01 rev 5 bases information and the Note associated with AA1 IC.

DAEC does not have automatic dose assessment capability.

Listing specific values with instrument IDs and names.

Consistent with current NEI EAL and last SER approved EAL.

NEI EAL RA1 is broken down into 3 DAEC EALs (RA1.1, 1.2 and 1.3).

NEI EAL RA2 is broken down into 3 DAEC EALs (RA1.4, 1.5, 1.6)

At DAEC, ODAM is the same as RETS/ODCM

Last SER Approved EALs NEI 99-01 Rev. 5 EALs Proposed EALs Difference or Deviation Explanation liquid releases indicates concentrations or release rates GREATER THAN 200 times ODAM limit and is expected to continue for 15 minutes or longer.

Last SER Approved EALs NEI 99-01 Rev. 5 EALs Proposed EALs Difference or Deviation Explanation RA2 Damage to Irradiated Fuel or Loss of Water Level that Has or Will Result in the Uncovering of Irradiated Fuel Outside the Reactor Vessel Operating Modes: ALL RA2.1 Report of any of the following:

Valid ARM Hi Rad alarm for the Refueling Floor North End (RM 9163), Refueling Floor South End (RM 9164), New Fuel Storage (RM 9153), or Spent Fuel Storage Area (RM 9178).

Valid Refueling Floor North End (RM-9163), Refueling Floor South End (RM-9164), or New Fuel Storage Area (RM-9153) ARM Reading GREATER THAN 10 mRem/hr Valid Spent Fuel Storage Area ARM (RM-9178) Reading GREATER THAN 100 mRem/hr OR RA2.2 Valid water level reading LESS THAN 450 inches as indicated on LI-4541 (floodup) for the Reactor Refueling Cavity that will result in Irradiated Fuel uncovering.

OR RA2.3 Valid Fuel Pool water level indication (LI-3413) LESS THAN 16 feet that will result in Irradiated Fuel uncovering.

AA2 Damage to irradiated fuel or loss of water level that has resulted or will result in the uncovering of irradiated fuel outside the reactor vessel.

Operating Modes: ALL

1.

A water level drop in the reactor refueling cavity, spent fuel pool or fuel transfer canal that will result in irradiated fuel becoming uncovered.

OR

2.

A VALID alarm or (site specific elevated reading) on ANY of the following due to damage to irradiated fuel or loss of water level.

(site specific radiation monitors)

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 Operating Modes: ALL 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).

b. VALID reading GREATER THAN 10 milliem/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)

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

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

DIFFERENCE Changing A to R for IC designator to signify Rad Table.

Changed irradiated to spent since NEI 99-01 Rev. 5 bases state this Initiating Condition applies to spent fuel requiring water coverage.

Reordered EAL 1 and 2 Listing specific values with instrument IDs and names Consistent with last SER Approved EAL and current NEI EAL.

VALID WR GEMAC Floodup indication (LI-4541) LESS THAN 450 inches is equivalent to A water level drop in the Previous EAL RA2.3 is redundant to (new) RA2.1a for RM-9178 and was therefore removed with this revision. EAL deleted as a result of Rad Engineering Calculation 10-002A, RA3 Release of Radioactive Material or Increases in Radiation Levels Within the Facility That Impedes Operation of Systems Required to Maintain Safe Operations or to Establish or to Maintain Cold Shutdown Operating Modes: ALL RA3.1 Valid area radiation levels GREATER THAN 15 mRem/hr in any of the following areas:

Control Room (RM 9162)

Central Alarm Station (by survey)

Secondary Alarm Station (by AA3 Rise in radiation levels within the facility that impedes operation of systems required to maintain plant safety functions.

Operating Modes: ALL Dose rate GREATER THAN 15 mR/hr in ANY of the following areas requiring continuous occupancy to maintain plant safety functions:

(site specific area list)

RA3 Rise in radiation levels within the facility that impedes operation of systems required to maintain plant safety functions.

Operating Modes: ALL 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)

DIFFERENCE Changing A to R for IC designator to signify Rad Table.

Consistent with last SER Approved EAL and current NEI EAL.

Removed requiring continuous occupancy to simplify and eliminate potential over analyzing by the Operating Crews.

Included site-specific loacations/instruments Remote Shutdown Panel EAL removed because it is not continuously occupied and in the event its being used because of Control Room habitability, ALERT HA5.1 would be classified.

Although the Rad Waste Control Room is continuously staffed, it is not required to maintain plant safety

Last SER Approved EALs NEI 99-01 Rev. 5 EALs Proposed EALs Difference or Deviation Explanation survey)

OR RA3.2 Valid area radiation monitor (RE-9168),

reading GREATER THAN 500 mRem/hr affecting the Remote Shutdown Panel, 1C388.

functions.

Changed mR/hr to millirem/hr in accordance with site expectations to spell it out to prevent potential miscommunication.

RS1 Offsite Dose Resulting from an Actual or Imminent Release of Gaseous Radioactivity Exceeds 100 mRem TEDE or 500 mRem CDE Thyroid for the Actual or Projected Duration of the Release Operating Modes: ALL RS1.1 Dose assessment using actual meteorology indicates doses GREATER THAN 100 mRem TEDE or 500 mRem 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 E-2 Ci/cc and is expected to continue for 15 minutes or longer.

Valid Offgas Stack rad monitor (Kaman 9/10) reading GREATER THAN 4 E+1 Ci/cc and is expected to continue for 15 minutes or longer.

OR RS1.3 Field survey results indicate closed window dose rates exceeding 100 mRem/hr expected to continue for more than one hour at or beyond the site boundary; or analyses of field survey samples indicate thyroid CDE of 500 mRem for one hour of inhalation at or beyond the site boundary.

AS1 Off-site dose resulting from an actual or IMMINENT release of gaseous radioactivity GREATER THAN 100 mrem TEDE or 500 mrem Thyroid CDE for the actual or projected duration of the release.

Operating Modes: ALL 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 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.

1.

VALID reading on ANY of the following radiation monitors GREATER THAN the reading shown for 15 minutes or longer:

(site specific monitor list and threshold values)

OR

2.

Dose assessment using actual meteorology indicates doses GREATER THAN 100 mrem TEDE or 500 mrem thyroid CDE at or beyond the site boundary.

OR

3.

VALID perimeter radiation monitoring system reading GREATER THAN 100 mR/hr for 15 minutes or longer. [for sites having telemetered perimeter monitors]

OR

4.

Field survey results indicate closed window dose rates GREATER THAN 100 mR/hr expected to continue for 60 minutes or longer; or analyses of field survey samples indicate thyroid CDE GREATER THAN 500 mrem for one hour of inhalation, at or beyond the site boundary.

RS1 Off-site dose resulting from an actual or IMMINENT release of gaseous radioactivity GREATER THAN 100 millirem TEDE or 500 millirem Thyroid CDE for the actual or projected duration of the release.

Operating Modes: ALL 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)

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 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+1 Ci/cc and is expected to continue for 15 minutes or longer.

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.

RS1.4 Analyses of field survey samples indicate DIFFERENCE Changing A to R for IC designator to signify Rad Table.

Provided clarification on ERO positions associated with who could be Emergency Director, depending on which facility is in command and control. This is consistent throughout this recognition category.

Consistent with last SER Approved EAL and current NEI EAL.

Changed order of EALs to put preferred method first.

NEI EAL RS1 is RS1.2.

Added at or beyond the site boundary to RS1.3 for the dose rate threshold to ensure that EAL Decision-Makers understand the boundaries of the thresholds.

Changed mR to millirem in accordance with site expectations to spell it out to prevent potential miscommunication.

Included site-specific instruments for when dose assessment is unavailable. Also included specific rad concentrations for thresholds.

Split EAL #4 into 2 separate EALs to differentiate between field survey readings and sample analysis results. This was done for human performance factors and will help to prevent the 2nd half of the EAL from getting missed.

DAEC does not have a perimeter radiation monitoring system.

Last SER Approved EALs NEI 99-01 Rev. 5 EALs Proposed EALs Difference or Deviation Explanation thyroid CDE GREATER THAN 500 millirem for one hour of inhalation at or beyond the site boundary.

RG1 Offsite Dose Resulting from an Actual or Imminent Release of Gaseous Radioactivity that Exceeds 1000 mRem TEDE or 5000 mRem CDE Thyroid for the Actual or Projected Duration of the Release Using Actual Meteorology Operating Modes: ALL RG1.1 Dose assessment using actual meteorology indicates doses GREATER THAN 1000 mRem TEDE or 5000 mRem 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 E-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 E+2 Ci/cc and is expected to continue for 15 minutes or longer.

OR RG1.3 Field survey results indicate closed window dose rates exceeding 1000 mRem/hr expected to continue for more than one hour at or beyond the site boundary; or analyses of field survey samples indicate thyroid CDE of 5000 mRem for one hour of inhalation at or beyond the site boundary.

AG1 Off-site dose resulting from an actual or IMMINENT release of gaseous radioactivity GREATER THAN 1000 mrem TEDE or 5000 mrem Thyroid CDE for the actual or projected duration of the release using actual meteorology.

Operating Modes: ALL 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 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.

1.

VALID reading on ANY of the following radiation monitors GREATER THAN the reading shown for 15 minutes or longer:

(site specific monitor list and threshold values)

OR

2.

Dose assessment using actual meteorology indicates doses GREATER THAN 1000 mrem TEDE or 5000 mrem thyroid CDE at or beyond the site boundary.

OR

3.

VALID perimeter radiation monitoring system reading GREATER THAN 1000 mR/hr for 15 minutes or longer. [for sites having telemetered perimeter monitors]

OR

4.

Field survey results indicate closed window dose rates GREATER THAN 1000 mR/hr expected to continue for 60 minutes or longer; or analyses of field survey samples indicate thyroid CDE GREATER THAN 5000 mrem for one hour of inhalation, at or beyond site boundary.

RG1 Off-site 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.

Operating Modes: ALL 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)

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 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 Ci/cc and is expected to continue for 15 minutes or longer.

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 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.

DIFFERENCE Changing A to R for IC designator to signify Rad Table.

Provided clarification on ERO positions associated with who could be Emergency Director, depending on which facility is in command and control. This is consistent throughout this recognition category.

Consistent with last SER Approved EAL and current NEI EAL.

Changed order of EALs to put preferred method first.

Changed mR to millirem in accordance with site expectations to spell it out to prevent potential miscommunication.

Included site-specific instruments for when dose assessment is unavailable. Also included specific rad concentrations for thresholds.

Added at or beyond the site boundary to RG1.3 for the dose rate threshold to ensure that EAL Decision-Makers understand the boundaries of the thresholds.

Split EAL #4 into 2 separate EALs to differentiate between field survey readings and sample analysis results. This was done for human performance factors and will help to prevent the 2nd half of the EAL from getting missed.

DAEC does not have a perimeter rad monitoring system.

Last SER Approved EALs NEI 99-01 Rev. 5 EALs Proposed EALs Difference or Deviation Explanation CU1 RCS Leakage Operating Modes: Cold Shutdown CU1.1 Unidentified or pressure boundary leakage GREATER THAN 10 GPM.

OR CU1.2 Identified leakage GREATER THAN 25 GPM.

CU1 RCS Leakage Operating Modes: Cold Shutdown 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 condition will likely exceed the applicable time.

1.

RCS leakage results in the inability to maintain or restore RPV level greater than (site specific low level RPS actuation setpoint) for 15 minutes or longer.

CU1 RCS Leakage Operating Modes: Cold Shutdown 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.

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

DIFFERENCE Provided clarification on ERO positions associated with who could be Emergency Director, depending on which facility is in command and control. This is consistent throughout this recognition category.

170 inches is DAECs low-level RPS actuation setpoint CU2 Unplanned Loss of RCS Inventory with Irradiated Fuel in the RPV Operating Modes: Refueling CU2.1 Unplanned RCS level decrease LESS THAN the RPV flange for 15 minutes or longer.

OR CU2.2 RPV Level cannot be monitored AND Loss of RPV inventory as indicated by unexplained Drywell/Reactor Building Equipment or Floor Drain sump, or Torus, level increase.

CU2 Unplanned loss of RCS/RPV inventory.

Operating Modes: Refueling 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 condition will likely exceed the applicable time.

1.

UNPLANNED RCS/RPV level drop as indicated by either of the following:

RCS/RPV water level drop below the RPV flange for 15 minutes or longer when the RCS/RPV level band is established above the RPV flange.

RCS/RPV water level drop below the RCS level band for 15 minutes or longer when the RCS/RPV level band is established below the RPV flange.

OR

2.

RCS/RPV level cannot be monitored with a loss of RCS/RPV inventory as indicated by an unexplained level rise in (site specific sump or tank).

CU2 UNPLANNED loss of RCS/RPV inventory.

Operating Modes: Refueling 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.

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.

DIFFERENCE Clarified Emergency Director.

Split the 1st EAL into 2 for human factoring in making a more timely declaration. As currently written in NEI 99-01, it could cause the Operating crews to spend too much time in trying to determine exactly which part of the EAL was exceeded. Splitting into 2 separate EALS makes it easier for the crews to delineate between level bands and RPV flange similarities. As a result of splitting the EAL statement, changed either to ALL as another means of human factoring to quickly aid the operating crews that both of the following conditions need to be met. Also changed the order of the 2 conditions (level drop AND level band), This did not change the intent of either of the 2 new EALs.

Tank levels not normally monitored by Operations. For BWRs, the sumps are monitored via sump timers and run times, which are used to calculate rate. For BWRs, sump and Torus level increases would be potentially indicative of RCS level loss if no other method existed for monitoring RCS level.

Last SER Approved EALs NEI 99-01 Rev. 5 EALs Proposed EALs Difference or Deviation Explanation CU3 Loss of All Offsite Power to Essential Busses for Greater Than 15 Minutes Operating Modes: Cold Shutdown, Refueling CU3.1 Loss of power to or from the Startup or Standby Transformer resulting in a loss of all offsite power to Emergency Busses 1A3 and 1A4 that is expected to last for greater than 15 minutes AND At least one Emergency Bus, 1A3 or 1A4, is powered by its Standby Diesel Generator.

CU3 AC power capability to emergency busses reduced to a single power source for 15 minutes or longer such that any additional single failure would result in station blackout.

Operating Modes: Cold Shutdown, Refueling 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 condition will likely exceed the applicable time.

1.

a.

AC power capability to (site specific emergency busses) reduced to a single power source for 15 minutes or longer.

AND

b.

Any additional single power source failure will result in station blackout.

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.

Operating Modes: Cold Shutdown, Refueling 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.

CU3.1

a. AC power capability to 1A3 or 1A4 busses reduced to a single power source for 15 minutes or longer.

AND

b. Any additional single power source failure will result in station blackout.

DIFFERENCE Clarified Emergency Director.

Using essential in Initiating Condition vice emergency. This guidance is approved per NEI FAQ 2006-17. This is also consistent with the last SER Approved EAL.

Defined emergency busses as 1A3 and 1A4 CU4 Unplanned Loss of Decay Heat Removal Capability with Irradiated Fuel in the RPV Operating Modes: Cold Shutdown, Refueling CU4.1 An unplanned event results in RCS temperature GREATER THAN 212 ºF OR CU4.2 Loss of all RCS temperature and RPV level indication for GREATER THAN 15 minutes.

CU4 Unplanned loss of decay heat removal capability with irradiated fuel in the RPV.

Operating Modes: Cold Shutdown, Refueling 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 condition will likely exceed the applicable time.

1.

UNPLANNED event results in RCS temperature exceeding the Technical Specification cold shutdown temperature limit.

OR

2.

Loss of all RCS temperature and RCS/RPV level indication for 15 minutes or longer.

CU4 UNPLANNED loss of decay heat removal capability with irradiated fuel in the RPV.

Operating Modes: Cold Shutdown, Refueling 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.

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

DIFFERENCE Clarified Emergency Director.

GREATER THAN 212 ºF is equivalent to exceeding the Technical Specification cold shutdown temperature limit. This is DAECs Technical Specification limit.

CU6 Unplanned Loss of All Onsite or Offsite CU6 Loss of All On-site or Off-site Communications CU6 Loss of all onsite or offsite communications DIFFERENCE Consistent with current NEI EAL Included site-specific list of communication methods

Last SER Approved EALs NEI 99-01 Rev. 5 EALs Proposed EALs Difference or Deviation Explanation Communications Capabilities Operating Modes: Cold Shutdown, Refueling CU6.1 Loss of ALL of the following onsite communication capabilities affecting the ability to perform routine operation:

Plant Operations Radio System In-Plant Telephones Plant Paging System OR CU6.2 Loss of ALL of the following offsite communications capability:

All telephone lines (commercial)

Microwave Phone System FTS Phone System Capabilities Operating Modes: Cold Shutdown, Refueling, Defueled

1.

Loss of all of the following on-site communication methods affecting the ability to perform routine operations:

(site specific list of communications methods)

OR

2.

Loss of all of the following off-site communication methods affecting the ability to perform offsite notifications:

(site specific list of communications methods) capabilities Operating Modes: Cold Shutdown, Refueling, Defueled 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 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 Added satellite and cell phones to the list for offsite communication methods as well as ENS phones to reflect the intent in the basis CU7 Unplanned Loss of Required DC Power For Greater Than 15 Minutes Operating Modes: Cold Shutdown, Refueling CU7.1 Unplanned Loss of Vital DC power to required DC busses based on bus voltage LESS THAN 105 VDC indicated.

AND Failure to restore power to at least one required DC bus within 15 minutes from the time of loss.

CU7 Loss of required DC power for 15 minutes or longer.

Operating Modes: Cold Shutdown, Refueling 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 condition will likely exceed the applicable time.

1.

Less than (site specific bus voltage indication) on required (site specific Vital DC busses) for 15 minutes or longer.

CU7 Loss of required 125 VDC power for 15 minutes or longer.

Operating Modes: Cold Shutdown, Refueling 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.

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

DIFFERENCE Clarified Emergency Director.

Added 125 V in the IC preceding DC to quickly aid the operating crews to help eliminate any confusion about which DC power.

105 VDC is the site-specific setpoint for a loss of 125 VDC busses CU8 Inadvertent criticality.

Operating Modes: Cold Shutdown, Refueling CU8 Inadvertent criticality.

Operating Modes: Cold Shutdown, Refueling NONE

Last SER Approved EALs NEI 99-01 Rev. 5 EALs Proposed EALs Difference or Deviation Explanation CU8.1 An unplanned extended positive period observed on nuclear instrumentation.

1.

UNPLANNED sustained positive period observed on nuclear instrumentation.

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

CA1 Loss of RCS inventory Operating Modes: Cold Shutdown CA1.1 Loss of RCS inventory as indicated by RPV level LESS THAN 119.5 inches.

OR CA1.2 Loss of RCS inventory as indicated by unexplained Drywell/Reactor Building Equipment or Floor Drain sump, or Torus, level increase and RCS level cannot be monitored for GREATER THAN 15 minutes CA2 Loss of RPV Inventory with Irradiated Fuel in the RPV Operating Modes: Refueling CA2.1 Loss of RPV inventory as indicated by RPV level LESS THAN 119.5 inches.

OR CA2.2 Loss of RPV inventory as indicated by unexplained Drywell/Reactor Building Equipment or Floor Drain sump, or Torus, level increase and RPV level cannot be monitored for GREATER THAN 15 minutes CA1 Loss of RCS/RPV inventory Operating Modes: Cold Shutdown, Refueling 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 condition will likely exceed the applicable time.

1.

Loss of RCS/RPV inventory as indicated by level less than (site specific level).

[Low-Low ECCS actuation setpoint / Level 2 (BWR)]

OR

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 (site specific sump or tank).

CA1 Loss of RCS/RPV inventory Operating Modes: Cold Shutdown, Refueling 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.

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

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 level.

DIFFERENCE Clarified Emergency Director.

119.5 inches is DAECs low-low ECCS actuation setpoint Tank levels not normally monitored by Operations. For BWRs, the sumps are monitored via sump timers and run times, which are used to calculate rate. For BWRs, sump and Torus level increases would be potentially indicative of RCS level loss if no other method existed for monitoring RCS level.

CA3 Loss of All Offsite Power and Loss of All Onsite AC Power to Essential Busses Operating Modes: Cold S/D, Refueling, Defueled CA3.1 Loss of power to or from the Startup or Standby Transformer resulting in a loss of all offsite power to Emergency Busses 1A3 and 1A4 AND Failure of A Diesel Generator (1G-31) and B Diesel Generator (1G-21) to supply power to emergency busses 1A3 and CA3 Loss of all Off-site and all On-Site AC power to emergency busses for 15 minutes or longer.

Operating Modes: Cold S/D, Refueling, Defueled 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 condition will likely exceed the applicable time.

1.

Loss of all Off-Site and all On-Site AC Power to (site specific emergency busses) for 15 minutes or longer.

CA3 Loss of all offsite and all onsite AC power to Essential Busses for 15 minutes or longer.

Operating Modes: Cold S/D, Refueling, Defueled 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.

CA3.1 Loss of all offsite and all On-Site AC Power to 1A3 and 1A4 for 15 minutes or longer.

DIFFERENCE Clarified Emergency Director.

Using essential in Initiating Condition vice emergency. This guidance is approved per NEI FAQ 2006-17. This is also consistent with the last SER Approved EAL.

Defined emergency busses as 1A3 and 1A4

Last SER Approved EALs NEI 99-01 Rev. 5 EALs Proposed EALs Difference or Deviation Explanation 1A4.

AND Failure to restore power to at least one emergency bus, 1A3 or 1A4, within 15 minutes from the time of loss of both offsite and onsite AC power.

CA4 Inability to Maintain Plant in Cold Shutdown with Irradiated Fuel in the RPV Operating Modes: Cold S/D, Refueling 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 either RCS integrity not established or RCS inventory reduced, an unplanned event results in RCS temperature GREATER THAN 212 ºF for GREATER THAN 20 minutes. (Note: If an RCS heat removal system is in operation within this time frame and RCS temperature is being reduced then this EAL is not applicable.)

OR CA4.3 An unplanned event results in RCS temperature GREATER THAN 212 ºF for GREATER THAN 60 minutes or results in an RCS pressure increase of GREATER THAN 10 psig. (Note: If an RCS heat removal system is in operation within this time frame and RCS temperature is being reduced then this EAL is not applicable.)

CA4 Inability to maintain plant in cold shutdown.

Operating Modes: Cold S/D, Refueling

1.

An UNPLANNED event results in RCS temperature greater than (site specific Technical Specification cold shutdown temperature limit) for greater than the specified duration on table.

OR

2.

An UNPLANNED event results in RCS pressure increase greater than 10 psi due to a loss of RCS cooling. (PWR-This EAL does not apply in Solid Plant conditions.)

CA4 Inability to maintain plant in cold shutdown.

Operating Modes: Cold S/D, Refueling 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 temperature is 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 ºF for 60 minutes or longer. Note: If an RCS heat removal system is in operation within the specified time frames and RCS temperature is 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.

DIFFERENCE Removed PWR-specific annotations Retained same sequence as previous approved EAL scheme (Rev. 4)

Changed pressure increase to pressure rise for consistency throughout the EALs.

212 ºF is DAECs tech spec cold shutdown temperature limit Clarified Containment Closure as being Secondary Containment in accordance with NEI guidance for BWRs.

Changed psi to psig consistent with measuring instrumentation While using elements of the table in NEI 99-01, kept the EAL statements separate for each (similar to previous approved DAEC EALs).

Using Established versus Intact since the Basis discussion refers to RCS Integrity being established.

Split old CA4.3 into 2 separate EALs.

Included the note from the Table in the EAL.

See table next page. This will be included on the EAL wall board.

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

RCS Integrity Secondary Containment Duration temperature GREATER THAN 212 ºF EAL Not Established Not Established 0 minutes CA4.1 Not Established Established 20 minutes (*)

CA4.2 Established N/A 60 minutes (*)

CA4.3

(*) If an RCS heat removal system is in operation within the specified time frames and RCS temperature is being reduced, then this EAL is not applicable.

Last SER Approved EALs NEI 99-01 Rev. 5 EALs Proposed EALs Difference or Deviation Explanation CS1 Loss of RPV Inventory Affecting Core Decay Heat Removal Capability Operating Modes: Cold S/D CS1.1 With Secondary Containment not established:

a. RPV inventory as indicated by RPV level is LESS THAN 113.5 inches OR

b. RPV level cannot be monitored for GREATER THAN 30 minutes with a loss of RPV inventory as indicated by unexplained Drywell/Reactor Building Equipment or Floor Drain sump, or Torus, level increase OR CS1.2 With Secondary Containment established:

a. RPV inventory as indicated by RPV level is LESS THAN

+15 inches OR

b. RPV level cannot be monitored for GREATER THAN 30 minutes with a loss of RPV inventory as indicated by either:

Unexplained Drywell/Reactor Building Equipment or Floor Drain sump, or Torus, level increase Erratic Source Range Monitor Indication CS1 Loss of RCS/RPV inventory affecting core decay heat removal capability.

Operating Modes: Cold S/D, Refueling 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 condition will likely exceed the applicable time.

1.

With CONTAINMENT CLOSURE not established, RCS/RPV level less than (site specific level).

[6" below the bottom ID of the RCS loop (PWR)]

[6" below the low-low ECCS actuation setpoint (BWR)]

OR

2.

With CONTAINMENT CLOSURE established, RCS/RPV level less than (site specific level for TOAF).

OR

3.

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

(Site specific radiation monitor) reading greater than (site specific value).

Erratic Source Range Monitor Indication.

Unexplained level rise in (site specific sump or tank).

CS1 Loss of RCS/RPV inventory affecting core decay heat removal capability.

Operating Modes: Cold S/D, Refueling 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.

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

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

DIFFERENCE Clarified Emergency Director.

Used RPV versus RPV/RCS when showing specific indications which are shown as RPV level at DAEC.

113.5 inches is derived from subtracting 6 inches from the low-low ECCS actuation setpoint. At DAEC, the low-low ECCS actuation setpoint = 119.5 inches.

Consistent with current NEI EAL.

+15 inches is DAECs TOAF Clarified Containment Closure as being Secondary Containment in accordance with NEI guidance for BWRs.

Included Containment High Range Rad Monitor as site-specific instrument.

Tank levels not normally monitored by Operations.

For BWRs, the sumps are monitored via sump timers and run times, which are used to calculate rate. For BWRs, sump and Torus level increases would be potentially indicative of RCS level loss if no other method existed for monitoring RCS level.

Last SER Approved EALs NEI 99-01 Rev. 5 EALs Proposed EALs Difference or Deviation Explanation CS2 Loss of RPV Inventory Affecting Core Decay Heat Removal Capability with Irradiated Fuel in the RPV Operating Modes: Cold S/D CS2.1 With SECONDARY CONTAINMENT NOT ESTABLISHED, EITHER of the following occurs:

(a)

RPV inventory as indicated by RPV level is LESS THAN 113.5 inches (b)

RPV level cannot be monitored with Indication of core uncovery as evidenced by one or more of the following:

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

Erratic Source Range Monitor Indication OR CS2.2 With SECONDARY CONTAINMENT ESTABLISHED, EITHER of the following occurs:

(a)

RPV inventory as indicated by RPV level is LESS THAN +15 inches (b)

RPV level cannot be monitored with Indication of core uncovery as evidenced by one or more of the following:

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

Erratic Source Range Monitor Indication

Last SER Approved EALs NEI 99-01 Rev. 5 EALs Proposed EALs Difference or Deviation Explanation CG1 Loss of RPV Inventory Affecting Fuel Clad Integrity with Containment Challenged with Irradiated Fuel in the RPV Operating Modes: Cold S/D, Refueling CG1.1 (1)

Loss of RPV inventory as indicated by unexplained Drywell/Reactor Building Equipment or Floor Drain sump, or Torus, level increase AND (2)

RPV Level:

(a)

LESS THAN +15 inches for GREATER THAN 30 minutes OR (b)

Cannot be monitored with Indication of core uncovery for GREATER THAN 30 minutes as evidenced by one or more of the following:

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

Erratic Source Range Monitor Indication AND (3)

Indication of Secondary Containment challenged as indicated by one or more of the following:

Drywell Hydrogen or Torus Hydrogen GREATER THAN 6%

AND Drywell Oxygen or Torus Oxygen GREATER THAN 5%

Containment Pressure GREATER THAN 53 psig Secondary Containment not established Two or more Reactor Building areas exceed Max Safe Radiation Levels CG1 Loss of RCS/RPV inventory affecting fuel clad integrity with containment challenged.

Operating Modes: Cold S/D, Refueling 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 condition will likely exceed the applicable time.

1. a.

RCS/RPV level less than (site specific level for TOAF) for 30 minutes or longer.

AND

b.

ANY containment challenge indication (see Table):

OR

2. a.

RCS/RPV level cannot be monitored with core uncovery indicated by ANY of the following for 30 minutes or longer.

(Site specific radiation monitor) reading greater than (site specific setpoint).

Erratic source range monitor indication UNPLANNED level rise in (site specific sump or tank).

[Other site specific indications]

AND

b.

ANY containment challenge indication (see Table):

CG1 Loss of RCS/RPV inventory affecting fuel clad integrity with containment challenged.

Operating Modes: Cold S/D, Refueling 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.

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):

DIFFERENCE Clarified Emergency Director.

Used RPV versus RPV/RCS when showing specific indications which are shown as RPV level at DAEC.

Consistent with current NEI EAL.

+15 inches is DAECs TOAF.

Incorporated elements of the table into CG1.1b and CG1.2c Added with irradiated fuel in the RPV for clarification.

Clarified that Containment Closure is referring to secondary containment.

DAEC chose to define explosive mixture.

Split CG1.2a into an AND statement to be consistent with CG1.1a EAL CG1.2.a: Clarified the meaning of the 30 minutes to indicate core uncovery for 30 minutes or longer.

This is consistent with the basis description and does not change the intent of the EAL. This will also aid the key decision makers from a human factors standpoint to eliminate confusion about the intent of the 30 minutes, thereby helping their timeliness in declaration.

As water level in the RPV lowers, the dos 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:

Tank levels not normally monitored by Operations.

For BWRs, the sumps are monitored via sump timers and run times, which are used to calculate rate. For BWRs, sump and Torus level increases would be potentially indicative of RCS level loss if no other method existed for monitoring RCS level.

Calculation of Drywell Radiation Monitor Reading Assuming 1% Gap Release NG-88-0966 value 20% Gap Release at 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 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

Last SER Approved EALs NEI 99-01 Rev. 5 EALs Proposed EALs Difference or Deviation Explanation EU1 Damage To A Loaded Cask Confinement Boundary Operating Modes: Not applicable EU1.1 Any one of the following natural phenomena events with resultant visible damage to or loss of a loaded cask confinement boundary:

(4)

Report by plant personnel of a tornado strike.

(5)

Report by plant personnel of a seismic event.

OR EU1.2 The following accident condition with resultant visible damage to or loss of a loaded cask confinement boundary:

A loaded transfer cask is dropped as a result of normal handling or transporting.

OR EU1.3 Any condition in the opinion of the Emergency Director that indicates loss of loaded fuel storage cask confinement boundary.

E-HU1 Damage to a loaded cask confinement BOUNDARY.

Operating Modes: Not applicable

1. Damage to a loaded cask CONFINEMENT BOUNDARY.

EU1 Damage to a loaded cask CONFINEMENT BOUNDARY.

Operating Modes: Not applicable EU1.1 Damage to the Dry Shielded Canister of a loaded cask.

DIFFERENCE Used EU1 instead of E-HU1 for consistency on our wallboard. This is also consistent with last SER approved EAL.

At DAEC, the Dry Shielded Canister is the CONFINEMENT BOUNDARY (according to the definition in NEI 99-01 Rev. 5)

Last SER Approved EALs NEI 99-01 Rev. 5 EALs Proposed EALs Difference or Deviation Explanation FUEL CLAD BARRIER RADIATION/CORE DAMAGE L

Fuel damage assessment (PASAP 7.2) indicates at least 5% fuel clad damage OR L

Drywell Area Hi Range Rad Monitor, RIM-9184A or B reading GREATER THAN 7E+2 Rem/hr OR L

Torus Area Hi Range Rad Monitor, RIM-9185A or B reading GREATER THAN 3E+1 Rem/hr OR L

Coolant activity GREATER THAN 300Ci/gm DOSE EQUIVALENT I-131 FUEL CLAD BARRIER L

Primary containment radiation monitor reading greater than (site specific value).

OR L/P (Site specific ) as applicable OR L

Primary coolant activity greater than (site specific value).

FUEL CLAD BARRIER RADIATION/CORE DAMAGE L

Fuel damage assessment (PASAP 7.2) indicates at least 5% fuel clad damage OR L

Drywell Area Hi Range Rad Monitor, RIM-9184A or B reading GREATER THAN 700 Rem/hr OR L

Torus Area Hi Range Rad Monitor, RIM-9185A or B reading GREATER THAN 30 Rem/hr OR L

Coolant activity GREATER THAN 300 Ci/gm DOSE EQUIVALENT I-131 DIFFERENCE Consistent with last SER Approved EAL and current NEI EAL.

Site-specific for DAEC includes the fuel damage assessment and Torus Area High Range Monitor indication as per original SER. PASAP 7.2 is the procedure used for fuel damage assessment.

Included site-specific values NOTE: Fission Barrier Table logic is as follows for all Fission Barrier EALs. This logic diagram is on DAECs EAL Wall Boards.

CLAD LOSS OF AT LEAST 2 BARRIERS?

L P RCS L P CNTMT L P CLAD L P RCS L P CNTMT L P CLAD L P RCS L P CNTMT L P 1/2 1/1 FU1 UNUSUAL EVENT FA1 ALERT ONE BARRIER AFFECTED TWO BARRIERS AFFECTED THREE BARRIERS AFFECTED 2/3 FS1 SITE AREA EMERGENCY 3/3 FG1 GENERAL EMERGENCY NO YES

Last SER Approved EALs NEI 99-01 Rev. 5 EALs Proposed EALs Difference or Deviation Explanation FUEL CLAD BARRIER RPV LEVEL L

RPV Level LESS THAN -25 Inches P

RPV Level LESS THAN +15 inches FUEL CLAD BARRIER L

RPV water level cannot be restored and maintained above (site specific RPV water level corresponding to the requirement for primary containment flooding)

P RPV water level cannot be restored and maintained above (site specific RPV water level corresponding to the top of active fuel) or cannot be determined.

FUEL CLAD BARRIER RPV LEVEL L

RPV Level cannot be restored and maintained above -25 Inches P

RPV Level cannot be restored and maintained above +15 inches or cannot be determined DIFFERENCE Consistent with last SER Approved EAL and current NEI EAL.

Stating RPV Level so Operators know what level to refer to.

-25 inches is DAECs Minimum Steam Cooling RPV Water Level.

+15 inches is DAECs TOAF.

FUEL CLAD BARRIER EMERGENCY DIRECTOR JUDGMENT Any condition in the opinion of the Emergency Director that indicates Loss or Potential Loss of the Fuel Clad Barrier.

FUEL CLAD BARRIER EMERGENCY DIRECTOR JUDGMENT L

Any condition in the opinion of the Emergency Director that indicates Loss of the Fuel Clad Barrier P

Any condition in the opinion of the Emergency Director that indicates Potential Loss of the Fuel Clad Barrier FUEL CLAD BARRIER EMERGENCY DIRECTOR JUDGMENT L

Any condition in the opinion of the Emergency Director (OSM, EC or ER&RD) that indicates Loss of the Fuel Clad Barrier P

Any condition in the opinion of the Emergency Director (OSM, EC or ER&RD) that indicates Potential Loss of the Fuel Clad Barrier DIFFERENCE Provided clarification on ERO positions associated with who could be Emergency Director, depending on which facility is in command and control. This is consistent throughout this recognition category.

RCS BARRIER RADIATION/CORE DAMAGE L

Drywell Area Hi Range Rad Monitor, RIM-9184A or B reading GREATER THAN 5 Rem/hr after reactor shutdown RCS BARRIER L

Primary containment radiation monitor reading greater than (site specific value).

OR L/P (Site specific ) as applicable RCS BARRIER RADIATION/CORE DAMAGE L

Drywell Area Hi Range Rad Monitor, RIM-9184A or B reading GREATER THAN 5 Rem/hr after reactor shutdown DIFFERENCE Listed specific equipment ID and name. Drywell rad monitor is DAECs equivalent to Primary Containment Radiation Monitor after reactor shutdown: 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.

Approved via last SER.

RCS BARRIER RPV LEVEL L

RPV Level LESS THAN +15 inches RCS BARRIER L

RPV water level cannot be restored and maintained above (site specific RPV water level corresponding to the top of active fuel) or cannot be determined.

RCS BARRIER RPV LEVEL L

RPV Level cannot be restored and maintained above +15 inches or cannot be determined DIFFERENCE Stating RPV Level so Operators know what level to refer to.

Consistent with last SER Approved EAL and current NEI EAL.

+15 inches is DAECs TOAF.

Last SER Approved EALs NEI 99-01 Rev. 5 EALs Proposed EALs Difference or Deviation Explanation RCS BARRIER LEAKAGE L

Unisolable Main Steamline Break as indicated by the failure of both MSIVs in any one line to close AND EITHER:

High MSL flow or high steam tunnel temperature annunciators Direct report of steam release P

RCS Leakage is GREATER THAN 50 GPM inside the drywell OR P

Unisolable primary system leakage outside the drywell as indicated by area temps or ARMs exceeding the Max Normal Limits per EOP 3, Table 6.

RCS BARRIER L

(site specific Indication of an UNISOLABLE Main Steamline, HPCI, Feedwater, RWCU, or RCIC break)

OR L

Emergency RPV Depressurization is required.

P RCS leakage GREATER THAN 50 gpm inside the drywell OR P

UNISOLABLE primary system leakage outside primary containment as indicated by exceeding EITHER of the following:

Max Normal Operating Temperature.

OR Max Normal Area Radiation RCS BARRIER LEAKAGE L

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 L

Emergency RPV Depressurization is required.

P RCS Leakage is GREATER THAN 50 GPM inside the drywell OR P

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

DIFFERENCE Added site-specific indication of an unisolable main steamline leak which includes failure of both isolation valves.

Consistent with current NEI EAL.

Referred to drywell vice primary containment to help with understanding of location of the leak and eliminate confusion.

Added the specific reference for Max Normal, no change in intent.

Max Normal Limits per EOP 3 are the values used to determine a Potential Loss of an RCS Barrier. Max Safe Limits per EOP 3 are the values used to determine a Loss of a Primary Containment Barrier.

RCS BARRIER PRIMARY CONTAINMENT ATMOSPHERE L

Drywell pressure GREATER THAN 2 psig and not caused by a loss of DW Cooling RCS BARRIER L

Primary containment pressure GREATER THAN (site-specific value) due to RCS leakage.

RCS BARRIER PRIMARY CONTAINMENT ATMOSPHERE L

Drywell pressure GREATER THAN 2 psig due to RCS Leakage (not caused by a loss of Drywell Cooling)

DIFFERENCE Consistent with last SER Approved EAL and current NEI EAL.

DAEC uses a GE Mark I Containment (drywell). 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.

Approved via last SER.

RCS BARRIER EMERGENCY DIRECTOR JUDGMENT Any condition in the opinion of the Emergency Director that indicates Loss or Potential Loss of the RCS Barrier RCS BARRIER EMERGENCY DIRECTOR JUDGMENT L

Any condition in the opinion of the Emergency Director that indicates Loss of the RCS Barrier P

Any condition in the opinion of the Emergency Director that indicates Potential Loss of the RCS Barrier RCS BARRIER EMERGENCY DIRECTOR JUDGMENT L

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

Any condition in the opinion of the Emergency Director (OSM, EC or ER&RD) that indicates Potential Loss of the RCS Barrier DIFFERENCE Provided clarification on ERO positions associated with who could be Emergency Director, depending on which facility is in command and control. This is consistent throughout this recognition category.

Last SER Approved EALs NEI 99-01 Rev. 5 EALs Proposed EALs Difference or Deviation Explanation PRIMARY CONTAINMENT BARRIER RADIATION/CORE DAMAGE P

Drywell Area Hi Range Rad Monitor, RIM-9184A or B reading GREATER THAN 3E+3 Rem/hr OR P

Torus Area Hi Range Rad Monitor, RIM-9185A or B reading GREATER THAN 1E+2 Rem/hr OR P

Fuel damage assessment (PASAP 7.2) indicates at least 20% fuel clad damage CONTAINMENT BARRIER P

Primary containment radiation monitor reading greater than (site specific value).

OR P/L (Site specific) as applicable PRIMARY CONTAINMENT BARRIER RADIATION/CORE DAMAGE P

Drywell Area Hi Range Rad Monitor, RIM-9184A or B reading GREATER THAN 3000 Rem/hr OR P

Torus Area Hi Range Rad Monitor, RIM-9185A or B reading GREATER THAN 100 Rem/hr OR P

Fuel damage assessment (PASAP 7.2) indicates at least 20% fuel clad damage DIFFERENCE Included site-specific instrumentation and values similar to previous EALs.

Site-specific for DAEC includes the fuel damage assessment and Torus Area High Range Monitor indication as per original SER. PASAP 7.2 is the procedure used for fuel damage assessment.

PRIMARY CONTAINMENT BARRIER RPV LEVEL P

Primary containment flooding required CONTAINMENT BARRIER P

Primary containment flooding required PRIMARY CONTAINMENT BARRIER RPV LEVEL P

Primary containment flooding required NONE PRIMARY CONTAINMENT BARRIER LEAKAGE L

Failure of both valves in any one line to close and a downstream pathway to the environment exists OR L

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 L

Primary containment venting per EOPs CONTAINMENT BARRIER L

Failure of all valves in any one line to close AND direct downstream pathway to the environment exists after primary containment isolation signal.

OR L

Intentional primary containment venting per EOPs OR L

UNISOLABLE primary system leakage outside primary containment as indicated by exceeding EITHER of the following:

Max Safe Operating Temperature.

OR Max Safe Area Radiation PRIMARY CONTAINMENT BARRIER LEAKAGE L

Failure of both isolation valves in any one line to close AND direct downstream pathway to the environment exists after primary containment isolation signal.

OR L

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

OR L

Intentional Primary Containment venting per EOPs DIFFERENCE Consistent with last SER Approved EAL and current NEI EAL.

Changed all valves to both isolation valves for ease in understanding to the Operating crews and decision-makers. This helps to meet the intent of a pathway from primary containment to the environment.

Referred to drywell vice primary containment to help with understanding of location of the leak and eliminate confusion.

Added the specific reference for Max Normal, no change in intent.

Max Normal Limits per EOP 3 are the values used to determine a Potential Loss of an RCS Barrier. Max Safe Limits per EOP 3 are the values used to determine a Loss of a Primary Containment Barrier.

Adding when Containment Isolation is required is in accordance with the basis for this EAL and is added to the IC to ensure consistent understanding of when this IC is applicable.

Last SER Approved EALs NEI 99-01 Rev. 5 EALs Proposed EALs Difference or Deviation Explanation PRIMARY CONTAINMENT BARRIER PRIMARY CONTAINMENT ATMOSPHERE L

Rapid unexplained decrease following initial increase in pressure OR L

Drywell pressure response not consistent with LOCA conditions P

Torus pressure reaches 53 psig and increasing OR P

Drywell or Torus H2 cannot be determined to be LESS THAN 6% and Drywell or Torus O2 cannot be determined to be LESS THAN 5%

CONTAINMENT BARRIER L

Primary containment pressure rise followed by a rapid unexplained drop in primary containment pressure.

OR L

Primary containment pressure response not consistent with LOCA conditions.

P Primary containment pressure greater than (site specific value) and rising.

OR P

Explosive mixture exists inside primary containment.

OR P

RPV pressure and suppression pool temperature cannot be maintained below the HCTL.

PRIMARY CONTAINMENT BARRIER PRIMARY CONTAINMENT ATMOSPHERE L

Drywell pressure rise followed by a rapid unexplained drop in Drywell pressure.

OR L

Drywell pressure response NOT consistent with LOCA conditions P

Torus pressure reaches 53 psig and rising OR P

Drywell or Torus H2 cannot be determined to be LESS THAN 6% and Drywell or Torus O2 cannot be determined to be LESS THAN 5%

OR P

RPV pressure and Torus water temperature cannot be maintained below the Heat Capacity Limit (EOP Graph 4)

DIFFERENCE Consistent with last SER Approved EAL and current NEI EAL.

Used drywell vice primary containment to be consistent.

Added in pressure to ensure clarity.

DAEC uses Torus pressure of 53 psig as its Primary Containment Pressure Limit as used in the EOPs.

DAEC chose to define explosive mixture.

DAEC uses Torus to be synonymous with suppression pool Identified the specific EOP graph for HCL (HCTL) to quickly draw Operator attention to the specific graph.

PRIMARY CONTAINMENT BARRIER EMERGENCY DIRECTOR JUDGMENT Any condition in the opinion of the Emergency Director that indicates Loss or Potential Loss of the Containment Barrier CONTAINMENT BARRIER EMERGENCY DIRECTOR JUDGMENT L

Any condition in the opinion of the Emergency Director that indicates Loss of the Containment Barrier P

Any condition in the opinion of the Emergency Director that indicates Potential Loss of the Containment Barrier PRIMARY CONTAINMENT BARRIER EMERGENCY DIRECTOR JUDGMENT L

Any condition in the opinion of the Emergency Director (OSM, EC or ER&RD) that indicates Loss of the Primary Containment Barrier P

Any condition in the opinion of the Emergency Director (OSM, EC or ER&RD) that indicates Potential Loss of the Primary Containment Barrier DIFFERENCE Provided clarification on ERO positions associated with who could be Emergency Director, depending on which facility is in command and control. This is consistent throughout this recognition category.

Consistent with last SER Approved EAL and current NEI EAL.

Last SER Approved EALs NEI 99-01 Rev. 5 EALs Proposed EALs Difference or Deviation Explanation HU1 Natural and Destructive Phenomena Affecting the Protected Area Operating Modes: All HU1.1 Earthquake detected per AOP 901, Earthquake OR HU1.2 Report of a tornado touching down within the Plant Protected Area with NO confirmed damage to a Safe Shutdown/Vital Area or Control Room indication of degraded performance of a System of Concern.

OR HU1.3 Report of winds greater than 95 mph within the Plant Protected Area with NO confirmed damage to a Safe Shutdown/Vital Area or Control Room indication of degraded performance of a System of Concern.

OR HU1.4 Vehicle crash into plant structures or systems within the Plant Protected Area with NO confirmed damage to a Safe Shutdown/Vital Area or Control Room indication of degraded performance of a System of Concern.

OR HU1.5 Report of an unanticipated explosion within the Plant Protected Area resulting in visible damage to permanent structures or equipment.

OR HU1.6 Report of turbine failure resulting in casing penetration or damage to turbine or generator seals.

OR HU1.7 River level above 757 feet.

OR HU1.8 Uncontrolled flooding in a Safe Shutdown/Vital Area that has the potential to affect safety related equipment needed for the current operating mode.

OR HU1.9 River level below 725 feet 6 inches.

HU1 Natural or destructive phenomena affecting the PROTECTED AREA.

Operating Modes: All

1.

Seismic event identified by ANY 2 of the following:

o Seismic event confirmed by (site specific indication or method) o Earthquake felt in plant o

National Earthquake Center

2.

Tornado striking within PROTECTED AREA boundary or high winds greater than (site specific mph).

3.

Internal flooding that has the potential to affect safety related equipment required by Technical Specifications for the current operating mode in ANY of the following areas:

(site specific area list)

4.

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

5.

(Site specific occurrences affecting the PROTECTED AREA).

HU1 Natural or destructive phenomena affecting the PROTECTED AREA.

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

Seismic event confirmed per AOP 901, Earthquake Earthquake felt in plant 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.

DEVIATION Consistent with last SER Approved EAL and current NEI EAL.

AOP 901, Earthquake is the procedure used to validate a seismic event. Included phone number to NEIC.

The tornado and high winds EAL was split into two separate EALs for human factors. Having multiple items to consider in one statement is a human performance trap in that the reader may focus on the first one and miss the second one. Splitting this into two EALs still meets the intent of the original EAL.

Added switchyard to ICs with Protected Area.

DEVIATION - Used on-site for high winds because the Met Tower which records wind speed is outside the Protected Area.

Added with NO confirmed damage to several ICs to clarify that this is a NOUE and to be consistent with NEI bases.

Added reference to the Safe Shutdown/Vital Areas Table and Systems of Concern Table in the applicable ICs to ensure consistent understanding of what area and/or systems we are concerned about. These tables are also included on the EAL Wall Boards and in the bases.

Included river level flooding which would affect the PA.

Deleted previous EAL HU1.9 for low river water level. It is an ALERT now.

Safe Shutdown/Vital 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 Sink /

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

Last SER Approved EALs NEI 99-01 Rev. 5 EALs Proposed EALs Difference or Deviation Explanation HU2 Fire Within Protected Area Boundary Not Extinguished Within 15 Minutes of Detection Operating Modes: ALL HU2.1 Fire in buildings or areas contiguous to any Safe Shutdown/Vital Area not extinguished within 15 minutes of control room notification or verification of a control room alarm.

HU2 FIRE within the PROTECTED AREA not extinguished within 15 minutes of detection or EXPLOSION within the PROTECTED AREA.

Operating Modes: ALL 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.

1.

FIRE not extinguished within 15 minutes of control room notification or verification of a control room FIRE alarm in ANY of the following areas:

(site specific area list)

OR

2.

EXPLOSION within the PROTECTED AREA.

HU2 FIRE within the PROTECTED AREA NOT extinguished within 15 minutes of detection or EXPLOSION within the PROTECTED AREA.

Operating Modes: ALL 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 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 Shutdown/Vital Area.

HU2.2.

EXPLOSION within the PROTECTED AREA.

DIFFERENCE Consistent with current NEI EAL.

Provided clarification on ERO positions associated with who could be Emergency Director, depending on which facility is in command and control. This is consistent throughout this recognition category.

Areas of concern are the Safe Shutdown/Vital Areas.

Chose to define a single source of affected areas for EALs as a human factors improvement opportunity. By using the Safe Shutdown/Vital Areas table as a consistent reference for applicable EALs, EAL decision-makers do not need to waste time referring to other tables or lists of impacted areas. This is a conservative decision that has been made to aid in timely EAL decision making.

HU3 Release of Toxic or Flammable Gases Deemed Detrimental to Normal Operation of the Plant.

Operating Modes: ALL HU3.1 Report or detection of toxic or flammable gases that has or could enter the site area boundary in amounts that can affect normal plant operations.

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

HU3 Release of toxic, corrosive, asphyxiant, or flammable gases deemed detrimental to NORMAL PLANT OPERATIONS.

Operating Modes:

All

1.

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

OR

2.

Report by local, county or state officials for evacuation or sheltering of site personnel based on an off-site event.

HU3 Release of toxic, corrosive, asphyxiant, or flammable gases deemed detrimental to NORMAL PLANT OPERATIONS.

Operating Modes:

All 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.

NONE HU5 Other Conditions Existing Which in the Judgment of the Emergency Director Warrant Declaration of a NOUE.

Operating Modes: ALL HU5.1 Other conditions exist which in the judgment HU5 Other conditions exist which in the judgment of the Emergency Director warrant declaration of a NOUE.

Operating Modes: ALL

1.

Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or have occurred which HU5 Other conditions exist which in the judgment of the Emergency Director (OSM, EC or ER&RD) warrant declaration of a NOUE.

Operating Modes: ALL HU5.1. Other conditions exist which in the judgment of the Emergency Director (OSM, EC or ER&RD) indicate that events are in progress DIFFERENCE Provided clarification on ERO positions associated with who could be Emergency Director, depending on which facility is in command and control. This is consistent throughout this recognition category.

Last SER Approved EALs NEI 99-01 Rev. 5 EALs Proposed EALs Difference or Deviation Explanation of the Emergency Director indicate that events are in process or have occurred which indicate a potential degradation of the level of safety of the plant. No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of safety systems occurs.

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.

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.

Last SER Approved EALs NEI 99-01 Rev. 5 EALs Proposed EALs Difference or Deviation Explanation HA1 Natural and Destructive Phenomena Affecting the Plant Vital Area Operating Modes: ALL HA1.1 Receipt of the Amber Operating Basis Earthquake Light and the wailing seismic alarm on 1C35 ( 0.06 gravity).

OR HA1.2 Report of Tornado or high winds greater than 95MPH within PROTECTED AREA boundary and resulting in VISIBLE DAMAGE to a Safe Shutdown/Vital Area or Control Room indication of degraded performance of a System of Concern.

OR HA1.3 Vehicle crash within PROTECTED AREA boundary and resulting in VISIBLE DAMAGE to a Safe Shutdown/Vital Area or Control Room indication of degraded performance of a System of Concern.

OR HA1.4 Turbine failure-generated missiles result in any VISIBLE DAMAGE to or penetration of any of a Safe Shutdown/Vital Area.

OR HA1.5 River level ABOVE 767 feet.

OR HA1.6 Uncontrolled flooding in a Safe Shutdown/Vital Area that results in degraded safety system performance as indicated in the Control Room or that creates an industrial safety hazards (e.g., electric shock) that precludes access necessary to operate or monitor safety equipment.

OR HA1.7 River level BELOW 724 feet 6 inches.

OR HA1.8 Report to control room of visible damage affecting a Safe Shutdown/Vital Area.

HA1 Natural or destructive phenomena affecting VITAL AREAS.

Operating Modes: ALL

1.

a.

Seismic event greater than Operating Basis Earthquake (OBE) as indicated by (site specific seismic instrumentation) reading (site specific OBE limit).

AND

b.

Earthquake confirmed by ANY of the following:

Earthquake felt in plant National Earthquake Center Control Room indication of degraded performance of systems required for the safe shutdown of the plant.

OR

2.

Tornado striking or high winds greater than (site specific mph) resulting in VISIBLE DAMAGE to ANY of the following structures containing safety systems or components OR control room indication of degraded performance of those safety systems:

(site specific structure list)

OR

3.

Internal flooding in ANY of the following areas resulting in an electrical shock hazard that precludes access to operate or monitor safety equipment OR control room indication of degraded performance of those safety systems:

(site specific area list)

OR

4.

Turbine failure-generated PROJECTILES resulting in VISIBLE DAMAGE to or penetration of ANY of the following structures containing safety systems or components OR control room indication of degraded performance of those safety systems:

HA1 Natural or destructive phenomena affecting VITAL AREAS.

Operating Modes: ALL 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:

Earthquake felt in plant 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 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 DIFFERENCE Consistent with last SER Approved EAL and current NEI EAL.

Changed the order to resemble more like HU1 AOP 901, Earthquake is the procedure used to validate a seismic event. Included phone number to NEIC.

Combined Tornado, High Winds and Vehicle crash into 1 EAL since the structures that could be impacted are similar, but not inclusive of the entire Safe Shutdown/Vital Area table.

Added reference to the Safe Shutdown/Vital Areas Table and Systems of Concern Table in the applicable ICs to ensure consistent understanding of what area and/or systems we are concerned about. These tables are also included on the EAL Wall Boards and in the bases.

Split HA1.2, HA1.3 and HA1.4 into a more discernible OR statement for the decision makers.

Included river level flooding which would affect the Vital Area.

Changed previous EAL HA1.7 for low river water level to supply pit low level as this is what would alarm in the control room for low flow or a blockage. Then moved this EAL near the high river water level EAL.

Maintained previous HA1.8 (now HA1.5) to include any other reports to the control room of visible damage not covered by the EALs.

The same Safe Shutdown/Vital Areas and Systems of Concern as HU1, for applicable EALs

Last SER Approved EALs NEI 99-01 Rev. 5 EALs Proposed EALs Difference or Deviation Explanation (site specific structure list)

OR

5.

Vehicle crash resulting in VISIBLE DAMAGE to ANY of the following structures containing safety systems or components OR control room indication of degraded performance of those safety systems:

(site specific structure list)

OR

6.

(Site specific occurrences) resulting in VISIBLE DAMAGE to ANY of the following structures containing safety systems or components OR control room indication of degraded performance of those safety systems:

(site specific structure list) o Reactor Building o

Condensate Storage Tank Area OR 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.

HA2 Fire or Explosion Affecting the Operability of Plant Safety Systems Required to Establish or Maintain Safe Shutdown Operating Modes: ALL HA2.1 Fire or explosion in any Safe Shutdown/Vital Area AND Affected system parameter indications show degraded performance or plant personnel report VISIBLE DAMAGE to permanent structures or equipment within the specified area.

HA2 FIRE or EXPLOSION affecting the operability of plant safety systems required to establish or maintain safe shutdown.

Operating Modes: ALL

1.

FIRE or EXPLOSION resulting in VISIBLE DAMAGE to ANY of the following structures containing safety systems or components OR control room indication of degraded performance of those safety systems:

(site specific structure list)

HA2 FIRE or EXPLOSION affecting the operability of plant safety systems required to establish or maintain safe shutdown.

Operating Modes: ALL HA2.1 FIRE or EXPLOSION resulting in VISIBLE DAMAGE to ANY Safe Shutdown/Vital Area or Control Room indication of degraded performance of a System of Concern.

DIFFERENCE Consistent with current NEI EAL.

Areas of concern are the Safe Shutdown/Vital Areas.

Chose to define a single source of affected areas for EALs as a human factors improvement opportunity. By using the Safe Shutdown/Vital Areas table as a consistent reference for applicable EALs, EAL decision-makers do not need to waste time referring to other tables or lists of impacted areas. This is a conservative decision that has been made to aid in timely EAL decision making.

HA3 Release of Toxic or Flammable Gases Within or Contiguous to a Vital Area Which Jeopardizes Operation of Systems Required to Maintain Safe Operations or Establish or Maintain Safe Shutdown.

Operating Modes: ALL HA3 Access to a VITAL AREA is prohibited due to toxic, corrosive, asphyxiant or flammable gases which jeopardize operation of operable equipment required to maintain safe operations or safely shutdown the reactor.

Operating Modes: ALL DELETED DIFFERENCE Operations in conjunction with Emergency Preparedness reviewed plant procedures associated with shutdown and cooldown. This review consisted of identifying those local actions required by Operators in areas requiring access for normal plant shutdown and cooldown.

No required local Operator actions, outside of the Control Room, were identified assuming no equipment failure, no

Last SER Approved EALs NEI 99-01 Rev. 5 EALs Proposed EALs Difference or Deviation Explanation HA3.1 Report or detection of toxic gases within or contiguous to a Safe Shutdown/Vital Area in concentrations that may result in an atmosphere Immediately Dangerous to Life and Health (IDLH).

OR HA3.2 Report or detection of gases in concentration greater than the Lower Flammability Limit within or contiguous to a Safe Shutdown/Vital Area.

Note: If the equipment in the stated area was already inoperable, or out of service, before the event occurred, then this EAL should not be declared as 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.

1.

Access to a VITAL AREA is prohibited due to toxic, corrosive, asphyxiant or flammable gases which jeopardize operation of systems required to maintain safe operations or safely shutdown the reactor.

coincident events and no accidents. Specific to this review, were the consideration of Station Blackout required actions and local action required for initiation of shutdown cooling.

Station Blackout required actions were reviewed and considered, however a Station Blackout is considered a coincident event and therefore does not apply to this EAL. In addition, Station Blackout is covered by the System Malfunction EALs SS1.1 and SG1.1.

Also reviewed local manual operation of the Low Pressure Cooling Injection (LPCI) system crosstie during Mode 3 for initiation of Shutdown Cooling.

Since LPCI is inop/removed from service as a result of this evolution, this also would not apply to this EAL when considering NEI guidance:

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 declared as 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.

As a result, DAEC is planning on removing EAL (HA3.1) from the scheme.

HA5 Control Room Evacuation Has Been Initiated Operating Modes: ALL HA5.1 Entry into AOP 915 for control room evacuation.

HA5 Control room evacuation has been initiated.

Operating Modes: ALL

1.

(Site-specific procedure) requires control room evacuation.

HA5 Control room evacuation has been initiated.

Operating Modes: ALL HA5.1 AOP 915, Shutdown Outside Control Room, entered requiring Control Room evacuation.

DIFFERENCE Included site-specific AOP for shutdown outside control room to help direct the Operating crews more quickly to the appropriate procedure and aid in a timely declaration.

HA6 Other Conditions Existing Which in the Judgment of the Emergency Director Warrant Declaration of an Alert.

Operating Modes: ALL HA6.1 Other conditions exist which in the judgment of the Emergency Director indicate that events are in process or have occurred which involve actual or likely potential substantial degradation of the level of safety of the plant.

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

HA6 Other conditions exist which in the judgment of the Emergency Director warrant declaration of an Alert.

Operating Modes: ALL

1.

Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant or a security event that involves probable life threatening risk to site personnel or damage to site equipment because of HOSTILE HA6 Other conditions exist which in the judgment of the Emergency Director (OSM, EC or ER&RD) warrant declaration of an Alert.

Operating Modes: ALL 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 DIFFERENCE Provided clarification on ERO positions associated with who could be Emergency Director, depending on which facility is in command and control. This is consistent throughout this recognition category.

Last SER Approved EALs NEI 99-01 Rev. 5 EALs Proposed EALs Difference or Deviation Explanation ACTION. Any releases are expected to be limited to small fractions of the EPA Protective Action Guideline exposure levels.

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

HS2 Control Room Evacuation Has Been Initiated and Plant Control Cannot Be Established Operating Modes: ALL HS2.1 Control Room evacuation has been initiated.

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

HS2 Control room evacuation has been initiated and plant control cannot be established.

Operating Modes: ALL

1.

a.

Control room evacuation has been initiated.

AND

b.

Control of the plant cannot be established within (site specific minutes).

HS2 Control room evacuation has been initiated and plant control cannot be established.

Operating Modes: ALL HS2.1. Control Room evacuation has been initiated.

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

DIFFERENCE Included site-specific AOP for shutdown outside control room.

Included site-specific timeframe of 20 minutes for establishing control of the plant.

HS3 Other Conditions Existing Which in the Judgment of the Emergency Director Warrant Declaration of Site Area Emergency.

Operating Modes: ALL HS3.1 Other conditions exist which in the judgment of the Emergency Director indicate that events are in process or have occurred which involve actual or likely major failures of plant functions needed for 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.

HS3 Other conditions exist which in the judgment of the Emergency Director warrant declaration of a Site Area Emergency.

Operating Modes: ALL

1.

Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or have occurred which involve actual or likely major failures of plant functions needed for protection of the public or HOSTILE ACTION that results in intentional damage or malicious acts; (1) toward site personnel or equipment that could lead to the likely failure of or; (2) that prevent effective access to equipment needed for the protection of the public. Any releases are not expected to result in exposure levels which exceed EPA Protective Action Guideline exposure levels beyond the site boundary.

HS3 Other conditions exist which in the judgment of the Emergency Director (OSM, EC or ER&RD) warrant declaration of a Site Area Emergency.

Operating Modes: ALL 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.

DIFFERENCE Provided clarification on ERO positions associated with who could be Emergency Director, depending on which facility is in command and control. This is consistent throughout this recognition category.

HG2 Other Conditions Existing Which in the Judgment of the Emergency Director Warrant Declaration of General Emergency.

Operating Modes: ALL HG2.1 Other conditions exist which in the judgment of the Emergency Director indicate that events are in process or have occurred which involve HG2 Other conditions exist which in the judgment of the Emergency Director warrant declaration of a General Emergency.

Operating Modes: ALL

1.

Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or have occurred which HG2 Other conditions exist which in the judgment of the Emergency Director (OSM, EC or ER&RD) warrant declaration of a General Emergency.

Operating Modes: ALL HG2.1. Other conditions exist which in the judgment of the Emergency Director (OSM, EC or ER&RD) indicate that events are in progress DIFFERENCE Provided clarification on ERO positions associated with who could be Emergency Director, depending on which facility is in command and control. This is consistent throughout this recognition category.

Last SER Approved EALs NEI 99-01 Rev. 5 EALs Proposed EALs Difference or Deviation Explanation actual or imminent substantial core degradation or melting with potential for loss of containment integrity. Releases can be reasonably expected to exceed EPA Protective Action Guideline exposure levels offsite for more than the immediate site area.

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.

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.

Last SER Approved EALs NEI 99-01 Rev. 5 EALs Proposed EALs Difference or Deviation Explanation SU1 Loss of All Offsite Power to Essential Busses for Greater Than 15 Minutes Operating Modes: Power Operation, Startup, Hot S/D SU1.1 Loss of power to or from the Startup or Standby Transformer resulting in a loss of all offsite power to Emergency Busses 1A3 and 1A4 that is expected to last for greater than 15 minutes.

AND Emergency Busses 1A3 and 1A4 are powered by their respective Standby Diesel Generators.

SU1 Loss of All Offsite AC Power to emergency busses for 15 Minutes or longer.

Operating Modes: Power Operation, Startup, Hot Standby, Hot Shutdown 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 condition has exceeded, or will likely exceed, the applicable time.

1.

Loss of all off-site AC power to (site specific emergency busses) for 15 minutes or longer.

SU1 Loss of all offsite AC power to Essential Busses for 15 Minutes or longer.

Operating Modes: Power Operation, Startup, Hot Shutdown 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.

DIFFERENCE Using essential in Initiating Condition vice emergency. This guidance is approved per NEI FAQ 2006-17. This is also consistent with the last SER Approved EAL.

Consistent with current NEI EAL.

Provided clarification on ERO positions associated with who could be Emergency Director, depending on which facility is in command and control. This is consistent throughout this recognition category.

DAEC states the emergency busses (site-specific information) powered by the applicable transformers to minimize any potential confusion whereby the transformer would have power, but the essential bus would not. The intent of the EAL is a loss of offsite power to the essential busses and each essential bus is getting power from its applicable Standby Diesel Generator.

1A3 and 1A4 is equivalent to DAECs Essential Busses throughout this recognition category.

SU2 Inability to Reach Required Shutdown Within Technical Specification Limits Operating Modes: Power Operation, Startup, Hot S/D SU2.1 Plant is not brought to required operating mode within applicable Technical Specifications LCO Action Statement Time.

SU2 Inability to reach required shutdown within Technical Specification limits.

Operating Modes: Power Operation, Startup, Hot Standby, Hot Shutdown Plant is not brought to required operating mode within Technical Specifications LCO Action Statement Time.

SU2 Inability to reach required shutdown within Technical Specification limits Operating Modes: Power Operation, Startup, Hot Shutdown SU2.1 Plant is not brought to required operating mode within applicable Technical Specifications LCO Action Statement Time.

NONE Consistent with last SER Approved EAL and current NEI EAL.

SU3 Unplanned Loss of Most or All Safety System Annunciation or Indication in the Control Room for Greater Than 15 Minutes Operating Modes: Power Operation, Startup, Hot S/D SU3.1 Unplanned loss of most or all 1C03, 1C04 and 1C05 annunciators or indicators associated with Safety Systems for greater than 15 minutes.

SU3 UNPLANNED loss of safety system annunciation or indication in the control room for 15 minutes or longer.

Operating Modes: Power Operation, Startup, Hot Standby, Hot Shutdown 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 condition has exceeded, or will likely exceed, the applicable time.

1.

UNPLANNED Loss of greater than approximately 75% of the following for 15 minutes or longer:

a.

(Site specific control room safety system annunciation)

OR SU3 UNPLANNED loss of safety system annunciation or indication in the control room for 15 minutes or longer.

Operating Modes: Power Operation, Startup, Hot Shutdown 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 DIFFERENCE Provided clarification on ERO positions associated with who could be Emergency Director, depending on which facility is in command and control. This is consistent throughout this recognition category.

For clarification purposes, changed loss of greater than approximately 75% to loss of approximately 75% or more without changing the intent.

Added any prior to the following to help clarify.

Site specific annunciators and indicators included (1C03, 1C04, 1C05)

Included radiation monitor indications because of language included in basis Site specific annunciators or indicators for this EAL must include those identified in the AOPs, in the EOPs and in other EALs (e.g. area, process, and/or effluent rad monitors, etc.)

Last SER Approved EALs NEI 99-01 Rev. 5 EALs Proposed EALs Difference or Deviation Explanation

b.

(Site specific control room safety system indication)

Radiation monitor indications SU4 Fuel Clad Degradation Operating Modes: Power Operation, Startup, Hot S/D SU4.1 Pretreatment Offgas System (RM-4104) Hi-Hi Radiation Alarm OR SU4.2 Reactor Coolant sample activity value GREATER THAN 2.0 Ci/gm dose equivalent I-131.

SU4 Fuel Clad Degradation.

Operating Modes: Power Operation, Startup, Hot Standby, Hot Shutdown

1.

(Site-specific radiation monitor readings indicating fuel clad degradation greater than Technical Specification allowable limits.)

OR

2.

(Site-specific coolant sample activity value indicating fuel clad degradation greater than Technical Specification allowable limits.)

SU4 Fuel Clad Degradation Operating Modes: Power Operation, Startup, Hot Shutdown SU4.1 Pretreatment Offgas System (RM-4104) Hi-Hi Radiation Alarm SU4.2 Reactor Coolant sample activity value GREATER THAN 2.0 Ci/gm dose equivalent I-131.

DIFFERENCE 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 the 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.

DAEC uses the term Reactor Coolant when referring the coolant sample.

2.0 Ci/gm dose equivalent I-131 is the maximum concentration in DAECs Technical Specifications therefore indicating fuel clad degradation.

SU5 RCS Leakage Operating Modes: Power Operation, Startup, Hot S/D SU5.1 Unidentified or pressure boundary leakage GREATER THAN 10 gpm.

OR SU5.2 Identified leakage GREATER THAN 25 gpm.

SU5 RCS Leakage.

Operating Modes: Power Operation, Startup, Hot Standby, Hot Shutdown

1.

Unidentified or pressure boundary leakage greater than 10 gpm.

OR

2.

Identified leakage greater than 25 gpm.

SU5 RCS Leakage Operating Modes: Power Operation, Startup, Hot Shutdown SU5.1 Unidentified drywell leakage GREATER THAN 10 gpm.

SU5.2 Identified drywell leakage GREATER THAN 25 gpm.

DIFFERENCE To avoid confusion, removed reference to pressure boundary leakage and simply referred to Drywell.

In SU5.2, added drywell to clarify the EAL statement and make it quicker for the Operators to determine the intent.

SU6 Unplanned Loss of All Onsite or Offsite Communications Capabilities Operating Modes: Power Operation, Startup, Hot S/D SU6.1 Loss of ALL of the following onsite communication capabilities affecting the ability to perform routine operation:

Plant Operations Radio System In-Plant Telephones Plant Paging System OR SU6.2 Loss of ALL of the following offsite communications capability:

All telephone lines (commercial)

Microwave Phone System SU6 Loss of all On-site or Off-site communications capabilities.

Operating Modes: Power Operation, Startup, Hot Standby, Hot Shutdown

1.

Loss of all of the following on-site communication methods affecting the ability to perform routine operations.

(site specific list of communications methods)

OR

2.

Loss of all of the following off-site communication methods affecting the ability to perform offsite notifications.

(site specific list of communications methods)

SU6 Loss of all onsite or offsite communications capabilities.

Operating Modes: Power Operation, Startup, Hot Shutdown 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 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 DIFFERENCE Consistent with current NEI EAL.

Included site-specific list of communication methods Added satellite and cell phones to the list for offsite communication methods as well as ENS phones to reflect the intent in the basis

Last SER Approved EALs NEI 99-01 Rev. 5 EALs Proposed EALs Difference or Deviation Explanation FTS Phone System phone system)

Control Room fixed satellite phones Microwave Phone System Emergency Notification System (ENS)

FTS Phone System SU8 Inadvertent Criticality Operating Modes: Hot S/D SU8.1 An unplanned extended positive period observed on nuclear instrumentation.

SU8 Inadvertent Criticality Operating Modes: Hot Standby, Hot Shutdown UNPLANNED sustained positive period observed on nuclear instrumentation.

SU8 Inadvertent Criticality Operating Modes: Hot Shutdown SU8.1 An UNPLANNED sustained positive period observed on nuclear instrumentation.

DIFFERENCE Consistent with current NEI EAL.

Used BWR specific EAL.

Added An at the beginning of the EAL SA2 Failure of Reactor Protection System Instrumentation to Complete or Initiate an Automatic Reactor Scram Once a Reactor Protection System Setpoint Has Been Exceeded and Manual Scram Was Successful Operating Modes: Power Operation, Startup SA2.1 Auto Scram failure AND ANY of the following operator actions to reduce power are successful in shutting down the reactor:

Manual Scram Pushbuttons Mode Switch to Shutdown Alternate Rod Insertion (ARI)

SA2 Automatic Scram (Trip) fails to shutdown the reactor and the manual actions taken from the reactor control console are successful in shutting down the reactor.

Operating Modes: Power Operation, Startup, Hot Standby

1.

a.

An automatic scram (trip) failed to shutdown the reactor.

AND

b.

Manual actions taken at the reactor control console successfully shutdown the reactor as indicated by (site specific indications of plant shutdown).

SA2 Automatic Scram fails to shutdown the reactor and the manual actions taken from 1C05 are successful in shutting down the reactor.

Operating Modes: Power Operation, Startup 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)

DIFFERENCE Used BWR-specific term of Scram versus Trip 1C05 is the Reactor Control Console at DAEC To avoid confusion regarding what qualifies as successfully shut down the reactor, DAEC has chosen to identify a specific value of 5% power as reactor shutdown. This is consistent with the current NEI EAL Reworded the logic statement AND b.Manual actions taken at the for human factoring while bringing immediate attention to the words ANY and are successful.

DAEC chose to define available methods of manually scramming the reactor from within the Control Room to clarify acceptable actions for the Alert level declaration or the Site Area Emergency declaration.

SA4 Unplanned Loss of Most or All Safety System Annunciation or Indication in Control Room With Either (1) a Significant Transient in Progress, or (2) Compensatory Non-Alarming Indicators Unavailable Operating Modes: Power Operation, Startup, Hot S/D 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.

Operating Modes: Power Operation, Startup, Hot Standby, Hot Shutdown Note: The Emergency Director should not wait until 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.

Operating Modes: Power Operation, Startup, Hot Shutdown Note: The Emergency Director (OSM, EC or DIFFERENCE Provided clarification on ERO positions associated with who could be Emergency Director, depending on which facility is in command and control. This is consistent throughout this recognition category.

For clarification purposes, changed loss of greater than approximately 75% to loss of approximately 75% or more without changing the intent.

Added any prior to the following to help clarify.

Last SER Approved EALs NEI 99-01 Rev. 5 EALs Proposed EALs Difference or Deviation Explanation SA4.1 Unplanned loss of most or all 1C03, 1C04 and 1C05 annunciators or indicators associated with Critical Safety Systems for greater than 15 minutes.

AND Either of the following conditions exist:

A significant plant transient is in progress.

Compensatory non-alarming indications are unavailable.

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.

1.

a.

UNPLANNED loss of greater than approximately 75% of the following for 15 minutes or longer:

(Site specific control room safety system annunciation)

OR (Site specific control room safety system indication)

b.

EITHER of the following:

A SIGNIFICANT TRANSIENT is in progress.

Compensatory indications are unavailable.

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:

1C03, 1C04 and 1C05 annunciators 1C03, 1C04 and 1C05 indicators Radiation monitor indications AND Either of the following conditions exist:

A SIGNIFICANT TRANSIENT is in progress.

OR Compensatory indications are unavailable.

Site specific annunciators and indicators included (1C03, 1C04, 1C05)

Included radiation monitor indications because of language included in basis Site specific annunciators or indicators for this EAL must include those identified in the AOPs, in the EOPs and in other EALs (e.g. area, process, and/or effluent rad monitors, etc.)

Added AND statement connecting 1a to 1b to be consistent with current SER approved EAL.

Added conditions exist to second IC. No change in intent.

SA5 AC Power Capability to Essential Busses Reduced to a Single Power Source for Greater Than 15 Minutes Such That Any Additional Single Failure Would Result in Station Blackout Operating Modes: Power Operation, Startup, Hot S/D SA5.1 AC power capability to 1A3 or 1A4 busses reduced to a single power source for greater than 15 minutes AND Any additional single failure will result in station blackout.

SA5 AC power capability to emergency busses reduced to a single power source for 15 minutes or longer such that any additional single failure would result in station blackout.

Operating Modes: Power Operation, Startup, Hot Standby, Hot Shutdown 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 condition has exceeded, or will likely exceed, the applicable time.

1.

a.

AC power capability to (site-specific emergency busses) reduced to a single power source for 15 minutes or longer.

AND

b.

Any additional single power source failure will result in station blackout.

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.

Operating Modes: Power Operation, Startup, Hot Shutdown 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.

NONE Using essential in Initiating Condition vice emergency. This guidance is approved per NEI FAQ 2006-17. This is also consistent with the last SER Approved EAL.

Provided clarification on ERO positions associated with who could be Emergency Director, depending on which facility is in command and control. This is consistent throughout this recognition category.

Consistent with current NEI EAL.

1A3 and 1A4 is equivalent to DAECs Essential Busses throughout this recognition category.

SS1 Loss of All Offsite Power and Loss of All Onsite AC SS1 Loss of all Off-site and all On-Site AC power to SS1 Loss of all offsite and all On-Site AC power to DIFFERENCE Using essential in Initiating Condition vice emergency. This guidance is approved per NEI FAQ

Last SER Approved EALs NEI 99-01 Rev. 5 EALs Proposed EALs Difference or Deviation Explanation Power to Essential Busses Operating Modes: Power Operation, Startup, Hot S/D SS1.1 Loss of power to or from the Startup or Standby Transformer resulting in a loss of all offsite power to Emergency Busses 1A3 and 1A4.

AND Failure of A Diesel Generator (1G-31) and B Diesel Generator (1G-21) to supply power to emergency busses 1A3 and 1A4.

AND Failure to restore power to at least one emergency bus, 1A3 or 1A4, within 15 minutes from the time of loss of both offsite and onsite AC power.

emergency busses for 15 minutes or longer.

Operating Modes: Power Operation, Startup, Hot Standby, Hot Shutdown 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 condition has exceeded, or will likely exceed, the applicable time.

1.

Loss of all Off-Site and all On-Site AC power to (site specific emergency busses) for 15 minutes or longer.

Essential Busses for 15 minutes or longer.

Operating Modes: Power Operation, Startup, Hot Shutdown 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.

2006-17. This is also consistent with the last SER Approved EAL.

Provided clarification on ERO positions associated with who could be Emergency Director, depending on which facility is in command and control. This is consistent throughout this recognition category.

Consistent with last SER Approved EAL and current NEI EAL.

1A3 and 1A4 is equivalent to DAECs Essential Busses throughout this recognition category.

SS2 Failure of Reactor Protection System Instrumentation to Complete or Initiate an Automatic Reactor Scram Once a Reactor Protection System Setpoint Has Been Exceeded and Manual Scram Was NOT Successful Operating Modes: Power Operation, Startup SS2.1 Auto Scram failure AND NONE of the following operator actions to reduce power are successful in shutting down the reactor:

Manual Scram Pushbuttons Mode Switch to Shutdown Alternate Rod Insertion (ARI)

SS2 Automatic Scram (Trip) fails to shutdown the reactor and manual actions taken from the reactor control console are not successful in shutting down the reactor.

Operating Modes: Power Operation, Startup

1.

a.

An automatic scram (trip) failed to shutdown the reactor.

AND

b.

Manual actions taken at the reactor control console do not shutdown the reactor as indicated by (site specific indications of reactor not shutdown).

SS2 Automatic Scram fails to shutdown the reactor and manual actions taken from 1C05 are NOT successful in shutting down the reactor.

Operating Modes: Power Operation, Startup 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 power below 5% power.

Manual Scram Pushbutton Mode Switch to Shutdown Alternate Rod Insertion (ARI)

DIFFERENCE Used BWR-specific term of Scram versus Trip DAEC chose to define available methods of manually scramming the reactor from within the Control Room to clarify acceptable actions for the Alert level declaration or the Site Area Emergency declaration.

1C05 is the Reactor Control Console at DAEC Reworded the logic statement AND b.Manual actions taken at the for human factoring while bringing immediate attention to the words NONE and are successful.

To avoid confusion regarding what qualifies as successfully shut down the reactor, DAEC has chosen to identify a specific value of 5% power as reactor shutdown. This is consistent with the current NEI EAL SS3 Loss of All Vital DC Power Operating Modes: Power Operation, Startup, Hot S/D SS3.1 Loss of Div 1 and Div 2 125V DC busses based on bus voltage LESS THAN 105 VDC indicated for greater than 15 minutes.

SS3 Loss of All Vital DC Power for 15 minutes or longer.

Operating Modes: Power Operation, Startup, Hot Standby, Hot Shutdown 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 condition has exceeded, or will likely exceed, the applicable time.

1.

Less than (site specific bus voltage indication)

SS3 Loss of All Vital DC Power for 15 minutes or longer.

Operating Modes: Power Operation, Startup, Hot Shutdown 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.

DIFFERENCE Provided clarification on ERO positions associated with who could be Emergency Director, depending on which facility is in command and control. This is consistent throughout this recognition category.

Provided more detail for IC, no change in intent.

Div 1 and Div 2 125v DC busses are considered vital at DAEC.

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

Last SER Approved EALs NEI 99-01 Rev. 5 EALs Proposed EALs Difference or Deviation Explanation on all (site specific Vital DC busses) for 15 minutes or longer.

SS3.1. Less than 105 VDC bus voltage on BOTH Div 1 and Div 2 125 VDC busses for 15 minutes or longer.

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.

105 VDC is the setpoint for a loss of 125 VDC busses.

SS4 Complete Loss of Heat Removal Capability Operating Modes: Power Operation, Startup, Hot S/D SS4.1 EOP Graph 4 Heat Capacity Limit is exceeded DELETED SS6 Inability to Monitor a Significant Transient in Progress Operating Modes: Power Operation, Startup, Hot S/D SS6.1 Significant transient in progress and ALL of the following:

Loss of most or all annunciators on Panels 1C03, 1C04 and 1C05.

Compensatory non-alarming indications are unavailable.

Indicators needed to monitor criticality, or core heat removal, or Fission Product Barrier status are unavailable.

SS6 Inability to Monitor a SIGNIFICANT TRANSIENT in Progress.

Operating Modes: Power Operation, Startup, Hot Standby, Hot Shutdown 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 condition has exceeded, or will likely exceed, the applicable time.

1.

a.

Loss of greater than approximately 75% of the following for 15 minutes or longer:

(Site specific control room safety system annunciation)

OR (Site specific control room safety system indication)

AND

b.

A SIGNIFICANT TRANSIENT is in progress.

AND

c.

Compensatory indications are unavailable.

SS6 Inability to monitor a SIGNIFICANT TRANSIENT in progress.

Operating Modes: Power Operation, Startup, Hot Shutdown 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 a.

Loss of approximately 75% or more of any of the following for 15 minutes or longer:

1C03, 1C04 and 1C05 annunciators OR 1C03, 1C04 and 1C05 indicators OR radiation monitor indications OR indicators needed to monitor criticality, or core heat removal, or Fission Product Barrier status AND

b.

A SIGNIFICANT TRANSIENT is in DIFFERENCE Provided clarification on ERO positions associated with who could be Emergency Director, depending on which facility is in command and control. This is consistent throughout this recognition category.

For clarification purposes, changed loss of greater than approximately 75% to loss of approximately 75% or more without changing the intent.

Added any prior to the following to help clarify.

Site specific annunciators and indicators included (1C03, 1C04, 1C05)

Included radiation monitor indications because of language included in basis Site specific annunciators or indicators for this EAL must include those identified in the AOPs, in the EOPs and in other EALs (e.g. area, process, and/or effluent rad monitors, etc.)

Last SER Approved EALs NEI 99-01 Rev. 5 EALs Proposed EALs Difference or Deviation Explanation progress.

AND

c.

Compensatory indications are unavailable.

SG1 Prolonged Loss of All Offsite Power and Prolonged Loss of All Onsite AC Power to Essential Busses Operating Modes: Power Operation, Startup, Hot S/D SG1.1 Loss of power to or from the Startup or Standby Transformer resulting in a loss of all offsite power to Emergency Busses 1A3 and 1A4.

AND Failure of A Diesel Generator (1G-31) and B Diesel Generator (1G-21) to supply power to emergency busses 1A3 and 1A4.

AND ANY ONE OF THE FOLLOWING:

Restoration of power to either Bus 1A3 or 1A4 is not likely within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

RPV level is indeterminate.

RPV Level is LESS THAN +15 inches.

SG1 Prolonged loss of all Off-site and all On-Site AC power to emergency busses.

Operating Modes: Power Operation, Startup, Hot Standby, Hot Shutdown

1.

a.

Loss of all off-site and all on-site AC power to (site specific emergency busses).

AND

b.

EITHER of the following:

Restoration of at least one emergency bus in less than (site specific hours) is not likely.

(Site specific indication of continuing degradation of core cooling based on Fission Product Barrier monitoring.)

SG1 Prolonged loss of all offsite and all onsite AC power to Essential Busses.

Operating Modes: Power Operation, Startup, Hot Shutdown SG1.1. a.

Loss of all offsite and all onsite AC power to 1A3 and 1A4.

AND

b.

ANY of the following:

Restoration of power to either 1A3 or 1A4 in less than 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is NOT likely.

RPV water level cannot be determined.

RPV water level is LESS THAN

+15 inches.

DIFFERENCE Using essential in Initiating Condition vice emergency. This guidance is approved per NEI FAQ 2006-17. This is also consistent with the last SER Approved EAL.

1A3 and 1A4 is equivalent to DAECs Essential Busses throughout this recognition category.

DAEC states the emergency busses (site-specific information) powered by the applicable transformers to minimize any potential confusion whereby the transformer would have power, but the essential bus would not. The intent of the EAL is a loss of offsite power to the essential busses, both Diesel Generators fail, and one essential bus cannot be restored within 4 hrs or an indication of potential Fission Product Barrier degradation exists.

Replaced EITHER with ANY. This did not change the intent.

Consistent with current NEI EAL.

restoration of power to either bus is equivalent to restoration of at least one since DAEC has only two emergency busses.

DAECs site specific indication of continuing degradation of core cooling based on Fission Product Barrier monitoring is:

RPV level cannot be determined - 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.

cannot be determined meets intent of is indeterminate as stated in current DAEC EAL.

This is also consistent with EOP language.

RPV Level is LESS THAN +15 inches - 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.

SG2 Failure of the Reactor Protection System to Complete an Automatic Scram and Manual Scram was NOT successful and There is Indication of an Extreme Challenge to the Ability to Cool the Core SG2 Automatic Scram (Trip) and all manual actions fail to shutdown the reactor and indication of an extreme challenge to the ability to cool the core exists.

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.

Operating Modes: Power Operation, Startup DIFFERENCE Used BWR-specific term of Scram versus Trip Consistent with last SER Approved EAL and current NEI EAL.

Reworded the logic statement AND b. All manual actions do not for human factoring.

Last SER Approved EALs NEI 99-01 Rev. 5 EALs Proposed EALs Difference or Deviation Explanation Operating Modes: Power Operation, Startup SG2.1 Auto Scram failure AND NONE of the following operator actions to reduce power are successful in shutting down the reactor:

Manual Scram Pushbuttons Mode Switch to Shutdown Alternate Rod Insertion (ARI)

AND Loss of adequate core cooling or decay heat removal capability as indicated by either:

RPV level cannot be maintained GREATER THAN -25 inches.

HCL Curve (EOP Graph 4) exceeded.

Operating Modes: Power Operation, Startup

1.

a.

An automatic scram (trip) failed to shutdown the reactor.

AND

b.

All manual actions do not shutdown the reactor as indicated by (site specific indications of reactor not shutdown).

AND

c.

EITHER of the following exist or have occurred due to continued power generation:

(Site specific indication that core cooling is extremely challenged.)

(Site specific indication that heat removal is extremely challenged.)

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.

OR Heat Capacity Limit Curve (EOP Graph 4) exceeded.

To avoid confusion, DAEC has chosen to identify a specific value of 5% power as reactor shutdown. This is consistent with the current NEI EAL

-25 inches, DAECs Minimum Steam Cooling RPV Water Level, and the HCL Curve (EOP Graph 4), gives an indication of challenged core cooling and/or decay heat removal capability.

Wording and format consistent with SA2 & SS2.

Exceeding the HCL curve from DAECs EOP Graph 4 is site-specific indication that heat removal is extremely challenged.

NG-12-0064 86 Pages Follow ENCLOSURE 3 REVISED EMERGENCY ACTION LEVEL DESIGN BASIS DOCUMENTS FOR NEXTERA ENERGY DUANE ARNOLD

EAL BASES DOCUMENT EBD-C COLD SHUTDOWN/REFUELING - SYSTEMS MALFUNCTION Rev. 1 next Page 1 of 28 Usage Level Information Use Effective Date:

TECHNICAL REVIEW Prepared by:

Date:

Reviewed by:

Date:

Emergency Planning Staff PROCEDURE APPROVAL I am responsible for the technical content of this procedure.

Approved by:

Date:

Manager, Emergency Planning

EAL BASES DOCUMENT EBD-C COLD SHUTDOWN/REFUELING - SYSTEMS MALFUNCTION Rev. 1 next Page 2 of 28 CU1 CU1 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:

EAL BASES DOCUMENT EBD-C COLD SHUTDOWN/REFUELING - SYSTEMS MALFUNCTION Rev. 1 next Page 3 of 28 CU1

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

EAL BASES DOCUMENT EBD-C COLD SHUTDOWN/REFUELING - SYSTEMS MALFUNCTION Rev. 1 next Page 4 of 28 CU2 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

EAL BASES DOCUMENT EBD-C COLD SHUTDOWN/REFUELING - SYSTEMS MALFUNCTION Rev. 1 next Page 5 of 28 CU2 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:

EAL BASES DOCUMENT EBD-C COLD SHUTDOWN/REFUELING - SYSTEMS MALFUNCTION Rev. 1 next Page 6 of 28 CU2

1.

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

EAL BASES DOCUMENT EBD-C COLD SHUTDOWN/REFUELING - SYSTEMS MALFUNCTION Rev. 1 next Page 7 of 28 CU3 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

EAL BASES DOCUMENT EBD-C COLD SHUTDOWN/REFUELING - SYSTEMS MALFUNCTION Rev. 1 next Page 8 of 28 CU3 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

EAL BASES DOCUMENT EBD-C COLD SHUTDOWN/REFUELING - SYSTEMS MALFUNCTION Rev. 1 next Page 9 of 28 CU4 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 ºF 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 0.0278 hours 1.653439e-4 weeks 3.805e-5 months 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

EAL BASES DOCUMENT EBD-C COLD SHUTDOWN/REFUELING - SYSTEMS MALFUNCTION Rev. 1 next Page 10 of 28 CU4 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

EAL BASES DOCUMENT EBD-C COLD SHUTDOWN/REFUELING - SYSTEMS MALFUNCTION Rev. 1 next Page 11 of 28 CU6 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.

EAL BASES DOCUMENT EBD-C COLD SHUTDOWN/REFUELING - SYSTEMS MALFUNCTION Rev. 1 next Page 12 of 28 CU6 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

EAL BASES DOCUMENT EBD-C COLD SHUTDOWN/REFUELING - SYSTEMS MALFUNCTION Rev. 1 next Page 13 of 28 CU7 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

EAL BASES DOCUMENT EBD-C COLD SHUTDOWN/REFUELING - SYSTEMS MALFUNCTION Rev. 1 next Page 14 of 28 CU8 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

EAL BASES DOCUMENT EBD-C COLD SHUTDOWN/REFUELING - SYSTEMS MALFUNCTION Rev. 1 next Page 15 of 28 CA1 CA1 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 0.0278 hours 1.653439e-4 weeks 3.805e-5 months 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

EAL BASES DOCUMENT EBD-C COLD SHUTDOWN/REFUELING - SYSTEMS MALFUNCTION Rev. 1 next Page 16 of 28 CA1 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 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months 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

EAL BASES DOCUMENT EBD-C COLD SHUTDOWN/REFUELING - SYSTEMS MALFUNCTION Rev. 1 next Page 17 of 28 CA3 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:

EAL BASES DOCUMENT EBD-C COLD SHUTDOWN/REFUELING - SYSTEMS MALFUNCTION Rev. 1 next Page 18 of 28 CA3

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

EAL BASES DOCUMENT EBD-C COLD SHUTDOWN/REFUELING - SYSTEMS MALFUNCTION Rev. 1 next Page 19 of 28 CA4 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 temperature is 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 ºF for 60 minutes or longer. Note: If an RCS heat removal system is in operation within the specified time frames and RCS temperature is 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 ºF, use the table to select the appropriate EAL.

RCS Integrity Secondary Containment Duration temperature GREATER THAN 212 ºF EAL Not Established Not Established 0 minutes CA4.1 Not Established Established 20 minutes CA4.2 Established N/A 60 minutes CA4.3

EAL BASES DOCUMENT EBD-C COLD SHUTDOWN/REFUELING - SYSTEMS MALFUNCTION Rev. 1 next Page 20 of 28 CA4 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 restoring an RCS heat removal system to operation and RCS temperature is 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 applicable if actions are successful in restoring an 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 remained less 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

EAL BASES DOCUMENT EBD-C COLD SHUTDOWN/REFUELING - SYSTEMS MALFUNCTION Rev. 1 next Page 21 of 28 CA4 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 ºF 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

EAL BASES DOCUMENT EBD-C COLD SHUTDOWN/REFUELING - SYSTEMS MALFUNCTION Rev. 1 next Page 22 of 28 CS1 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

EAL BASES DOCUMENT EBD-C COLD SHUTDOWN/REFUELING - SYSTEMS MALFUNCTION Rev. 1 next Page 23 of 28 CS1 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 0.0278 hours 1.653439e-4 weeks 3.805e-5 months 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 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 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

EAL BASES DOCUMENT EBD-C COLD SHUTDOWN/REFUELING - SYSTEMS MALFUNCTION Rev. 1 next Page 24 of 28 CS1 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

EAL BASES DOCUMENT EBD-C COLD SHUTDOWN/REFUELING - SYSTEMS MALFUNCTION Rev. 1 next Page 25 of 28 CG1 CG1 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):

EAL BASES DOCUMENT EBD-C COLD SHUTDOWN/REFUELING - SYSTEMS MALFUNCTION Rev. 1 next Page 26 of 28 CG1 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 Operation at Commercial Nuclear Power Plants in the United States, and, NUMARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management.

EAL BASES DOCUMENT EBD-C COLD SHUTDOWN/REFUELING - SYSTEMS MALFUNCTION Rev. 1 next Page 27 of 28 CG1 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 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 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

EAL BASES DOCUMENT EBD-C COLD SHUTDOWN/REFUELING - SYSTEMS MALFUNCTION Rev. 1 next Page 28 of 28 CG1 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

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

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

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EAL BASES DOCUMENT EBD-H HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY Rev. 11 next PAGE 2 of 30 HU1 HU1 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

EAL BASES DOCUMENT EBD-H HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY Rev. 11 next PAGE 3 of 30 HU1 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 HA1.

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.

EAL BASES DOCUMENT EBD-H HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY Rev. 11 next PAGE 4 of 30 HU1 Safe Shutdown/Vital 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 Sink /

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

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 BASES DOCUMENT EBD-H HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY Rev. 11 next PAGE 5 of 30 HU1 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

EAL BASES DOCUMENT EBD-H HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY Rev. 11 next PAGE 6 of 30 HU2 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 Shutdown/Vital 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.

EAL BASES DOCUMENT EBD-H HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY Rev. 11 next PAGE 7 of 30 HU2 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 Shutdown/Vital 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 Sink /

Coolant Supply Torus Room, Intake Structure, Pumphouse Containment Drywell, Torus Emergency Systems NE, NW, SE Corner Rooms, HPCI Room, RCIC Room, RHR Valve Room, North CRD Area, South CRD Area, CSTs Other 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

EAL BASES DOCUMENT EBD-H HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY Rev. 11 next PAGE 8 of 30 HU3 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

EAL BASES DOCUMENT EBD-H HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY Rev. 11 next PAGE 9 of 30 HU4 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/compromise to site security, threat/risk to site personnel, or a potential degradation to the level of safety of the plant. A SECURITY CONDITION does not involve a HOSTILE ACTION.

Security 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 sites 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.

EAL BASES DOCUMENT EBD-H HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY Rev. 11 next PAGE 10 of 30 HU4 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

EAL BASES DOCUMENT EBD-H HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY Rev. 11 next PAGE 11 of 30 HU5 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

EAL BASES DOCUMENT EBD-H HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY Rev. 11 next PAGE 12 of 30 HA1 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 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

EAL BASES DOCUMENT EBD-H HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY Rev. 11 next PAGE 13 of 30 HA1 OR 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

EAL BASES DOCUMENT EBD-H HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY Rev. 11 next PAGE 14 of 30 HA1 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

EAL BASES DOCUMENT EBD-H HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY Rev. 11 next PAGE 15 of 30 HA1 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 Electrical Power 1G31 DG and Day Tank Rooms, 1G21 DG and Day Tank Rooms, Battery Rooms, Essential Switchgear Rooms, Cable Spreading Room Heat Sink /

Coolant Supply Torus Room, Intake Structure, Pumphouse Containment Drywell, Torus Emergency Systems NE, NW, SE Corner Rooms, HPCI Room, RCIC Room, RHR Valve Room, North CRD Area, South CRD Area, CSTs Other Control Building, Remote Shutdown Panel 1C388 Area, Panel 1C55/56 Area, SBGT Room 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 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 BASES DOCUMENT EBD-H HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY Rev. 11 next PAGE 16 of 30 HA1 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

EAL BASES DOCUMENT EBD-H HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY Rev. 11 next PAGE 17 of 30 HA2 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 Shutdown/Vital 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 Shutdown/Vital 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 Sink /

Coolant Supply Torus Room, Intake Structure, Pumphouse Containment Drywell, Torus Emergency Systems NE, NW, SE Corner Rooms, HPCI Room, RCIC Room, RHR Valve Room, 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.

EAL BASES DOCUMENT EBD-H HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY Rev. 11 next PAGE 18 of 30 HA2 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 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 BASES DOCUMENT EBD-H HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY Rev. 11 next PAGE 19 of 30 HA4 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 ISFSIs that may be outside the PROTECTED AREA but still within the OWNER CONTROLLED AREA.

EAL BASES DOCUMENT EBD-H HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY Rev. 11 next PAGE 20 of 30 HA4 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

EAL BASES DOCUMENT EBD-H HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY Rev. 11 next PAGE 21 of 30 HA5 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

EAL BASES DOCUMENT EBD-H HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY Rev. 11 next PAGE 22 of 30 HA6 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

EAL BASES DOCUMENT EBD-H HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY Rev. 11 next PAGE 23 of 30 HS2 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 Director is 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

EAL BASES DOCUMENT EBD-H HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY Rev. 11 next PAGE 24 of 30 HS2 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

EAL BASES DOCUMENT EBD-H HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY Rev. 11 next PAGE 25 of 30 HS3 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

EAL BASES DOCUMENT EBD-H HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY Rev. 11 next PAGE 26 of 30 HS4 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.

EAL BASES DOCUMENT EBD-H HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY Rev. 11 next PAGE 27 of 30 HS4 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

EAL BASES DOCUMENT EBD-H HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY Rev. 11 next PAGE 28 of 30 HG1 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 Shutdown/Vital 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 Shutdown/Vital 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.

EAL BASES DOCUMENT EBD-H HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY Rev. 11 next PAGE 29 of 30 HG1

REFERENCES:

1. UFSAR Section 6.4, Habitability Systems

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

EAL BASES DOCUMENT EBD-H HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY Rev. 11 next PAGE 30 of 30 HG2 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

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

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EAL BASES DOCUMENT EBD-R ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT Rev. 11 next PAGE 2 of 28 RU1 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 Ci/cc and is expected to continue for 60 minutes or longer.

OR RU1.2 VALID Offgas Stack rad monitor (Kaman 9/10) reading GREATER THAN 2.0 E-1 Ci/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 Ci/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-1997) 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.

EAL BASES DOCUMENT EBD-R ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT Rev. 11 next PAGE 3 of 28 RU1 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.

EAL BASES DOCUMENT EBD-R ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT Rev. 11 next PAGE 4 of 28 RU1 GASEOUS EFFLUENT EALS Offgas Stack Kaman 9/10 Turbine Bldg (Kaman 1/2)

Reactor Bldg (Kaman 3/4, 5/6, 7/8)

Assumed flow (CFM) 10,000 72,000 93,000 Release Limits Concentration (Ci/cc)

Release Rate (Ci/sec)

Concentration (Ci/cc)

Release Rate (Ci/sec)

Concentration (Ci/cc)

Release Rate (Ci/sec)

Tech Spec 1.1E-1 5.2E+5 6.2E-4 2.1E+4 4.8E-4 2.1E+4 Unusual Event (2 x TS) 2.0E-1 1.0E+6 1.0E-3 4.2E+4 1.0E-3 4.2E+4 Alert (60 x TS) 6.0E+0 3.0E+7 3.0E-2 1.3E+6 3.0E-2 1.3E+6 LLRPSF Kaman 12 Assumed flow (CFM) 75,000 Release Limits Concentration (Ci/cc)

Release Rate (Ci/sec)

Tech Spec 5.9E-4 2.1E+4 Unusual Event (2 x TS) 1.0E-3 4.2E+4 Alert (200 x TS) 1.0E-1 4.2E+6

EAL BASES DOCUMENT EBD-R ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT Rev. 11 next PAGE 5 of 28 RU1 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 Ci/cc. Rounded off, this corresponds to 1E-3 Ci/cc.

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

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

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

EAL BASES DOCUMENT EBD-R ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT Rev. 11 next PAGE 6 of 28 RU1 Monitor TS Limit Reading Unusual Event Level Alert Level GSW 1,555 CPS 1.5 X 103 CPS 3.0 X 103 CPS 3.0 X 105 CPS RHRSW & ESW to cooling tower 413 CPS 4.0 X 102 CPS 8.0 X 102 CPS 8.0 X 104 CPS RHRSW & ESW to Discharge Canal 507 CPS 5.0 X 102 CPS 1.0 X 103 CPS 1.0 X 105 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

EAL BASES DOCUMENT EBD-R ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT Rev. 11 next PAGE 7 of 28 RU2 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 reading in the past twenty-four hours excluding the current peak value.

DAEC EAL INFORMATION:

Unplanned means 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.

EAL BASES DOCUMENT EBD-R ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT Rev. 11 next PAGE 8 of 28 RU2 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 LI3413 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.

EAL BASES DOCUMENT EBD-R ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT Rev. 11 next PAGE 9 of 28 RU2

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) 3.0.0.0-01PA, 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.

EAL BASES DOCUMENT EBD-R ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT Rev. 11 next PAGE 10 of 28 RA1 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 Ci/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 Ci/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 Ci/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.

EAL BASES DOCUMENT EBD-R ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT Rev. 11 next PAGE 11 of 28 RA1 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 µCi/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.

EAL BASES DOCUMENT EBD-R ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT Rev. 11 next PAGE 12 of 28 RA1 GASEOUS EFFLUENT EALS Offgas Stack Kaman 9/10 Turbine Bldg (Kaman 1/2)

Reactor Bldg (Kaman 3/4, 5/6, 7/8)

Assumed flow (CFM) 10,000 72,000 93,000 Release Limits Concentratio n (Ci/cc)

Release Rate (Ci/sec)

Concentration (Ci/cc)

Release Rate (Ci/sec)

Concentration (Ci/cc)

Release Rate (Ci/sec)

Tech Spec 1.1E-1 5.2E+5 6.2E-4 2.1E+4 4.8E-4 2.1E+4 Unusual Event (2 x TS) 2.0E-1 1.0E+6 1.0E-3 4.2E+4 1.0E-3 4.2E+4 Alert (60 x TS) 6.0E+0 3.0E+7 3.0E-2 1.3E+6 3.0E-2 1.3E+6 LLRPSF Kaman 12 Assumed flow (CFM) 75,000 Release Limits Concentration (Ci/cc)

Release Rate (Ci/sec)

Tech Spec 5.9E-4 2.1E+4 Unusual Event (2 x TS) 1.0E-3 4.2E+4 Alert (200 x TS) 1.0E-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.

EAL BASES DOCUMENT EBD-R ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT Rev. 11 next PAGE 13 of 28 RA1 For the Turbine Building, 60X the TS setpoint is 3.7E-2 Ci/cc. Rounded off, this corresponds to 3E-2 Ci/cc.

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

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

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

EAL BASES DOCUMENT EBD-R ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT Rev. 11 next PAGE 14 of 28 RA1 Monitor TS Limit Reading Unusual Event Level Alert Level GSW 1,555 CPS 1.5 X 103 CPS 3.0 X 103 CPS 3.0 X 105 CPS RHRSW & ESW to cooling tower 413 CPS 4.0 X 102 CPS 8.0 X 102 CPS 8.0 X 104 CPS RHRSW & ESW to Discharge Canal 507 CPS 5.0 X 102 CPS 1.0 X 103 CPS 1.0 X 105 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

EAL BASES DOCUMENT EBD-R ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT Rev. 11 next PAGE 15 of 28 RA2 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,

EAL BASES DOCUMENT EBD-R ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT Rev. 11 next PAGE 16 of 28 RA2 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 (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 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.

EAL BASES DOCUMENT EBD-R ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT Rev. 11 next PAGE 17 of 28 RA2 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

EAL BASES DOCUMENT EBD-R ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT Rev. 11 next PAGE 18 of 28 RA2

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.

EAL BASES DOCUMENT EBD-R ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT Rev. 11 next PAGE 19 of 28 RA3 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

EAL BASES DOCUMENT EBD-R ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT Rev. 11 next PAGE 20 of 28 RA3

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

EAL BASES DOCUMENT EBD-R ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT Rev. 11 next PAGE 21 of 28 RS1 RS1 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 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+1 Ci/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.

EAL BASES DOCUMENT EBD-R ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT Rev. 11 next PAGE 22 of 28 RS1 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

µCi/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.

DAECs 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

EAL BASES DOCUMENT EBD-R ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT Rev. 11 next PAGE 23 of 28 RS1 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 Site Area Emergency RS1 General Emergency RG1 Valid Turbine or Reactor Building ventilation rad monitor (KAMAN) reading, for 15 minutes or longer, above:

0.06 Ci/cc 0.6 Ci/cc Valid Offgas Stack ventilation rad monitor (Kaman) reading, for 15 minutes or longer, above:

40 Ci/cc 400 Ci/cc 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.

EAL BASES DOCUMENT EBD-R ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT Rev. 11 next PAGE 24 of 28 RS1

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 & 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 Protective Actions for Nuclear Incidents

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

EAL BASES DOCUMENT EBD-R ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT Rev. 11 next PAGE 25 of 28 RG1 RG1 Offsite 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 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 Ci/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.

EAL BASES DOCUMENT EBD-R ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT Rev. 11 next PAGE 26 of 28 RG1 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 µCi/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.

DAECs 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:

EAL BASES DOCUMENT EBD-R ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT Rev. 11 next PAGE 27 of 28 RG1 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 Site Area Emergency RS1 General Emergency RG1 Valid Turbine or Reactor Building ventilation rad monitor (KAMAN) reading, for 15 minutes or longer, above:

0.06 Ci/cc 0.6 Ci/cc Valid Offgas Stack ventilation rad monitor (Kaman) reading, for 15 minutes or longer, above:

40 Ci/cc 400 Ci/cc 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.

EAL BASES DOCUMENT EBD-R ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT Rev. 11 next PAGE 28 of 28 RG1 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 Protective Actions for Nuclear Incidents

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

v t e Letter Sequence Response to RAI
TAC:ME6508, (Approved, Closed)
Initiation Acceptance... Administration Questions 17 January 2012 7 February 2012 1 April 2012 Answers 16 March 2012 Results Approval Other: ML12011A060, ML12135A007
MONTHYEAR2012-03-16 [Table View]

NG-12-0064, Response to Request for Additional Information Amendment to Change Emergency Action Levels

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NG-12-0064, Response to Request for Additional Information Amendment to Change Emergency Action Levels

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