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{{Adams
#REDIRECT [[0CAN121102, Proposed Emergency Action Levels Using NEI 99-01 Revision 5 Scheme]]
| number = ML113350317
| issue date = 12/01/2011
| title = Proposed Emergency Action Levels Using NEI 99-01 Revision 5 Scheme
| author name = Pyle S L
| author affiliation = Entergy Operations, Inc
| addressee name =
| addressee affiliation = NRC/NRR, NRC/Document Control Desk
| docket = 05000313, 05000368
| license number = DPR-051, NPF-006
| contact person =
| case reference number = 0CAN121102
| document type = Emergency Preparedness-Emergency Plan Implementing Procedures, Letter
| page count = 390
}}
 
=Text=
{{#Wiki_filter:0CAN121102  
 
December 1, 2011
 
U.S. Nuclear Regulatory Commission
 
Attn: Document Control Desk
 
Washington, DC  20555
 
==SUBJECT:==
Proposed Emergency Action Levels Using NEI 99-01 Revision 5 Scheme  
 
Arkansas Nuclear One - Units 1 and 2
 
Docket Nos. 50-313 and 50-368
 
License Nos. DPR-51 and NPF-6
 
==References:==
: 1. Letter from Christopher G. Miller (U.S. Nuclear Regulatory Commission) to Alan Nelson (Nuclear Energy Institute) - "
U.S. Nuclear Regulatory Commission Review and Endorsement of NEI 99-01, Revision 5, dated February 2008
", dated February 22, 2008 (ML080430535)
: 2. Entergy letter dated July 18, 2011, "Proposed Emergency Action Levels Using NEI 99-01 Revision 5 Scheme,"
TAC Nos. ME6719 and 6720 (ML112000124) (0CAN071102)
: 3. Entergy letter dated July 27, 2011, "Supplement to Proposed Emergency Action Levels Using NEI 99-01 Revision 5 Scheme,"
TAC Nos. ME6719 and ME6720 (ML112082804) (0CAN071105)
 
==Dear Sir or Madam:==
 
Pursuant to 10 CFR 50, Appendix E, Section IV.B(1), Entergy Operations, Inc. (Entergy) hereby
 
requests NRC review and approval of the Arkansas Nuclear One (ANO) proposed revision to
 
the Emergency Plan (EP) Emergency Action Levels (EALs). The proposed changes involve
 
revisions to ANO's current EP EAL scheme which is based on NUREG 0654, "Criteria for
 
Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in
 
Support of Nuclear Power Plants" (the current "implemented" version of the EALs at ANO). 
 
Entergy is proposing to adopt the EAL scheme based on the guidance provided in NEI 99-01, Revision 5, which has been endorsed by the NRC in Reference 1 above.
 
By Reference 2, as supplemented by Reference 3, Entergy submitted a request for NRC review and approval of the ANO proposed revision to the EP EALs. Following these submittals, ANO
 
began development of Operator training modules that would be used to familiarize station Operators with the new EAL schemes. In so doing, changes were identified with regard to the Entergy Operations, Inc.
1448 S.R. 333 Russellville, AR  72802
 
Tel  479-858-4704 Stephenie L Pyle Manager, Licensing A rkansas Nuclear One
 
0CAN121102 Page 2 of 2
 
original ANO submittal (Reference 2) that were found to be necessary to support proper
 
implementation of the new EAL schemes, onc e approved. This submittal, therefore, supersedes the previous submittals described in References 2 and 3 above, and includes the
 
changes identified during the aforementioned dev elopment of Operator training modules.
 
Based on the above, Entergy requests the Reference 2 and 3 submittals be withdrawn from
 
NRC review. Upon NRC approval of the EP EAL changes included in this submittal, Entergy
 
requests a period of 6 months to implement the change.
 
This letter contains no new commitments. If you have any questions or require additional
 
information, please contact me.
 
Sincerely, Original signed by Stephenie L. Pyle SLP/dbb
 
Attachments: 1. NEI 99-01 Revision 5 Deviation-Difference Document 2. Proposed Technical Basis Document (Markup) 3. Proposed Technical Basis Document (Clean)
: 4. Proposed EAL Matrix Chart and Review Table (for information)
: 5. Supporting Referenced Document Pages
 
cc: Mr. Elmo Collins Regional Administrator
 
U. S. Nuclear Regulatory Commission
 
Region IV
 
612 E. Lamar Blvd., Suite 400 Arlington, TX 76011-4125
 
NRC Senior Resident Inspector
 
Arkansas Nuclear One
 
P.O. Box 310
 
London, AR 72847
 
U. S. Nuclear Regulatory Commission
 
Attn: Mr. Kaly Kalyanam
 
MS O-8 B1
 
One White Flint North
 
11555 Rockville Pike
 
Rockville, MD 20852 
 
Attachment 1 to 0CAN121102 NEI 99-01 Revision 5 Deviation-Difference Document    to 0CAN121102
 
Page 1 of 110
 
ARKANSAS NUCLEAR ONE DEVIATIONS AND DIFFERENCES FROM  NEI 99-01, REV 5 EMERGENCY ACTION LEVELS to 0CAN121102
 
Page 2 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document INTRODUCTION This document presents the Arkansas Nuclear One (ANO) site-specific deviations and
 
differences from the Nuclear Energy Institute (NEI) 99-01, Revision 5, Emergency Action Levels (EALs).
 
The following definitions from Supplements 1 and 2 to Regulatory Information Summary (RIS) 2003-18 were used when determining the categorization of differences between the
 
NEI 99-01, Revision 5, Initiating Conditions (ICs) and example EALs, and the proposed ANO
 
ICs and EALs:
 
Deviation: An EAL change where the basis scheme guidance (NUREG, Nuclear Management and Resources Council, NEI) differs in wording and is altered in
 
meaning or intent, such that the classification of the event could be different
 
between the basis scheme guidance and the site-specific proposed EAL. 
 
Examples of deviations include the use of altered mode applicability, altering
 
key words or time limits, or changing words of physical reference (protected
 
area, safety-related equipment, etc.).
There are no deviations in the ANO proposed EAL scheme.
Difference:
A difference is an EAL change where the basis scheme guidance differs in wording, but agrees in meaning and intent, such that classification of an
 
event would be the same, whether using the basis scheme guidance or the
 
site-specific proposed EAL. Examples of differences include the use of
 
site-specific terminology or administrative re-formatting of site-specific EALs.
 
Administrative changes that do not actually change the text are neither differences nor
 
deviations. Likewise, any format change that does not alter the wording of the IC or EAL is
 
considered neither a difference nor a deviation.
 
Formatting such as ALL CAPS, bold , and underline is utilized to aid the user in applying these EALs, particularly to set apart units, time frames, or quality of a value or data (such as the term
 
"valid"). Such formatting is neither a deviation nor a difference in accordance with the
 
definitions provided above because it does not alter the wording of the IC or EAL.
 
In addition, due to the nature of this submittal acronyms in the EALs are not necessarily defined.
 
to 0CAN121102
 
Page 3 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document The following differences are generic in nature and apply throughout the proposed ANO EALs:
: 1. In general, NEI 99-01, Revision 5, bases developer notes are not included in the ANO bases, but were used in their development. NEI 99-01, Revision 5, developer note bases
 
information was selectively provided in the ANO bases where it was viewed that the
 
developer notes would provide useful training information or aid the decision maker in
 
evaluating the event. In some cases, these developer notes are reworded from the
 
NEI 99-01 EALs, but the intent is retained.
: 2. Formatting choices may also involve minor grammatical differences between the ANO EALs and NEI 99-01 such as "that exceeds" vice "exceeding," use of "If, then" statements
 
for conditional statements, or the use of symbols (>, <). Such formatting differences
 
between the ANO EALs and NEI 99-01 are not noted in this document as differences or
 
deviations when they represent format choices alone and do not change the intent or
 
materially change the content of NEI 99-01 ICs or EALs.
: 3. At ANO, the emergency classification of Notification of Unusual Event is indicated by "Notification of Unusual Event" or the abbreviation "NUE."
: 4. At ANO, the Radiological Effluent Technical Specifications (RETS) are included in the Offsite Dose Calculation Manual (ODCM); theref ore, "ODCM" is used in place of references to RETS. 
: 5. "Shift Manager (SM)/Technical Support Center (TSC) Director/Emergency Operations Facility (EOF) Director" or "SM" is used instead of "Emergency Director".
: 6. "Safeguards Contingency Plan" is the term used to encompass all security plans/documents.
: 7. At ANO the "refueling canal" performs the functions of the "reactor refueling cavity" and "fuel transfer canal."
: 8. The term "reactor vessel" was used in place of "reactor pressure vessel (RPV)."
: 9. The term "release permit" was used in place of "radioactivity discharge permit."
: 10. In the Fission Product Barrier EALs, the EAL numbers are preceded by "FCB" for the Fuel Clad Barrier EALs, "RCB" for the Reactor Coolant System (RCS) Barrier EALs, and "CNB"
 
for the Containment Barrier EALs.
: 11. The term "threshold" is not used in every case as it is used in NEI 99-01. Replacement terms such as "EAL" are used as necessary based on context. to 0CAN121102
 
Page 4 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document NEI 99-01:  AU1 Any release of gaseous or liquid radioactivity to the environment greater than 2 times the
 
RETS/ODCM for 60 minutes or longer
 
Operating Mode Applicability: All  Example Emergency Action Levels: (1 or 2 or 3 or 4 or 5)
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)
: 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.
: 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.
: 4. VALID reading on perimeter radiation moni toring system greater than 0.10 mR/hr above normal* background for 60 minutes or longer. [for sites having telemetered perimeter
 
monitors]
: 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. to 0CAN121102
 
Page 5 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document ANO:  AU1 Any release of gaseous or liquid radioactivity to the environment > 2 times the ODCM limits for  60 minutes Operating Mode Applicability:
All  Emergency Action Level(s):
(1 or 2 or 3)  Note: The SM 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 Channel 7 on any of the following radiation monitors > the reading shown for  60 minutes:
MONITORS - UNIT 1 LIMIT RX-9820 Containment Purge 5.90E-2 &#xb5;Ci/cc RX-9825 Radwaste Area 5.36E-2 &#xb5;Ci/cc RX-9830 Fuel Handling Area 4.54E-2 &#xb5;Ci/cc RX-9835 Emergency Penetration Room 9.56E-1 &#xb5;Ci/cc MONITORS - UNIT 2 LIMIT 2RX-9820 Containment Purge 4.46E-2 &#xb5;Ci/cc 2RX-9825 Radwaste Area 3.32E-2 &#xb5;Ci/cc 2RX-9830 Fuel Handling Area 4.46E-2 &#xb5;Ci/cc 2RX-9835 Emergency Penetration Room 8.84E-1 &#xb5;Ci/cc 2RX-9840 Post Accident Sampling Building 4.42E-1 &#xb5;Ci/cc 2RX-9845 Aux. Building Extension 1.26E-1 &#xb5;Ci/cc 2RX-9850 Low-Level Radwaste Storage Building1.77E-1 &#xb5;Ci/cc OR    to 0CAN121102
 
Page 6 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document ANO:  AU1 (Cont'd)
: 2. VALID reading on any of the following radiation monitors > 2 times the alarm setpoint established by a current release permit for  60 minutes:
EFFLUENT MONITORS - Unit 1 RX-9820 Containment Purge (Channel 7 or 9) RE-4830 Waste Gas Radiation Monitor RE-4642 Liquid Radwaste Monitor EFFLUENT MONITORS - Unit 2 2RX-9820 Containment Purge (Channel 7 or 9) 2RE-2429 Waste Gas Decay Tank Vent Line Radiation Monitor 2RE-2330 Regenerative Waste Discharge Monitor 2RE-4423 Radwaste Liquid Discharge Monitor 2RE-4425 SG Blowdown to Flume Radiation Monitor OR  3. Confirmed grab sample analyses for gaseous or liquid releases indicates concentrations or release rates > 2 times the applicable values of the ODCM for  60 minutes.
Deviations:
None. Differences:
The second (superfluous) "reading" is deleted in ANO EAL #2.
The radiation monitor channel is identified in EAL #1 in order to provide site-specific detail.
A table is provided in the EAL document that lists the applicable radiation monitors for EAL #2.
ANO has installed telemetered perimeter m onitoring devices. These devices, however, are neither qualified nor intended to provide an accurate indication of the dose rate at the perimeter
 
of the site. They are only intended to provide early indication of a potential unmonitored offsite
 
release. Therefore, ANO has not included an EAL comparable to NEI 99-01 Revision 5 AU1
 
EAL #4. The Radiological Dose Assessment Computer Sy stem (RDACS) is ANO's current real-time dose assessment system, but is out-dated. RDA CS is currently scheduled to be replaced prior to implementation of the new EALs with a system that is not a "real-time" system; therefore, an EAL comparable to NEI 99-01, Revision 5, AU1 EAL #5 has not been included. to 0CAN121102
 
Page 7 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document NEI 99-01:  AU2 UNPLANNED rise in plant radiation levels
 
Operating Mode Applicability:
All  Example Emergency Action Levels:
(1 or 2) 
: 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).
: 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.
 
to 0CAN121102
 
Page 8 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document ANO:  AU2 UNPLANNED rise in plant radiation levels
 
Operating Mode Applicability:
All  Emergency Action Level(s): (1 or 2)
: 1. a. UNPLANNED lowering of water level in the refueling canal or spent fuel pool as indicated by:
Personnel observation, refueling crew report, indication on area security camera, borated water source (BWST or RWT) level drop due to makeup
 
demands. AND  b. VALID Area Radiation Monitor reading rise on any of the following:
Unit 1 RE-8009 Spent Fuel Area RE-8017 Fuel Handling Area Unit 2 2RE-8914 Spent Fuel Area 2RE-8915 Spent Fuel Area 2RE-8916 Spent Fuel Area 2RE-8912 Containment Incore Instrumentation OR  2. UNPLANNED VALID Area Radiation Monitor readings or survey results indicate a rise by a factor of 1000 over normal* levels NOTE: For area radiation monitors with ranges incapable of measuring 1000 times normal* levels, classification shall be based on valid full scale indication unless surveys confirm that area radiation levels are below 1000 times normal* within 15 minutes of the Area Radiation Monitor indications going to full scale indication.
* Normal can be considered as the highest reading in the past twenty-four hours excluding the current peak value.
Deviations:
None. to 0CAN121102
 
Page 9 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document ANO:  AU2 (Cont'd)
Differences:
A note is added to EAL #2 to address the condition where 1,000 times normal levels may
 
provide a value beyond the upper range of the applicable Area Radiation Monitor. to 0CAN121102
 
Page 10 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document NEI 99-01:  AA1 Any release of gaseous or liquid radioactivity to the environment greater than 200 times the
 
RETS/ODCM for 15 minutes or longer
 
Operating Mode Applicability:
All  Example Emergency Action Levels:
(1 or 2 or 3 or 4 or 5)
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)
: 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.
: 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.
: 4. VALID reading on perimeter radiation moni toring system reading greater than 10.0 mR/hr above normal* background for 15 minutes or longer. [for sites having telemetered perimeter
 
monitors]
: 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. to 0CAN121102
 
Page 11 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document ANO:  AA1 Any release of gaseous or liquid radioactivity to the environment > 200 times the ODCM limits for  15 minutes Operating Mode Applicability:
All  Emergency Action Level(s):
(1 or 2 or 3)  Note: The SM / TSC Director / EOF 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 Channel 7 on any of the following radiation monitors > the reading shown for  15 minutes:
MONITORS - UNIT 1 LIMIT RX-9820 Containment Purge 5.90E0 &#xb5;Ci/cc RX-9825 Radwaste Area 5.36E0 &#xb5;Ci/cc RX-9830 Fuel Handling Area 4.54E0 &#xb5;Ci/cc RX-9835 Emergency Penetration Room 9.56E+1 &#xb5;Ci/cc MONITORS - UNIT 2 LIMIT 2RX-9820 Containment Purge 4.46E0 &#xb5;Ci/cc 2RX-9825 Radwaste Area 3.32E0 &#xb5;Ci/cc 2RX-9830 Fuel Handling Area 4.46E0 &#xb5;Ci/cc 2RX-9835 Emergency Penetration Room 8.84E+1 &#xb5;Ci/cc 2RX-9840 Post Accident Sampling Building 4.42E+1 &#xb5;Ci/cc 2RX-9845 Aux. Building Extension 1.26E+1 &#xb5;Ci/cc 2RX-9850 Low-Level Radwaste Storage Building 1.77E+1 &#xb5;Ci/cc OR    to 0CAN121102
 
Page 12 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document ANO:  AA1 (Cont'd)
: 2. EITHER VALID reading on any of the following radiation monitors > 200 times the alarm setpoint established by a current release permit for  15 minutes OR VALID reading greater than the value listed for  15 minutes:
MONITORS - UNIT 1 LIMIT RX-9820 Containment Purge (Channel 7 or 9) N/A RE-4830 Waste Gas Radiation Monitor 9.5E7 cpm RE-4642 Liquid Radwaste Monitor 9.5E7 cpm MONITORS - UNIT 2 LIMIT 2RX-9820 Containment Purge (Channel 7 or 9) N/A 2RE-2429 Waste Gas Decay Tank Vent Line Radiation Monitor 9.5E5 cpm 2RE-2330 BMS Liquid Discharge Monitor 9.5E5 cpm 2RE-4423 Regenerative Waste Discharge Monitor 9.5E5 cpm 2RE-4425 SG Blowdown to Flume Radiation Monitor 9.5E5 cpm OR  3. Confirmed grab sample analyses for gaseous or liquid releases indicates concentrations or release rates > 200 times the applicable values of the ODCM for  15 minutes.
Deviations:
None.
to 0CAN121102
 
Page 13 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document Differences:
The radiation monitor channel is identified in EAL #1 in order to provide site-specific detail.
 
The second (superfluous) "reading" is deleted in the ANO EAL #2.
 
EAL #2 and its associated basis information are revised and a table added to provide site-
 
specific information for radiation monitors that may not be capable of providing values within the
 
calibrated range of the monitor at or above the 200 multiple for an alarm setpoint established by
 
a radioactivity discharge permit.
 
ANO has installed telemetered perimeter m onitoring devices. These devices, however, are neither qualified nor intended to provide an accurate indication of the dose rate at the perimeter
 
of the site. They are only intended to provide early indication of a potential unmonitored offsite
 
release. Therefore, ANO has not included an EAL comparable to NEI 99-01, Revision 5, AA1
 
EAL #4.
 
The Radiological Dose Assessment Computer Sy stem (RDACS) is ANO's current real-time dose assessment system, but is out-dated. RDA CS is currently scheduled to be replaced prior to implementation of the new EALs with a system that is not a "real-time" system; therefore, an EAL comparable to NEI 99-01, Revision 5, AU1 EAL #5 has not been included. to 0CAN121102
 
Page 14 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document NEI 99-01:  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 Mode Applicability:
All  Example Emergency Action Levels: (1 or 2)
: 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.
: 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)
 
ANO:  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 Mode Applicability: All  Emergency Action Level(s): (1 or 2)
: 1. A water level drop in the refueling canal or spent fuel pool that will result in irradiated fuel becoming uncovered.
OR    to 0CAN121102
 
Page 15 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document ANO:  AA2 (Cont'd)
: 2. VALID alarm on any of the following radiation monitors due to damage to irradiated fuel or loss of water level:
Unit 1 RX-9820 Containment Purge (Channel 7 or 9) RX-9825 Radwaste Area (Channel 7 or 9) RX-9830 Fuel Handling Area (Channel 7 or 9) RE-8060 Containment High Range Radiation Monitors RE-8061 Containment High Range Radiation Monitors RE-8009 Spent Fuel Area RE-8017 Fuel Handling Unit 2 2RX-9820 Containment Purge (Channel 7 or 9) 2RX-9825 Radwaste Area (Channel 7 or 9) 2RX-9830 Fuel Handling Area (Channel 7 or 9) 2RE-8905 Containment Equipment Hatch Area 2RE-8909 Containment Personnel Hatch Area 2RE-8925-1 Containment High Range Radiation Monitors 2RE-8925-2 Containment High Range Radiation Monitors 2RE-8914 Spent Fuel Area 2RE-8915 Spent Fuel Area 2RE-8916 Spent Fuel Area 2RE-8912 Containment Incore Inst.
Deviations:
None.
Differences:
Site-specific elevated readings for EAL #2 are not provided for ANO but instead multiple
 
radiation monitors for each unit on which an alarm could be received are provided. to 0CAN121102
 
Page 16 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document NEI 99-01:  AA3 Rise in radiation levels within the facility that impedes operation of systems required to maintain
 
plant safety functions
 
Operating Mode Applicability:
All  Example Emergency Action Levels:
: 1. 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)
 
ANO:  AA3 Rise in radiation levels within the facility that impedes operation of systems required to maintain
 
plant safety functions
 
Operating Mode Applicability:
All  Emergency Action Level(s):
: 1. Dose rate > 15 mR/hr in any of the following areas requiring continuous occupancy to maintain plant safety functions:
Unit 1 Control Room  Unit 2 Control Room  Central Alarm Station Deviations:
None.
Differences:
None.
to 0CAN121102
 
Page 17 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document NEI 99-01:  AS1 Off-site dose resulting from an actual or IMMINENT release of gaseous radioactivity greater
 
than 100 mrem Total Effective Dose Equivalent (TEDE) or 500 mrem Thyroid Committed Dose
 
Equivalent (CDE) for the actual or projected duration of the release
 
Operating Mode Applicability:
All  Example Emergency Action Levels:
(1 or 2 or 3 or 4)
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)
: 2. Dose assessment using actual meteorology indicates doses greater than 100 mrem TEDE or 500 mrem thyroid CDE at or beyond the site boundary.
: 3. VALID perimeter radiation monitoring system reading greater than 100 mR/hr for 15 minutes or longer. [for sites having telemetered perimeter monitors]
: 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. to 0CAN121102
 
Page 18 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document ANO:  AS1 Offsite dose resulting from an actual or imminent release of gaseous radioactivity > 100 mR
 
TEDE or 500 mR child thyroid CDE for the actual or projected duration of the release
 
Operating Mode Applicability: All  Emergency Action Level(s): (1 or 2 or 3)
Note: The SM / TSC Director / EOF 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, the
 
classification should be based on EAL #2 instead of EAL #1. Do not delay declaration awaiting dose assessment results.
: 1. VALID reading on Channel 9 on any of the following radiation monitors > the reading shown for  15 minutes:
MONITORS - UNIT 1 LIMIT RX-9820 Containment Purge 5.90E+1 &#xb5;Ci/cc RX-9825 Radwaste Area 5.36E+1 &#xb5;Ci/cc RX-9830 Fuel Handling Area 4.54E+1 &#xb5;Ci/cc RX-9835 Emergency Penetration Room 9.56E+2 &#xb5;Ci/cc MONITORS - UNIT 2 LIMIT 2RX-9820 Containment Purge 4.46E+1 &#xb5;Ci/cc 2RX-9825 Radwaste Area 3.32E+1 &#xb5;Ci/cc 2RX-9830 Fuel Handling Area 4.46E+1 &#xb5;Ci/cc 2RX-9835 Emergency Penetration Room 8.84E+2 &#xb5;Ci/cc 2RX-9840 Post Accident Sampling Building 4.42E+2 &#xb5;Ci/cc 2RX-9845 Aux. Building Extension 1.26E+2 &#xb5;Ci/cc 2RX-9850 Low-Level Radwaste Storage Building 1.77E+2 &#xb5;Ci/cc OR  2. Dose assessment using actual meteorology indicates doses > 100 mR TEDE or 500 mR child thyroid CDE at or beyond the site boundary.
OR    to 0CAN121102
 
Page 19 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document ANO:  AS1 (Cont'd)
: 3. Field survey results indicate closed window dose rates >100 mR/hr expected to continue for  60 minutes; or analyses of field survey samples indicate child thyroid CDE > 500 mR for one hour of inhalation, at or beyond the site boundary.
Deviations:
None.
Differences:
Child thyroid CDE is used in place of adult thyroid CDE in the IC and EALs #2 and #3 because
 
the State of Arkansas uses the child thyroid in the dose assessment methods. This difference
 
provides consistency with dose assessment methods used by the State of Arkansas.
 
The references to dose assessment and plant monitoring data are replaced in the EAL section
 
note with the corresponding specific EAL numbers. This change is provided for ease of use
 
only and does not change the intent of the note.
 
The radiation monitor channel is identified in EAL #1 in order to provide site-specific detail.
 
ANO has installed telemetered perimeter m onitoring devices. These devices, however, are neither qualified nor intended to provide an accurate indication of the dose rate at the perimeter
 
of the site. They are only intended to provide early indication of a potential unmonitored offsite
 
release. Therefore, ANO has not included an EAL comparable to NEI 99-01 Revision 5 AS1
 
EAL #3.
 
Additional information is provided in the bases for field monitoring team surveys (ANO EAL #3). to 0CAN121102
 
Page 20 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document NEI 99-01:  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 Mode Applicability:
All  Example Emergency Action Levels:
(1 or 2 or 3 or 4)
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)
: 2. Dose assessment using actual meteorology indicates doses greater than 1000 mrem TEDE or 5000 mrem thyroid CDE at or beyond the site boundary.
: 3. VALID perimeter radiation monitoring system reading greater than 1000 mR/hr for 15 minutes or longer. [for sites having telemetered perimeter monitors]
: 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. to 0CAN121102
 
Page 21 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document ANO:  AG1 Offsite dose resulting from an actual or imminent release of gaseous radioactivity > 1000 mR
 
TEDE or 5000 mR child thyroid CDE for the actual or projected duration of the release using
 
actual meteorology
 
Operating Mode Applicability: All  Emergency Action Level(s): (1 or 2 or 3)
Note: The SM / TSC Director / EOF 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, the
 
classification should be based on EAL #2 instead of EAL #1. Do not delay declaration awaiting dose assessment results.
: 1. VALID reading on Channel 9 on any of the following radiation monitors > the reading shown for  15 minutes:
MONITORS - UNIT 1 LIMIT RX-9820 Containment Purge 5.90E+2 (Ci/cc) RX-9825 Radwaste Area 5.36E+2 (Ci/cc) RX-9830 Fuel Handling Area 4.54E+2 (Ci/cc) RX-9835 Emergency Penetration Room 9.56E+3 (Ci/cc)  MONITORS - UNIT 2 LIMIT 2RX-9820 Containment Purge 4.46E+2 (Ci/cc) 2RX-9825 Radwaste Area 3.32E+2 (Ci/cc) 2RX-9830 Fuel Handling Area 4.46E+2 (Ci/cc) 2RX-9835 Emergency Penetration Room 8.84E+3 (Ci/cc) 2RX-9840 Post Accident Sampling Building 4.42E+3 (Ci/cc) 2RX-9845 Aux. Building Extension 1.26E+3 (Ci/cc) 2RX-9850 Low-Level Radwaste Storage Building 1.77E+3 (Ci/cc)  OR  2. Dose assessment using actual meteorology indicates doses > 1000 mR TEDE or 5000 mR child thyroid CDE at or beyond the site boundary.
OR  3. Field survey results indicate closed window dose rates >1000 mR/hr expected to  continue for  60 minutes; or analyses of field survey samples indicate child thyroid CDE > 5000 mR for one hour of inhalation, at or beyond the site boundary. to 0CAN121102
 
Page 22 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document ANO:  AG1 (Cont'd)
Deviations:
None.
Differences:
Child thyroid CDE is used in place of adult thyroid CDE in the IC and EALs #2 and #3 because
 
the State of Arkansas uses the child thyroid in their dose assessment methods. This difference
 
provides consistency with dose assessment methods used by the State of Arkansas.
 
The references to dose assessment and plant monitoring data are replaced with the
 
corresponding specific EAL numbers in the EAL section note. This change is provided for ease
 
of use only and does not change the intent of the note.
 
The radiation monitor channel is identified in EAL #1 in order to provide site-specific detail.
 
ANO has installed telemetered perimeter m onitoring devices. These devices, however, are neither qualified nor intended to provide an accurate indication of the dose rate at the perimeter
 
of the site. They are only intended to provide early indication of a potential unmonitored offsite
 
release. Therefore, ANO has not included an EAL comparable to NEI 99-01, Revision 5, AG1
 
EAL #3.
 
Additional information is provided in the bases for field monitoring team surveys (ANO EAL #3). to 0CAN121102
 
Page 23 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document NEI 99-01:  CU1 RCS leakage
 
Operating Mode Applicability:
Cold Shutdown Example Emergency Action Levels:
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. [
BWR] 
: 1. RCS leakage results in the inability to maintain or restore level within (site-specific pressurizer or RCS/RPV level target band) for 15 minutes or longer. [
PWR] 
 
ANO:  CU1 RCS leakage
 
Operating Mode Applicability:
Cold Shutdown (Mode 5)
Emergency Action Level(s):
Note: The SM 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 level within Pressurizer or RCS level target band for  15 minutes.
Deviations:
None.
Differences:
None. to 0CAN121102
 
Page 24 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document NEI 99-01:  CU2 UNPLANNED loss of RCS/RPV inventory
 
Operating Mode Applicability:
Refueling Example Emergency Action Levels:
(1 or 2)  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.
: 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).
 
ANO:  CU2 UNPLANNED loss of RCS / reactor vessel inventory
 
Operating Mode Applicability:
Refueling (Mode 6)
Emergency Action Level(s):  (1 or 2)
Note: The SM 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 / reactor vessel level drop as indicated by either of the following:
: a. RCS / reactor vessel water level drop below the reactor vessel flange for  15 minutes when the RCS / reactor vessel level band is established above the reactor vessel
 
flange. OR  b. RCS / reactor vessel water level drop below the RCS / reactor vessel level band for  15 minutes when the RCS / reactor vessel level band is established below the reactor vessel flange.
OR    to 0CAN121102
 
Page 25 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document ANO:  CU2 (Cont'd)
: 2. RCS / reactor vessel level cannot be monitored with a loss of RCS / reactor vessel inventory as indicated by an unexplained level rise in the Reactor Building Sump, Reactor
 
Drain Tank, Aux. Building Equipment Drain Tank, Aux. Building Sump, or Quench Tank.
Deviations:
None.
Differences:
None. to 0CAN121102
 
Page 26 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document NEI 99-01:  CU4 UNPLANNED loss of decay heat removal capability with irradiated fuel in the RPV
 
Operating Mode Applicability:
Cold Shutdown Refueling Example Emergency Action Levels:
(1 or 2)  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.
: 2. Loss of all RCS temperature and RCS/RPV level indication for 15 minutes or longer.
 
ANO:  CU3 UNPLANNED loss of decay heat removal capability with irradiated fuel in the reactor vessel
 
Operating Mode Applicability:
Cold Shutdown (Mode 5)
Refueling (Mode 6)
Emergency Action Level(s): (1 or 2)
Note: The SM 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 200 &deg;F.
 
OR  2. Loss of all RCS temperature and RCS/reactor vessel level indication for  15 minutes.
Deviations:
None.
Differences:
NEI 99-01 CU4 is renumbered to ANO CU3 for formatting purposes based on site preference for order of ICs alone. to 0CAN121102
 
Page 27 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document NEI 99-01:  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 Mode Applicability:
Cold Shutdown Refueling Example Emergency Action Level:
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.
ANO:  CU5 AC power capability to Vital 4.16 KV busses reduced to a single power source  15 minutes such that any additional single failure would result in station blackout
 
Operating Mode Applicability:
Cold Shutdown (Mode 5 Refueling (Mode 6)
Emergency Action Level(s):
Note: The SM 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 Vital 4.16 KV busses reduced to a single power source  15 minutes.
AND  b. Any additional single power source failure will result in station blackout.
Deviations:
None. Differences:
NEI 99-01 CU3 is renumbered to ANO CU5 for formatting purposes based on site preference for order of ICs alone.
The site-specific term "Vital 4.16 KV" is used in the IC and EAL to define emergency busses. to 0CAN121102
 
Page 28 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document NEI 99-01:  CU7 Loss of required DC power for 15 minutes or longer
 
Operating Mode Applicability:
Cold Shutdown Refueling Example Emergency Action Level:
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.
 
ANO:  CU6 Loss of required DC power  15 minutes Operating Mode Applicability:
Cold Shutdown (Mode 5)
Refueling (Mode 6)
Emergency Action Level(s):
Note: The SM 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. < 105 volts on required Vital DC bus  15 minutes.
Deviations:
None.
Differences:
NEI 99-01 CU7 is renumbered to ANO CU6 for formatting purposes based on site preference for order of ICs alone. to 0CAN121102
 
Page 29 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document NEI 99-01:  CU8 Inadvertent criticality
 
Operating Mode Applicability:
Cold Shutdown Refueling Example Emergency Action Levels:
: 1. UNPLANNED sustained positive period observed on nuclear instrumentation. (BWR)
: 1. UNPLANNED sustained positive startup rate observed on nuclear instrumentation. (PWR)
 
ANO:  CU7 Inadvertent criticality
 
Operating Mode Applicability: Cold Shutdown (Mode 5)
Refueling (Mode 6)
Emergency Action Level(s):
: 1. UNPLANNED sustained positive startup rate observed on nuclear instrumentation.
 
Deviations:
None.
Differences:
NEI 99-01 CU8 is renumbered to ANO CU7 for formatting purposes based on site preference for order of ICs alone. to 0CAN121102
 
Page 30 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document NEI 99-01:  CU6 Loss of all On-site or Off-site communications capabilities
 
Operating Mode Applicability:
Cold Shutdown Refueling
 
Defueled  Example Emergency Action Levels:
(1 or 2) 
: 1. Loss of all of the following on-site communication methods affecting the ability to perform routine operations:
(site-specific list of communications methods)
: 2. Loss of all of the following off-site communication methods affecting the ability to perform offsite notifications:
(site-specific list of communications methods)
 
ANO:  CU8 Loss of all onsite or offsite communications capabilities
 
Operating Mode Applicability:
Cold Shutdown (Mode 5)
Refueling (Mode 6)
 
Defueled  Emergency Action Level(s): (1 or 2)
: 1. Loss of all Table C2 onsite communication methods affecting the ability to perform routine operations.
OR  2. Loss of all Table C3 offsite communication methods affecting the ability to perform offsite notifications.
Table C2 Onsite Communications Methods  Table C3 Offsite Communications Methods Station radio system  All telephone lines (commercial and microwave) Plant paging system  Emergency Notification System (ENS) In-plant telephones  Gaitronics      to 0CAN121102
 
Page 31 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document ANO:  CU8 (Cont'd)
Deviations:
None.
Differences:
NEI 99-01 CU6 is renumbered to ANO CU8 for formatting purposes based on site preference for order of ICs alone.
 
Onsite and offsite communications methods in tables are presented and the tables are
 
referenced causing a minor difference in EAL language from that in NEI 99-01. to 0CAN121102
 
Page 32 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document NEI 99-01:  CA1 Loss of RCS/RPV inventory
 
Operating Mode Applicability:
Cold Shutdown Refueling Example Emergency Action Levels:
(1 or 2)  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)
]  [Bottom ID of the RCS loop (PWR)
] 
: 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).
 
ANO:  CA1 Loss of RCS / reactor vessel inventory
 
Operating Mode Applicability:
Cold Shutdown (Mode 5)
Refueling (Mode 6)
Emergency Action Level(s): (1 or 2)
Note: The SM / TSC Director / EOF 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 / reactor vessel inventory as indicated by:
 
Unit 1:  RVLMS Levels 1 through 8 indicate DRY Unit 2:  RVLMS Levels 1 through 5 indicate DRY OR  Unit 1:  Reactor vessel level < 368 ft., 0 in. (bottom of the hot leg)
Unit 2:  Reactor vessel level < 369 ft., 1.5 in. (bottom of the hot leg)
OR    to 0CAN121102
 
Page 33 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document ANO:  CA1 (Cont'd)
: 2. RCS / reactor vessel level cannot be monitored for  15 minutes with a loss of RCS /
reactor vessel inventory as indicated by an unexplained level rise in the Reactor Building
 
Sump, Reactor Drain Tank, Auxiliary Buildi ng Equipment Drain Tank, Auxiliary Building Sump, or Quench Tank.
Deviations:
None.
Differences:
Site-specific bases information is provided for reactor vessel level monitoring in Mode 6. to 0CAN121102
 
Page 34 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document NEI 99-01:  CA4 Inability to maintain plant in cold shutdown
 
Operating Mode Applicability:
Cold Shutdown Refueling Example Emergency Action Levels:
(1 or 2) 
: 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. Table:  RCS Reheat Duration Thresholds RCS Containment Closure Duration Intact (but not RCS Reduced Inventory [PWR]) N/A 60 minutes* Established 20 minutes*
Not intact or RCS Reduced Inventory (PWR) Not Established 0 minutes
* If an RCS heat removal system is in operation within this time frame and RCS temperature is being reduced, the EAL is not applicable.
: 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)    to 0CAN121102
 
Page 35 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document ANO:  CA3  Inability to maintain plant in cold shutdown
 
Operating Mode Applicability:
Cold Shutdown (Mode 5)
Refueling (Mode 6)
Emergency Action Level(s): (1 or 2)
: 1. An UNPLANNED event results in RCS temperature > 200 &deg;F > the specified duration in Table C1. Table C1 RCS Reheat Duration Thresholds RCS Containment Closure Duration Intact (but not RCS Lowered Inventory) N/A 60 minutes* Established 20 minutes*
Not intact or RCS Lowered Inventory Not Established 0 minutes
* If an RCS heat removal system is in operation within this time frame and RCS temperature is being reduced, the EAL is not applicable.
OR  Note: EAL #2 does not apply in solid plant conditions.
: 2. An UNPLANNED event results in RCS pressure rise > 10 psi due to a loss of RCS cooling.
 
Deviations:
None.
Differences:
NEI 99-01 CA4 is renumbered to ANO CA3 for formatting purposes based on site preference for order of ICs alone. to 0CAN121102
 
Page 36 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document NEI 99-01:  CA3 Loss of all Off-site and all On-Site AC power to emergency busses for 15 minutes or longer
 
Operating Mode Applicability:
Cold Shutdown Refueling
 
Defueled  Example Emergency Action Level:
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.
 
ANO:  CA5 Loss of all offsite and all onsite AC power to Vital 4.16KV busses  15 minutes Operating Mode Applicability:
Cold Shutdown (Mode 5)
Refueling (Mode 6)
 
Defueled Emergency Action Level(s):
Note: The SM / TSC Director / EOF 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 offsite and all onsite AC power to Vital 4.16KV busses  15 minutes.
Deviations:
None.
Differences:
NEI 99-01 CA3 is renumbered to ANO CA5 for formatting purposes based on site preference for order of ICs alone.
 
The site-specific term "Vital 4.16KV" is used in the IC and EAL to define emergency busses. to 0CAN121102
 
Page 37 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document NEI 99-01:  CS1 Loss of RCS/RPV inventory affecting core decay heat removal capability
 
Operating Mode Applicability:
Cold Shutdown Refueling Example Emergency Action Levels:
(1 or 2 or 3)
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 Emergency Core Cooling System (ECCS) actuation setpoint (BWR)
]  OR  2. With CONTAINMENT CLOSURE established, RCS/RPV level less than (site-specific level for the top of active fuel (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).
 
to 0CAN121102
 
Page 38 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document ANO:  CS1  Loss of RCS / reactor vessel inventory affecting core decay heat removal capability
 
Operating Mode Applicability:
Cold Shutdown (Mode 5)
Refueling (Mode 6)
Emergency Action Level(s):
(1 or 2 or 3)
Note: The SM / TSC Director / EOF 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:
Unit 1:  Reactor Vessel Level Monitor System (RVLMS) Levels 1 through 9 indicate DRY Unit 2:  RVLMS Levels 1 through 6 indicate DRY OR  2. With CONTAINMENT CLOSURE established, core exit thermocouples (CETs) indicate superheat.
OR  3. RCS / reactor vessel level cannot be monitored for  30 minutes with a loss of RCS /
reactor vessel inventory as indicated by any of the following:
Containment High Range Radiation Monitor reading >10 R/hr  Erratic source range monitor indication  Unexplained level rise in Reactor Building Sump, Reactor Drain Tank, Quench Tank, Aux. Building Equipment Drain Tank, or Aux. Building Sump.
Deviations:
None.
to 0CAN121102
 
Page 39 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document ANO:  CS1 (Cont'd)
Differences:
The ANO units do not have the capability to monitor reactor vessel level at or below the top of
 
active fuel and, therefore, superheat indication on core exit thermocouples (CETs) is used for
 
EAL #2. CET superheat is used to indicate the level below the top of active fuel and core
 
uncovery. As level falls below the top of active fuel, CETs will begin to indicate superheat
 
conditions. This difference of using CET superheat rather than an actual level indication
 
corresponding to the NEI EAL is due to plant design. The treatment of this EAL provides a logic
 
path consistent with a Site Area Emergency in that the RCS level drop indicates a loss of the
 
RCS barrier and the superheat condition indicates a potential loss (or loss) of the fuel clad
 
barrier. Escalation to a General Emergency occurs if containment closure is not established or
 
other indications of containment barrier loss or potential loss exist. The plant's cold shutdown
 
RCS level monitoring capability represents the same technical capability as that for the
 
NEI 99-01, Revision 4, based ANO EAL scheme that was previously approved by NRC in a letter dated October 6, 2005 (ADAMS Accession No. ML052720568). The current proposed use of the CET superheat indication in CS1 is consistent with the use of this indication in CG1 in
 
the approved EAL scheme. to 0CAN121102
 
Page 40 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document NEI 99-01:  CG1 Loss of RCS/RPV inventory affecting fuel clad integrity with containment challenged
 
Operating Mode Applicability:
Cold Shutdown Refueling Example Emergency Action Level:
(1 or 2)  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):
: 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):
Table:  Containment Challenge Indications CONTAINMENT CLOSURE not established.  (Site-specific explosive mixture) inside containment. UNPLANNED rise in containment pressure. Secondary containment radiation monitor reading above (site-specific value). [
BWR only]    to 0CAN121102
 
Page 41 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document ANO:  CG1  Loss of RCS / reactor vessel inventory affecting fuel clad integrity with containment challenged
 
Operating Mode Applicability:
Cold Shutdown (Mode 5)
Refueling (Mode 6)
Emergency Action Level(s):
(1 or 2)  Note: The SM / TSC Director / EOF 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. Core Exit Thermocouples indicate superheat for  30 minutes.
AND  b. Any of the following containment challenge indications:
CONTAINMENT CLOSURE not established  Explosive mixture inside containment  UNPLANNED rise in containment pressure OR  2. a. RCS / reactor vessel level cannot be monitored with core uncovery indicated by any of the following for  30 minutes:
Containment High Range Radiation Monitor reading > 10R/hr  Erratic source range monitor indication  Unexplained level rise in Reactor Building Sump, Reactor Drain Tank, Quench Tank, Auxiliary Building Equipment Drain Tank, or Auxiliary Building Sump AND  b. Any of the following containment challenge indications:
CONTAINMENT CLOSURE not established  Explosive mixture inside containment  UNPLANNED rise in containment pressure Deviations:
None. to 0CAN121102
 
Page 42 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document ANO:  CG1 (Cont'd)
Differences:
The ANO units do not have the capability to monitor reactor vessel level at or below the top of
 
active fuel and therefore uses superheat indication on CETs for EAL #1. CET superheat is
 
used to indicate level below the top of active fuel and core uncovery. As level falls below the
 
top of active fuel, CETs will begin to indicate superheat conditions. This difference of using
 
CET superheat rather than an actual level indication corresponding to the NEI EAL is because
 
of plant design. The plant's cold shutdown RCS level monitoring capability represents the same
 
technical capability as that for the NEI 99-01, Revision 4, based ANO EAL scheme that was
 
previously approved by NRC in a letter dated October 6, 2005 (ADAMS Accession No.
 
ML052720568). The current proposed use of the CET superheat indication, though formatted
 
differently, is consistent with its us e in the previously approved EAL scheme.
 
The term "unexplained" is used instead of "UNPLANNED" for the level rise indication in the third
 
bullet of EAL 2.a. This is the same term used by NEI 99-01 for the Site Area Emergency (SAE)
 
condition. If level rise cannot be explained, then it encompasses the term "UNPLANNED" and
 
therefore meets the NEI intent.
 
A table is not used to present the containment challenge indications for EALs 1 and 2.b. The
 
NEI indications are presented in bullet format with all NEI content retained in the ANO EAL. to 0CAN121102
 
Page 43 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document NEI 99-01:  E-HU1 Damage to a loaded cask CONFINEMENT BOUNDARY
 
Operating Mode Applicability:
Not applicable Example Emergency Action Level:
: 1. Damage to a loaded cask CONFINEMENT BOUNDARY.
 
ANO:  E-HU1 Damage to a loaded cask CONFINEMENT BOUNDARY
 
Operating Mode Applicability:
All  Emergency Action Level(s):
: 1. Damage to a loaded cask CONFINEMENT BOUNDARY.
 
Deviations:
None.
Differences:
An operating mode applicability of "all" is used vice the NEI designation of "N/A."  The net effect
 
is that this event is applicable regardless of operating mode and therefore, the same as the NEI
 
intent. to 0CAN121102
 
Page 44 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document NEI 99-01:  FU1 ANY Loss or ANY Potential Loss of Containment
 
Operating Mode Applicability:
Power Operations Startup Hot Standby
 
Hot Shutdown
 
ANO:  FU1 ANY loss or ANY potential loss of containment
 
Operating Mode Applicability:
Power Operations (Mode 1)
Startup (Mode 2)
 
Hot Standby (Mode 3)
 
Hot Shutdown (Mode 4)
Deviations:
None.
Differences:
None. to 0CAN121102
 
Page 45 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document NEI 99-01:  FA1 ANY Loss or ANY Potential Loss of EITHER Fuel Clad OR RCS
 
Operating Mode Applicability:
Power Operations Startup Hot Standby
 
Hot Shutdown
 
ANO:  FA1 ANY loss or ANY potential loss of EITHER fuel clad or RCS
 
Operating Mode Applicability:
Power Operations (Mode 1)
Startup (Mode 2)
 
Hot Standby (Mode 3)
 
Hot Shutdown (Mode 4)
Deviations:
None.
Differences:
None. to 0CAN121102
 
Page 46 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document NEI 99-01:  FS1 Loss or Potential Loss of ANY Two Barriers
 
Operating Mode Applicability:
Power Operations Startup Hot Standby
 
Hot Shutdown
 
ANO:  FS1 Loss or potential loss of ANY two barriers
 
Operating Mode Applicability:
Power Operations (Mode 1)
Startup (Mode 2)
 
Hot Standby (Mode 3)
 
Hot Shutdown (Mode 4)
Deviations:
None.
Differences:
None. to 0CAN121102
 
Page 47 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document NEI 99-01:  FG1 Loss of ANY Two Barriers AND Loss or Potential Loss of the third barrier
 
Operating Mode Applicability:
Power Operations Startup Hot Standby
 
Hot Shutdown
 
ANO:  FG1 Loss of ANY two barriers AND loss or potential Loss of the third barrier
 
Operating Mode Applicability:
Power Operations (Mode 1)
Startup (Mode 2)
 
Hot Standby (Mode 3)
 
Hot Shutdown (Mode 4)
Deviations:
None.
Differences:
None. to 0CAN121102
 
Page 48 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document NEI 99-01:  FUEL CLAD BARRIER Fuel Clad Barrier Thresholds LOSS POTENTIAL LOSS
: 1. Critical Safety Function Status A. Core Cooling - Red Entry Conditions MetA. Core Cooling - Orange Entry C onditions Met OR B. Heat Sink - Red Entry Conditions Met OR 2. Primary Coolant Activity Level A. Coolant activity greater than (site-specific) value Not Applicable OR 3. Core Exit Thermocouple Readings A. Core exit thermocouples reading greater than (site-specific &deg;F) A. Core exit thermocouples reading greater t han (site-specific &deg;F)
OR 4. Reactor Vessel Water Level Not Applicable A. RCS/RPV level le ss than (site-specific level for TOAF)
OR 5. Not Applicable Not Applicable Not Applicable OR 6. Containment Radiation Monitoring A. Containment radiation monitor reading greater than (site-specific value)
Not Applicable OR 7. Other Site-Specific Indications A. (site-specific) as applicable A. (site-specific) as applicable OR 8. Emergency Director Judgment A. Any condition in the opinion of the Emergency Director that indicates Loss of
 
the Fuel Clad Barrier A. Any condition in the opinion of the Emergency Director that indicates Po tential Loss of the Fuel Clad Barrier    to 0CAN121102
 
Page 49 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document ANO:  FUEL CLAD BARRIER Fuel Clad Barrier EALs LOSS POTENTIAL LOSS 1. Safety Function Status (FCB1)
Not Applicable Not Applicable 1. Primary Activity Level (FCB1)
: 1. Coolant activity > 300 &#xb5;Ci/gm dose equivalent I-131 activity by Chemistry
 
sample OR 2. Radiation levels > 1000 MR/hr Unit 1: at SA-229 Unit 2: at 2TCD-19 None 2. Core Exit Thermocouple Readings (FCB2)
> 1200 &deg;F CET temperature Unit 1: Inadequate Core Cooling (ICC) exists as evidenced by CETs indicating s uperheated conditions Unit 2: Average CETs indicate superheat for current RCS pressure 3. Reactor Vessel Water Level (FCB3)
None Unit 1: RVLMS Levels 1 through 9 indicate DRY Unit 2: RVLMS Levels 1 through 7 indicate DRY
: 4. Containment Radiation Monitoring (FCB4)
Containment high range radiation monitor reading > 1000 R/hr None 5. Core Damage Assessment (FCB5)
At least 5% fuel clad damage as determined from core damage assessment None 6. Emergency Director Judgment (FCB6)
Any condition in the opinion of the SM / TSC Director
/ EOF Director that indicates Loss or Potential Loss of the Fuel Clad Barrier to 0CAN121102
 
Page 50 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document ANO:  FUEL CLAD BARRIER (Cont'd)
Deviations:
None. Differences:
EALs for Unit 1 and Unit 2 that correspond to the Critical Safety Function Status Trees (CSFSTs) included in the NEI 99-01, Revision 5, EALs are not provided. Unit 1 (a B&W plant)
 
does not have a Safety Function process similar to that of Westinghouse units (which are
 
included in the NEI 99-01, Revision 5, example EALs).
 
Unit 2 (a CE plant) does have methodologies similar to CSFSTs with some differences. Unit 2
 
performs Standard Post Trip Actions (SPTA) and verifies the status of its Safety Functions (Reactivity, Vital Auxiliaries, RCS Inventory, RCS Pressure, Core Heat Removal, RCS Heat Removal, & Containment) upon any reactor trip. An evaluation is made of each of the seven
 
Safety Functions comparing plant response and critical parameters to standard, expected
 
values. Each Safety Function is then marked on a tracking sheet as "SATISIFIED" or "NOT
 
SATISIFIED."  If any Safety Function is not satisfied, the condition is announced to the Control
 
Room staff and a diagnostic flow chart is referenced to determine the proper Optimal Recovery
 
Procedure (ORP) to enter. The diagnostic flowchart may direct the Operator to an ORP or to the
 
Functional Recovery Procedure (FRP). The intent of this diagnostic action is to direct the
 
Operator to the ORPs for a single event, and to the FRP for multiple events. Each ORP
 
contains Safety Function Status Checks (SFSC) which are performed every 15 minutes. These
 
checks ensure the Operator's utilization of the ORP is properly addressing plant critical
 
parameters. If the SFSC is not met then the FRP is entered. Criteria for FRP entry is:
: 1. ANY event in progress which can NOT be diagnosed as a single event. 2. Actions taken have NOT satisfied SFSC acceptance criteria.
: 3. Entry is directed by Diagnostic Actions.
 
While all ORPs have SFSC criteria for all Safety Functions, all ORP acceptance criteria are not
 
the same. For example; the SFSC criteria required to SATISIFY the requirements for RCS Heat
 
Removal (a Safety Function) is different in the SPTAs than in Loss of Coolant Accident (LOCA)
 
ORP, which is different than those contained in the Steam Generator Tube Rupture (SGTR)
 
ORP, which is different than those in the Loss of Feedwater (LOF) ORP, and so on.
 
To compare NEI 99-01 Safety Function intent with that of Unit 2 above, a Loss of Offsite Power (LOOP) event is correctly classified per NEI 99-01 as System Malfunction - Loss of AC Power (SU1), "Loss of all offsite AC power to Vital 4.16 KV busses  15 minutes."  This event is correctly classified as a NUE. However, a LOOP results in the loss of the operating Reactor
 
Coolant Pumps due to loss of non-vital power sources. During performance of the SPTAs, the
 
Core Heat Removal Safety Function will be assessed as "NOT SATISIFIED."  If NEI 99-01, Fission Product Barrier Malfunction - Barriers, is referenced, FCB1 would be applicable and the
 
event would require declaration of an ALERT.
 
to 0CAN121102
 
Page 51 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document ANO:  FUEL CLAD BARRIER (Cont'd)
Differences (Cont'd):
Entergy believes erroneous classification of a standard NUE condition as a ALERT is
 
inappropriate. Natural circulation of the reactor coolant provides adequate core cooling and no
 
significant challenge to the core heat removal functions are expected to occur. Other EALs are
 
referenced when plant conditions change or degrade. The remaining EALs used in the Fuel
 
Clad Barrier (FCB) section provide the Operators with the necessary information to classify events appropriately based on the actual coolant temperatures, pressures, vessel level, margin-
 
to-sat values, etc., without the need for a reference to Safety Functions.
 
In summary, the ORP SFSC criteria were not established to meet EAL classification
 
requirements. The SFCSs ensure Safety Function status is verified and updated at regular
 
intervals so that changes in plant conditions may be recognized promptly and to enable trending
 
of plant parameters important to safety. Based on the above, Entergy proposes to not adopt the
 
use of Safety Functions in the Fuel Clad Barrier EAL. Subsequently, the NEI 99-01 FCBs are
 
renumbered in the ANO EALs for formatting purposes based on the non-use of the Safety
 
Function Status FCB criteria.
 
NEI provides a developer's note indicating that radiation levels observed on a sample may also
 
be used for the primary coolant activity level EAL. Radiation levels are provided at one foot
 
from the sample lines for this value. This is an equivalent use of the developer's note on
 
sample radiation levels.
 
Additional bases information for CET temperature readings are provided to assist the decision
 
maker in relating the associated potential losses and losses for all three fission product barriers.
 
ANO differs from NEI 99-01 guidance in that the potential loss reactor vessel level EAL does not
 
represent the top of active fuel (TOAF) level, but a level above it. The RVLMS at ANO does not
 
provide positive indication of core uncovery.
The above core level indication provided is used to monitor the approach to and recovery from Inadequate Core Cooling conditions, but the CETs
 
are used to identify core uncovery and are the only positive indication of core uncovery. The
 
ANO EAL represents the lowest point that can be monitored that is above the top of active fuel. 
 
This difference represents an appropriately conservative value that continues to indicate a
 
significant challenge in the ability to adequately cool the fuel cladding, is readily observable to
 
the operators, and represents the lowest level that can be measured using installed
 
instrumentation in the reactor vessel. Therefore, this difference is in accordance with plant
 
design.
 
Site-specific detail is provided in the bases for RVLMS.
 
An additional EAL labeled "core damage assessm ent (ANO FCB5)" is provided where NEI provides for "other indications."
 
The Fuel Clad Barrier EALs are not provided in a table in the Basis Document. The EALs are
 
presented as text. A table is used in the EAL Matrix document.
 
to 0CAN121102
 
Page 52 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document NEI 99-01:  RCS BARRIER RCS Barrier Example Thresholds LOSS POTENTIAL LOSS
: 1. Critical Safety Function Status Not Applicable A. RCS Integrit y - Red Entry Conditions Met OR B. Heat Sink - Red Entry Conditions Met OR 2. RCS Leak Rate A. RCS leak rate greater than available makeup capacity as indicated by a loss of
 
RCS subcooling B. RCS leak rate indicated greater than (site-specific capacity of once charging pump in the normal
 
charging mode) with Letdown isolated OR 3. Not Applicable Not Applicable Not Applicable OR 4. SG Tube Rupture A. RUPTURED Steam Generator (SG) results in an ECCS (SI) actuation Not Applicable OR 5. Not Applicable Not Applicable Not Applicable OR 6. Containment Radiation Monitoring A. Containment radiation monitor reading greater than (site-specific value)
Not Applicable OR 7. Other Site-Specific Indications A. (site-specific) as applicable A. (site-specific) as applicable 8. Emergency Director Judgment A. Any condition in the opinion of the Emergency Director that indicates Loss of
 
the RCS Barrier A. Any condition in the opinion of the Emergency Director that indicates Potential Loss of the RCS Barrier    to 0CAN121102
 
Page 53 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document ANO:  RCS BARRIER RCS Barrier EALs LOSS POTENTIAL LOSS 1. Safety Function Status (RCB1)
Not Applicable Not applicable
: 1. RCS Leak Rate (RCB1)
RCS leak rate > available makeup capacity as indicated by:
 
Unit 1: Loss of adequate subcooling margin  Unit 2: RCS subcooling (margin-to-saturation or MTS) can NOT
 
be maintained at least 30 &deg;F Unit 1: UNISOLABLE RCS leak > 50 gpm with Letdown isolated Unit 2: UNISOLABLE RCS leak > 44 gpm with Letdown isolated
: 2. SG Tube Rupture (RCB2)
Steam Generator Tube Rupture (SGTR) that results in an ECCS Safety
 
Injection (SI) actuation None 3. Containment Radiation Monitoring (RCB3)
Containment high range radiation monitor reading >
100 R/hr None 4. Emergency Director Judgment (RCB4)
Any condition in the opinion of the SM / TSC Director / EOF Director that indicates Loss or Potential Loss of the RCS Barrier.
Deviations:
None.
Differences:
EALs for Unit 1 and Unit 2 that correspond to the Critical Safety Function Status Trees (CSFSTs) included in the NEI 99-01, Revision 5, EALs are not provided. Unit 1 (a B&W plant)
 
does not have a Safety Function process similar to that of Westinghouse units (which are
 
included in the NEI 99-01, Revision 5, example EALs).
 
to 0CAN121102
 
Page 54 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document ANO: RCS BARRIER (Cont'd)
Differences (cont'd):
Unit 2 (a CE plant) does have methodologies similar to CSFSTs with some differences. Unit 2
 
performs Standard Post Trip Actions (SPTA) and verifies the status of its Safety Functions (Reactivity, Vital Auxiliaries, RCS Inventory, RCS Pressure, Core Heat Removal, RCS Heat Removal, & Containment) upon any reactor trip. An evaluation is made of each of the seven
 
Safety Functions comparing plant response and critical parameters to standard, expected
 
values. Each Safety Function is then marked on a tracking sheet as "SATISIFIED" or "NOT
 
SATISIFIED."  If any Safety Function is not satisfied, the condition is announced to the Control
 
Room staff and a diagnostic flow chart is referenced to determine the proper Optimal Recovery
 
Procedure (ORP) to enter. The diagnostic flowchart may direct the Operator to an ORP or to the
 
Functional Recovery Procedure (FRP). The intent of this diagnostic action is to direct the
 
Operator to the ORPs for a single event, and to the FRP for multiple events. Each ORP
 
contains Safety Function Status Checks (SFSC) which are performed every 15 minutes. These
 
checks ensure the Operator's utilization of the ORP is properly addressing plant critical
 
parameters. If the SFSC is not met then the FRP is entered. Criteria for FRP entry is:
: 1. ANY event in progress which can NOT be diagnosed as a single event. 2. Actions taken have NOT satisfied SFSC acceptance criteria.
: 3. Entry is directed by Diagnostic Actions.
 
While all ORPs have SFSC criteria for all Safety Functions, all ORP acceptance criteria are not
 
the same. For example; the SFSC criteria required to SATISIFY the requirements for RCS Heat
 
Removal (a Safety Function) is different in the SPTAs than in Loss of Coolant Accident (LOCA)
 
ORP, which is different than those contained in the Steam Generator Tube Rupture (SGTR)
 
ORP, which is different than those in the Loss of Feedwater (LOF) ORP, and so on.
 
To compare NEI 99-01 Safety Function intent with that of Unit 2 above, a Loss of Offsite Power (LOOP) event is correctly classified per NEI 99-01 as System Malfunction - Loss of AC Power (SU1), "Loss of all offsite AC power to Vital 4.16 KV busses  15 minutes."  This event is correctly classified as a NUE. However, a LOOP results in the loss of the operating Reactor
 
Coolant Pumps due to loss of non-vital power sources. During performance of the SPTAs, the
 
Core Heat Removal Safety Function will be assessed as "NOT SATISIFIED."  If NEI 99-01, Fission Product Barrier Malfunction - Barriers, is referenced, FCB1 would be applicable and the
 
event would require declaration of an ALERT.
 
Entergy believes erroneous classification of a standard NUE condition as a ALERT is
 
inappropriate. Natural circulation of the reactor coolant provides adequate core cooling and no
 
significant challenge to the core heat removal functions are expected to occur. Other EALs are
 
referenced when plant conditions change or degrade. The remaining EALs used in the RCS
 
Barrier (RCB) section provide the Operators with the necessary information to classify events appropriately based on the actual radiation levels, RCS leakage, margin-to-sat values, etc.,
without the need for a reference to Safety Functions. to 0CAN121102
 
Page 55 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document ANO: RCS BARRIER (Cont'd)
Differences (cont'd):
In summary, the ORP SFSC criteria were not established to meet EAL classification
 
requirements. The SFCSs ensure Safety Function status is verified and updated at regular
 
intervals so that changes in plant conditions may be recognized promptly and to enable trending
 
of plant parameters important to safety. Based on the above, Entergy proposes to not adopt the
 
use of Safety Functions in the RCS Barrier EAL. Subsequently, the NEI 99-01 RCBs are
 
renumbered in the ANO EALs for formatting purposes based on the non-use of the Safety
 
Function Status RCB criteria.
 
The ANO Unit 2 design uses three positive displacement charging pumps for normal RCS
 
makeup. The plant design specifics of the positiv e displacement pump is that the flow from a charging pump is not variable, but provides 44 gpm discrete flow and flow rises or lowers based
 
on the number of pumps in service. Therefore, ANO uses 44 gpm for the EAL vice the
 
NEI 99-01 developer's note guidance of 50 gpm for plants with low capacity charging pumps. 
 
The use of this 44 gpm value supports the NEI basis statement that "additional charging pumps
 
being required is indicative of a substantial RCS leak."
 
The term "loss of adequate subcooling margin" is used for Unit 1 and the inability to maintain
 
RCS subcooling at least 30 &deg;F for Unit 2 to indicate a loss of subcooling.
 
The term "unisolable" is included for the potential loss RCS leakage description for consistency
 
of EAL application. Use of this term provides clarity and does not alter intent.
 
Site-specific detail in the bases is provided to describe makeup systems and to clarify the
 
indications for RCS leakage potential loss.
 
The term "SGTR" for steam generator tube rupture instead of the NEI term "RUPTURED SG" is
 
used.  "SGTR" is a term readily recognized by the Operations staff and has the same meaning
 
as the NEI term.
 
The EALs refer to the ECCS (SI) actuation for ANO RCB2 indicator as being caused either
 
manually or automatically in the bases to clarify the information presented.
 
Site-specific information is provided in the bases regarding containment high range monitor
 
readings during the initial fifteen minutes after a thermal event.
 
Additional EALs that represent a loss or potential loss of the RCS barrier were not identified for
 
inclusion in the EAL scheme for this IC. A review of EOPs and station procedures was
 
performed to ensure additional EAL thresholds should not be considered in the "Other" category. This conforms to NEI guidance because an appropriately diverse mix of EALs is
 
provided and the NEI 99-01 guidance does not specify that any particular additional EALs be
 
provided, but specifies "as applicable."
 
The Reactor Coolant Barrier EALs are not provided in a table in the Basis Document. The EALs
 
are presented as text. A table is used in the EAL Matrix document. to 0CAN121102
 
Page 56 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document NEI 99-01:  CONTAINMENT BARRIER Containment Barrier Example Thresholds LOSS POTENTIAL LOSS
: 1. Critical Safety Function Status Not Applicable A. Containment - Red Entry Conditions Met OR 2. Containment Pressure A. A containment pressure rise followed by a rapid unexplained drop in
 
containment pressure OR B. Containment pressure or sump level response not consistent with Loss of
 
Coolant Accident (LOCA) conditions  A. Containment pressure greater than (site-specific value) and rising OR B. Explosive mixture exists inside containment OR C. a. Pressure greater than containment depressurization actuation setpoint AND b. Less than one full train of depressurization equipment operating OR 3. Core Exit Thermocouple Readings Not Applicable A. a. Core exit thermocouples in excess of (site-specific) &deg;F AND b. Restoration procedures not effective within 15 minutes OR B. a. Core exit thermocouples in excess of (site-specific) &deg;F AND b. Reactor vessel level below (site-specific level) AND c. Restoration procedures not effective within 15 minutes OR    to 0CAN121102
 
Page 57 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document NEI 99-01:  CONTAINMENT BARRIER (Cont'd)
Containment Barrier Example Thresholds LOSS POTENTIAL LOSS
: 4. SG Secondary Side Release With P-to -S Leakage A. RUPTURED SG is also FAULTED outside of containment OR B. a. Primary-to-Secondary leakrate greater than 10 gpm AND b. UNISOLABLE steam release from affected SG to the environment Not Applicable OR 5. Containment Isolation Failure or Bypass A. a. Failure of all valves in any one line to close AND b. Direct downstream path to the environment exists after
 
containment isolation signal Not Applicable OR 6. Containment Radiation Monitoring Not Applicable A. Containment radiation monitor reading greater than  (site-specific value)
OR 7. Other Site-Specific) Indications A. (site-specific) as applicable A. (site-specific) as applicable OR 8. Emergency Director Judgment A. Any condition in the opinion of the Emergency Director that indicates
 
Loss of the Containment Barrier A. Any condition in the opinion of the Emergency Director that indicates Potential Loss of the
 
Containment Barrier to 0CAN121102
 
Page 58 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document ANO: CONTAINMENT BARRIER Containment Barrier EALs LOSS POTENTIAL LOSS 1. Safety Function Status (CNB1)
Not Applicable Not applicable
: 1. Containment Pressure (CNB1)
: 1. Rapid unexplained drop in containment pressure following an
 
initial rise in containment pressure OR 2. Containment pressure or sump level response not consistent with
 
LOCA conditions
: 1. Unit 1: Containment pressure 73.7 PSIA (59 PSIG) and rising Unit 2: Containment pressure 73.7 PSIA and rising (59 PSIG)
OR 2. Explosive mixture exists inside Containment OR 3. a. Containment Pressure > containment spray actuation setpoint Unit 1: 44.7 PSIA (30 PSIG)
Unit 2: 23.3 PSIA (8.6 PSIG)
AND b. LESS THAN one full train of spray operating
: 2. Core Exit Thermocouple Readings (CNB2)
None 1. a. CETs indicate > 1200 &deg;F AND b. Restoration procedures not effective within 15 minutes OR 2. a. CETs indicate > 700 &deg;F AND b. RVLMS indicates:
Unit 1: Levels 1 through 9 DRY Unit 2: Levels 1 through 7 DRY AND c. Restoration procedures not effective within 15 minutes    to 0CAN121102
 
Page 59 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document ANO: CONTAINMENT BARRIER (Cont'd)
Containment Barrier EALs LOSS POTENTIAL LOSS 3. SG Secondary Side Release With Primary-to-Secondary Leakage (CNB3)
: 1. RUPTURED steam generator is also FAULTED outside of
 
Containment OR 2. a. Primary-to-secondary leakrate
>10 gpm AND b. UNISOLABLE steam release from affected steam generator
 
to the environment None 4. Containment Isolation Failure or Bypass (CNB4)
: 1. UNISOLABLE breach of containment AND 2. Direct downstream pathway to the environment exists after
 
containment isolation signal None 5. Containment Radiation Monitoring (CNB5)
None Containment high range radiation monitor reading
> 4000 R/hr
: 6. Other Indications (CNB6)
Elevated readings on the following radiation monitors that indicate loss or potential loss of the
 
Containment barrier:
MONITORS - Unit 1 MONITORS - UNIT 2 RX-9820 Containment Purge 2RX-9820 Containment Purge RX-9825 Radwaste Area 2RX-9825 Radwaste Area RX-9830 Fuel Handling Area 2RX-9830 Fuel Handling Area RX-9835 Emergency Penetration Room 2RX-9835 Emergency Penetration Room  2RX-9840 Post Accident Sampling Building 7. Emergency Director Judgment (CNB7)
Any condition in the opinion of the SM / TSC Director / EOF Director that indicates Loss or Potential Loss of the Containment Barrier    to 0CAN121102
 
Page 60 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document ANO: CONTAINMENT BARRIER (Cont'd)
Deviations:
None.
Differences:
EALs for Unit 1 and Unit 2 that correspond to the Critical Safety Function Status Trees (CSFSTs) included in the NEI 99-01, Revision 5, EALs are not provided. Unit 1 (a B&W plant)
 
does not have a Safety Function process similar to that of Westinghouse units (which are
 
included in the NEI 99-01, Revision 5, example EALs).
 
Unit 2 (a CE plant) does have methodologies similar to CSFSTs with some differences. Unit 2
 
performs Standard Post Trip Actions (SPTA) and verifies the status of its Safety Functions (Reactivity, Vital Auxiliaries, RCS Inventory, RCS Pressure, Core Heat Removal, RCS Heat Removal, & Containment) upon any reactor trip. An evaluation is made of each of the seven
 
Safety Functions comparing plant response and critical parameters to standard, expected
 
values. Each Safety Function is then marked on a tracking sheet as "SATISIFIED" or "NOT
 
SATISIFIED."  If any Safety Function is not satisfied, the condition is announced to the Control
 
Room staff and a diagnostic flow chart is referenced to determine the proper Optimal Recovery
 
Procedure (ORP) to enter. The diagnostic flowchart may direct the Operator to an ORP or to the
 
Functional Recovery Procedure (FRP). The intent of this diagnostic action is to direct the
 
Operator to the ORPs for a single event, and to the FRP for multiple events. Each ORP
 
contains Safety Function Status Checks (SFSC) which are performed every 15 minutes. These
 
checks ensure the Operator's utilization of the ORP is properly addressing plant critical
 
parameters. If the SFSC is not met then the FRP is entered. Criteria for FRP entry is:
: 1. ANY event in progress which can NOT be diagnosed as a single event. 2. Actions taken have NOT satisfied SFSC acceptance criteria.
: 3. Entry is directed by Diagnostic Actions.
 
While all ORPs have SFSC criteria for all Safety Functions, all ORP acceptance criteria are not
 
the same. For example; the SFSC criteria required to SATISIFY the requirements for RCS Heat
 
Removal (a Safety Function) is different in the SPTAs than in Loss of Coolant Accident (LOCA)
 
ORP, which is different than those contained in the Steam Generator Tube Rupture (SGTR)
 
ORP, which is different than those in the Loss of Feedwater (LOF) ORP, and so on.
 
To compare NEI 99-01 Safety Function intent with that of Unit 2 above, a Loss of Offsite Power (LOOP) event is correctly classified per NEI 99-01 as System Malfunction - Loss of AC Power (SU1), "Loss of all offsite AC power to Vital 4.16 KV busses  15 minutes."  This event is correctly classified as a NUE. However, a LOOP results in the loss of the operating Reactor
 
Coolant Pumps due to loss of non-vital power sources. During performance of the SPTAs, the
 
Core Heat Removal Safety Function will be assessed as "NOT SATISIFIED."  If NEI 99-01, Fission Product Barrier Malfunction - Barriers, is referenced, FCB1 would be applicable and the
 
event would require declaration of an ALERT.
 
to 0CAN121102
 
Page 61 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document ANO: CONTAINMENT BARRIER (Cont'd)
Differences (cont'd):
Entergy believes erroneous classification of a standard NUE condition as a ALERT is
 
inappropriate. Natural circulation of the reactor coolant provides adequate core cooling and no
 
significant challenge to the core heat removal functions are expected to occur. Other EALs are
 
referenced when plant conditions change or degrade. The remaining EALs used in the
 
Containment Barrier (CNB) section provide the Operators with the necessary information to classify events appropriately based on the actual radiation levels, reactor vessel level, containment pressure, etc., without the need for a reference to Safety Functions.
 
In summary, the ORP SFSC criteria were not established to meet EAL classification
 
requirements. The SFCSs ensure Safety Function status is verified and updated at regular
 
intervals so that changes in plant conditions may be recognized promptly and to enable trending
 
of plant parameters important to safety. Based on the above, Entergy proposes to not adopt the
 
use of Safety Functions in the Containment Barrier EAL. Subsequently, the NEI 99-01 CNBs
 
are renumbered in the ANO EALs for formatting purposes based on the non-use of the Safety
 
Function Status CNB criteria.
 
The wording "rapid unexplained drop in containment pressure following an initial rise in
 
containment pressure" is utilized vice the NEI (inverse) wording of "a containment pressure rise followed by a rapid unexplained drop in containment pressure."  This revised wording is
 
provided for clarity and has the same meaning as the NEI wording.
 
Additional basis information is provided to descr ibe the existence of an explosive mixture inside containment.
 
Additional site-specific basis information is provided to describe the NUE condition related to
 
RCS leakage for SG secondary side release with primary-to-secondary leakage.
 
Additional basis information is provided for containment isolation failure or bypass that
 
addresses the valve failure EAL (threshold) criteria in NEI 99-05 and clarifies classification as it
 
relates to attempted isolation from the Control Room.
 
Broader terminology for ANO CNB4 other than limiting the condition to isolation valves is used. 
 
This meets the NEI 99-01 intent because the NEI condition is bounded by the EAL chosen for
 
ANO and the basis information is unchanged.
 
Basis information is added for containment radiation monitoring that states that the value
 
provided represents in itself indication for a General Emergency classification because the
 
value also exceeds the loss EALs for fuel clad and RCS.
 
Other indications of the loss of the containment barrier based on radiation monitor readings are
 
provided.
 
The Containment Barrier EALs are not provided in a table in the Basis Document. The EALs
 
are presented as text. A table is used in the EAL Matrix document. to 0CAN121102
 
Page 62 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document NEI 99-01:  HU4 Confirmed SECURITY CONDITION or threat which indicates a potential degradation in the level
 
of safety of the plant
 
Operating Mode Applicability:
All  Example Emergency Action Levels:
(1 or 2 or 3)
: 1. A SECURITY CONDITION that does NOT involve a HOSTILE ACTION as reported by the (site-specific security shift supervision).
: 2. A credible site-specific security threat notification.
: 3. A validated notification from NRC providing information of an aircraft threat.
 
ANO:  HU1  Confirmed SECURITY CONDITION or threat which indicates a potential degradation in the level
 
of safety of the plant
 
Operating Mode Applicability:
All  Emergency Action Level(s): (1 or 2 or 3)
: 1. A SECURITY CONDITION that does not involve a HOSTILE ACTION as reported by ANO Security Shift Supervision.
OR  2. A credible site-specific security threat notification.
 
OR  3. A validated notification from NRC providing information of an aircraft threat.
 
Deviations:
None.
Differences:
NEI 99-01 HU4 is renumbered to ANO HU1 for formatting purposes based on site preference for order of ICs alone. to 0CAN121102
 
Page 63 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document NEI 99-01:  HU5 Other conditions exist which in the judgment of the Emergency Director warrant declaration of a
 
NOUE.
Operating Mode Applicability:
All  Example Emergency Action Level:
: 1. Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or have occurred which indicate a potential degradation of the level of
 
safety of the plant or indicate a security threat to facility protection has been initiated. No
 
releases of radioactive material requiring off-site response or monitoring are expected
 
unless further degradation of safety systems occurs.
 
ANO:  HU2 Other conditions exist which in the judgment of the SM warrant declaration of an NUE
 
Operating Mode Applicability:
All  Emergency Action Level(s):
: 1. Other conditions exist which in the judgment of the SM indicate that events are in progress or have occurred which indicate a potential degradation of the level of safety of the plant or
 
indicate a security threat to facility protection has been initiated. No releases of radioactive
 
material requiring offsite response or monitoring are expected unless further degradation of safety systems occurs.
Deviations:
None.
Differences:
NEI 99-01 HU5 is renumbered to ANO HU2 for formatting purposes based on site preference for order of ICs alone. to 0CAN121102
 
Page 64 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document NEI 99-01:  HU2 FIRE within the PROTECTED AREA not extinguished within 15 minutes of detection or EXPLOSION within the PROTECTED AREA
 
Operating Mode Applicability:
All  Example Emergency Action Level:
(1 or 2)  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)
: 2. EXPLOSION within the PROTECTED AREA.
 
to 0CAN121102
 
Page 65 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document ANO:  HU4 FIRE within the PROTECTED AREA not extinguished within 15 minutes of detection or EXPLOSION within the PROTECTED AREA Operating Mode Applicability:
All  Emergency Action Level(s):
(1 or 2)  Note: The SM 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 in any Table H1 structure or area not extinguished 1) within 15 minutes of Control Room notification or 2) within 15 minutes of verification of a Control Room FIRE alarm.
Table H1 Unit 1 Unit 2 CA-1 & HP Office Area Condensate Demineralizer Room
 
Corridor 98
 
Fire Area C
 
Lower North Electrical Penetration Room (LNEPR) Lower South Electrical Equipment Room (LSEER)
/ Air Compressor Room Lower South Electrical Penetration Room (LSEPR) Lower South Piping Penetration Room (LSPPR)
 
Main Steam Isolation Violation (MSIV) Room
 
North Engineered Safeguards (ES) SWGR Room (A4) South ES SWGR Room
 
Turbine Building  A1, A2, H1, H2 SWGR area  354' Bowling Alley north end west of Breathing Air compressor room  368' West Heater Deck from LSEER (orange door) along east wall of ES SWGR Rooms to
 
Corridor 98 door Upper North Electrical Penetration Room (UNEPR) / Hot Tool Room / Decon Room Upper South Electrical Penetration Room (USEPR) Upper South Piping Penetration Room (USPPR) 2A3 Room 2A4, 2D02, & East Battery Room
 
2B53 Room
 
2B63 Room
 
2B9/2B10 Room
 
2Y11/13 Equipment Room
 
Auxiliary Building 317' General Access
 
Auxiliary Building 335'
 
Auxiliary Building 354'
 
'B' Engineered Safeguards Features (ESF)
Room Corridor Behind Door 340
 
Turbine Building  2A1, 2A2, 2H1, 2H2 Area  354' West wall of Demineralizer area  368' West Heater Deck north of north Switchgear (SWGR) Room (2A3) and
 
East of LNEPR Intake Structure  354' or 366'
 
LNEPR LSEPR Motor-Generator (MG) Set Room
 
Steam Pipe Area
 
Hot Machine Shop
 
UNEPR, UNPPR, LNPPR, USPPR    to 0CAN121102
 
Page 66 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document ANO:  HU4 (Cont'd)
OR  2. EXPLOSION within the PROTECTED AREA.
 
Deviations:
None.
Differences:
NEI 99-01 HU2 is renumbered to ANO HU4 for formatting purposes based on site preference for order of ICs alone.
 
Site-specific areas for EAL #1 are presented in a table and the table is referenced causing a
 
minor difference in EAL language from that in NEI 99-01. to 0CAN121102
 
Page 67 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document NEI 99-01:  HU3 Release of toxic, corrosive, asphyxiant, or flammable gases deemed detrimental to NORMAL PLANT OPERATIONS
 
Operating Mode Applicability:
All  Example Emergency Action Levels:
(1 or 2) 
: 1. Toxic, corrosive, asphyxiant or flammable gas es in amounts that have or could adversely affect NORMAL PLANT OPERATIONS.
: 2. Report by local, county or state officials for evacuation or sheltering of site personnel based on an off-site event.
 
ANO:  HU5 Release of toxic, corrosive, asphyxiant, or flammable gases deemed detrimental to NORMAL PLANT OPERATIONS
 
Operating Mode Applicability:
All  Emergency Action Level(s):
(1 or 2) 
: 1. Toxic, corrosive, asphyxiant or flammable gas es 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 offsite event.
Deviations:
None.
Differences:
NEI 99-01 HU3 is renumbered to ANO HU5 for formatting purposes based on site preference for order of ICs alone. In an attempt to group "families" of emergency classes together with the
 
same last digit Arabic numeral designation (such as HU1, AU1, AS1, etc.), an IC-labeled HU3 is
 
not provided. All NEI 99-01 NOUE ICs for the Hazards and Other Conditions Affecting Plant
 
Safety category are still addressed in the ANO EALs. to 0CAN121102
 
Page 68 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document NEI 99-01:  HU1 Natural or destructive phenomena affecting the PROTECTED AREA
 
Operating Mode Applicability:
All  Example Emergency Action Levels:
(1 or 2 or 3 or 4 or 5)
: 1. Seismic event identified by ANY 2 of the following:  Seismic event confirmed by (site-specific indication or method)  Earthquake felt in plant  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).
 
ANO:  HU6 Natural or destructive phenomena affecting the PROTECTED AREA
 
Operating Mode Applicability:
All  Emergency Action Level(s):  (1 or 2 or 3 or 4 or 5 or 6)
: 1. Seismic event identified by any 2 of the following:  Seismic event confirmed by annunciation of the 0.01g acceleration alarm  Earthquake felt in plant  National Earthquake Center OR  2. Tornado striking within PROTECTED AREA boundary or high winds > 67 mph.
 
OR    to 0CAN121102
 
Page 69 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document ANO:  HU6 (Cont'd)
: 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 structures or areas in Table H1. Table H1 Unit 1 Unit 2 CA-1 & HP Office Area Condensate Demineralizer Room
 
Corridor 98
 
Fire Area C
 
Lower North Electrical Penetration Room (LNEPR)
 
Lower South Electrical Equipment Room (LSEER) /
Air Compressor Room Lower South Electrical Penetration Room (LSEPR)
 
Lower South Piping Penetration Room (LSPPR)
 
Main Steam Isolation Violation (MSIV) Room
 
North Engineered Safeguards (ES) SWGR Room (A4) South ES SWGR Room
 
Turbine Building  A1, A2, H1, H2 SWGR area  354' Bowling Alley north end west of Breathing Air compressor room  368' West Heater Deck from LSEER (orange door) along east wall of ES SWGR Rooms to
 
Corridor 98 door.
Upper North Electrical Penetration Room (UNEPR) /
Hot Tool Room / Decon Room Upper South Electrical Penetration Room (USEPR)
 
Upper South Piping Penetration Room (USPPR) 2A3 Room 2A4, 2D02, & East Battery Room
 
2B53 Room
 
2B63 Room
 
2B9/2B10 Room
 
2Y11/13 Equipment Room
 
Auxiliary Building 317' General Access
 
Auxiliary Building 335'
 
Auxiliary Building 354'
 
'B' Engineered Safeguards Features (ESF) Room Corridor Behind Door 340
 
Turbine Building  2A1, 2A2, 2H1, 2H2 Area  354' West wall of Demineralizer area 368' West Heater Deck north of north Switchgear (SWGR) Room
 
(2A3) and East of LNEPR Intake Structure  354' or 366'
 
LNEPR LSEPR Motor-Generator (MG) Set Room
 
Steam Pipe Area
 
Hot Machine Shop
 
UNEPR, UNPPR, LNPPR, USPPR OR  4. Turbine failure resulting in casing penetration or damage to turbine or generator seals.
OR  5. Lake Dardanelle level < 335 feet.
OR  6. Lake Dardanelle level > 345 feet. to 0CAN121102
 
Page 70 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document ANO:  HU6 (Cont'd)
Deviations:
None.
Differences:
NEI 99-01 HU1 is renumbered to ANO HU6 for formatting purposes based on site preference for order of ICs alone.
 
Site-specific information is added on Safety Analys is Report (SAR) design basis in the bases for EAL #2.
 
Site-specific areas for EAL #3 are presented in a table and the table is referenced causing a
 
minor difference in EAL language from that in NEI 99-01.
 
The reference to VISIBLE DAMAGE is removed from the bases for EAL #3 for escalation
 
information because NEI 99-01 does not refer to VISIBLE DAMAGE for the corresponding Alert
 
classification in HA1 EAL #3. 
 
An evaluation was performed to determine if any other site-specific occurrences were applicable
 
to ANO. Lake Dardanelle levels were determined to be the only other site-specific occurrences
 
that warrant declaration of an NUE. These EALs were added as EALs #5 and #6. to 0CAN121102
 
Page 71 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document NEI 99-01:  HA4 HOSTILE ACTION within the OWNER CONTROLLED AREA or airborne attack threat
 
Operating Mode Applicability:
All  Example Emergency Action Levels: (1 or 2)
: 1. A HOSTILE ACTION is occurring or has occurred within the OWNER CONTROLED AREA as reported by the (site-specific security shift supervision).
: 2. A validated notification from NRC of an airliner attack threat within 30 minutes of the site.
 
ANO:  HA1 HOSTILE ACTION within the OWNER CONTROLLED AREA or airborne attack threat
 
Operating Mode Applicability:
All  Emergency Action Level(s): (1 or 2)
: 1. A HOSTILE ACTION is occurring or has occurred within the OWNER CONTROLLED AREA as reported by ANO Security Shift Supervision.
OR  2. A validated notification from NRC of an airliner attack threat within 30 minutes of the site.
 
Deviations:
None.
Differences:
NEI 99-01 HA4 is renumbered to ANO HA1 for formatting purposes based on site preference for order of ICs alone. to 0CAN121102
 
Page 72 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document NEI 99-01:  HA6 Other conditions exist which in the judgment of the Emergency Director warrant declaration of
 
an Alert.
 
Operating Mode Applicability:
All  Example Emergency Action Level:
: 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
 
ACTION. Any releases are expected to be limited to small fractions of the Environmental
 
Protection Agency (EPA) Protective Action Guideline exposure levels.
 
ANO:  HA2 Other conditions exist which in the judgment of the SM / TSC Director / EOF Director warrant
 
declaration of an Alert.
 
Operating Mode Applicability:
All  Emergency Action Level(s):
: 1. Other conditions exist which in the judgment of the SM / TSC Director / EOF Director indicate that events are in progress or have occurred which involve an actual or potential substantial
 
degradation of the level of safety of the plant or a security event that involves probable life
 
threatening risk to site personnel or damage to site equipment because of HOSTILE
 
ACTION. Any releases are expected to be limited to small fractions of the EPA Protective
 
Action Guideline exposure levels.
Deviations:
None.
Differences:
NEI 99-01 HA6 is renumbered to ANO HA2 for formatting purposes based on site preference for order of ICs alone. to 0CAN121102
 
Page 73 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document NEI 99-01:  HA5 Control room evacuation has been initiated.
 
Operating Mode Applicability:
All  Example Emergency Action Level:
: 1. (Site-specific procedure) requires control room evacuation.
 
ANO:  HA3 Control Room evacuation has been initiated.
 
Operating Mode Applicability:
All  Emergency Action Level(s):
: 1. Alternate Shutdown procedure requires Control Room evacuation:
 
Unit 1:  1203.002, "Alternate Shutdown" Unit 2:  2203.014, "Alternate Shutdown" Deviations:
None.
Differences:
NEI 99-01 HA5 is renumbered to ANO HA3 for formatting purposes based on site preference for order of ICs alone. to 0CAN121102
 
Page 74 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document NEI 99-01:  HA2 FIRE or EXPLOSION affecting the operability of plant safety systems required to establish or
 
maintain safe shutdown
 
Operating Mode Applicability:
All  Example Emergency Action Level:
: 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)    to 0CAN121102
 
Page 75 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document ANO:  HA4 FIRE or EXPLOSION affecting the operability of plant safety systems required to establish or
 
maintain safe shutdown
 
Operating Mode Applicability:
All  Emergency Action Level(s):
: 1. FIRE or EXPLOSION resulting in VISIBLE DAMAGE to any Table H1 structure or area containing safety systems or components or Control Room indication of degraded performance of those safety systems.
Table H1 Unit 1 Unit 2 CA-1 & HP Office Area Condensate Demineralizer Room
 
Corridor 98
 
Fire Area C
 
Lower North Electrical Penetration Room (LNEPR) Lower South Electrical Equipment Room (LSEER)
/ Air Compressor Room Lower South Electrical Penetration Room (LSEPR) Lower South Piping Penetration Room (LSPPR)
 
Main Steam Isolation Violation (MSIV) Room
 
North Engineered Safeguards (ES) SWGR Room (A4) South ES SWGR Room
 
Turbine Building  A1, A2, H1, H2 SWGR area  354' Bowling Alley north end west of Breathing Air compressor room  368' West Heater Deck from LSEER (orange door) along east wall of ES SWGR Rooms to
 
Corridor 98 door.
Upper North Electrical Penetration Room (UNEPR) / Hot Tool Room / Decon Room Upper South Electrical Penetration Room (USEPR) Upper South Piping Penetration Room (USPPR) 2A3 Room 2A4, 2D02, & East Battery Room
 
2B53 Room
 
2B63 Room
 
2B9/2B10 Room
 
2Y11/13 Equipment Room
 
Auxiliary Building 317' General Access
 
Auxiliary Building 335'
 
Auxiliary Building 354'
 
'B' Engineered Safeguards Features (ESF)
Room Corridor Behind Door 340
 
Turbine Building  2A1, 2A2, 2H1, 2H2 Area  354' West wall of Demineralizer area  368' West Heater Deck north of north Switchgear (SWGR) Room (2A3) and
 
East of LNEPR Intake Structure  354' or 366'
 
LNEPR LSEPR Motor-Generator (MG) Set Room
 
Steam Pipe Area
 
Hot Machine Shop
 
UNEPR, UNPPR, LNPPR, USPPR to 0CAN121102
 
Page 76 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document ANO:  HA4 (Cont'd)
Deviations:
None.
Differences:
NEI 99-01 HA2 is renumbered to ANO HA4 for formatting purposes based on site preference for order of ICs alone. to 0CAN121102
 
Page 77 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document NEI 99-01:  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 Mode Applicability:
All  Example Emergency Action Levels:
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.
 
ANO:  HA5 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 Mode Applicability:
All  Emergency Action Level(s):
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.
Deviations:
None.
Differences:
NEI 99-01 HA3 is renumbered to ANO HA5 for formatting purposes based on site preference for order of ICs alone. to 0CAN121102
 
Page 78 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document NEI 99-01:  HA1 Natural or Destructive Phenomena Affecting VITAL AREAs
 
Operating Mode Applicability:
All  Example Emergency Action Levels:  (1 or 2 or 3 or 4 or 5 or 6)
: 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 perform ance of systems required for the safe shutdown of the plant
: 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 perfo rmance of those safety systems:
(site-specific structure list)
: 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)
: 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 performanc e of those safety systems:
(site-specific structure list)
: 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)
: 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:    to 0CAN121102
 
Page 79 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document ANO:  HA6 Natural or destructive phenomena affecting VITAL AREAS
 
Operating Mode Applicability:
All  Emergency Action Level(s): (1 or 2 or 3 or 4 or 5 or 6)
: 1. a. Seismic event > Operating Basis Earthquake (OBE) as indicated by annunciation of the 0.1g acceleration alarm.
AND  b. Earthquake confirmed by any of the following:
Earthquake felt in plant  National Earthquake Center  Control Room indication of degraded perform ance of systems required for the safe shutdown of the plant OR  2. Tornado striking or high winds > 67 mph resulting in VISIBLE DAMAGE to any of the following structures/equipment contai ning safety systems or components or Control Room indication of degraded performance of those safety systems:
Reactor Building Turbine Building Intake Structure Q Condensate Storage Tank (QCST)
Ultimate Heat Sink Control Room Startup Transformers Auxiliary Building Diesel Fuel Vault Borated Water Storage Tank (BWST)
 
Refueling Water Tank (RWT)
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:
Intake Structure Turbine Building Ultimate Heat Sink Control Room BWST / RWT Startup Transformers Auxiliary Building Diesel Fuel Vault
 
QCST    to 0CAN121102
 
Page 80 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document ANO:  HA6 (Cont'd)
Emergency Action Level(s) (Cont'd)
OR  4. Turbine failure-generated PROJECTILES resulting in VISIBLE DAMAGE to or penetration of any of the structures/equipment in Table H2 containing safety systems or components or Control Room indication of degraded perfo rmance of those safety systems.
Table H2 Reactor Building Turbine Building Intake Structure QCST Ultimate Heat Sink Control Room BWST/RWT Startup Transformers Auxiliary Building Diesel Fuel Vault OR  5. Lake Dardanelle level < 335 feet and Emergency Cooling Pond inoperable.
 
OR  6. Vehicle crash resulting in VISIBLE DAMAGE to any of the structures/equipment in Table H2 containing safety systems or components or Control Room indication of degraded performance of those safety systems.
Table H2 Reactor Building Turbine Building Intake Structure QCST Ultimate Heat Sink Control Room BWST/RWT Startup Transformers Auxiliary Building Diesel Fuel Vault Deviations:
None.
to 0CAN121102
 
Page 81 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document ANO:  HA6 (Cont'd)
Differences:
NEI 99-01 HA1 is renumbered to ANO HA6 for formatting purposes based on site preference for order of ICs alone.
 
Site-specific information is added on SAR design basis in the bases for EAL #2.
 
Site-specific areas for EALs #4 and #6 are presented in a table and the table is referenced
 
causing a minor difference in EAL language from that in NEI 99-01.
 
An evaluation was performed to determine if any other site-specific occurrences were applicable
 
to ANO. Lake Dardanelle low level with the Emergency Cooling Pond inoperable was
 
determined to be the only other site-specific occurrence that warrants declaration of an Alert.
EAL #5 was added. The NEI EAL #5 was renumbered as EAL #6 to provide consistency
 
between NUE and Alert EALs for lake level. to 0CAN121102
 
Page 82 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document NEI 99-01:  HS4 HOSTILE ACTION within the PROTECTED AREA
 
Operating Mode Applicability:
All  Example Emergency Action Level:
: 1. A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by the (site security shift supervision).
 
ANO:  HS1 HOSTILE ACTION within the PROTECTED AREA
 
Operating Mode Applicability:
All  Emergency Action Level(s):
: 1. A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by ANO Security Shift Supervision.
Deviations:
None.
Differences:
NEI 99-01 HS4 is renumbered to ANO HS1 for formatting purposes based on site preference for order of ICs alone. to 0CAN121102
 
Page 83 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document NEI 99-01:  HS3 Other conditions exist which in the judgment of the Emergency Director warrant declaration of a
 
Site Area Emergency.
 
Operating Mode Applicability:
All  Example Emergency Action Level:
: 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.
 
ANO:  HS2 Other conditions exist which in the judgment of the SM / TSC Director / EOF Director warrant
 
declaration of a Site Area Emergency
 
Operating Mode Applicability:
All  Emergency Action Level(s):
: 1. Other conditions exist which in the judgment of the SM / TSC Director / EOF 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.
Deviations:
None.
Differences:
NEI 99-01 HS3 is renumbered to ANO HS2 for formatting purposes based on site preference for order of ICs alone. to 0CAN121102
 
Page 84 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document NEI 99-01:  HS2 Control room evacuation has been initiated and plant control cannot be established.
 
Operating Mode Applicability:
All  Example Emergency Action Level:
: 1. a. Control room evacuation has been initiated.
 
AND  b. Control of the plant cannot be established within (site-specific minutes).
 
ANO:  HS3 Control Room evacuation has been initiated and plant control cannot be established
 
Operating Mode Applicability:
All  Emergency Action Level(s):
: 1. a. Control Room evacuation has been initiated
 
AND  b. Control of the plant cannot be established in accordance with the following procedures within 15 minutes:
Unit 1:  1203.002, "Alternate Shutdown" Unit 2:  2203.014, "Alternate Shutdown" Deviations:
None.
Differences:
NEI 99-01 HS2 is renumbered to ANO HS3 for formatting purposes based on site preference for order of ICs alone.
 
The procedural reference is included in the EAL for consistency with HA3. to 0CAN121102
 
Page 85 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document NEI 99-01:  HG1 HOSTILE ACTION resulting in loss of physical control of the facility.
 
Operating Mode Applicability:
All  Example Emergency Action Level:
(1 or 2) 
: 1. A HOSTILE ACTION has occurred such that plant personnel are unable to operate equipment required to maintain safety functions.
: 2. A HOSTILE ACTION has caused failure of Spent Fuel Cooling Systems and IMMINENT fuel damage is likely for a freshly off-loaded reactor core in pool.
 
ANO:  HG1 HOSTILE ACTION resulting in loss of physical control of the facility
 
Operating Mode Applicability:
All  Emergency Action Level(s): (1 or 2)
: 1. A HOSTILE ACTION has occurred such that plant personnel are unable to operate equipment required to maintain safety functions.
OR  2. A HOSTILE ACTION has caused failure of Spent Fuel Cooling Systems and IMMINENT fuel damage is likely for a freshly off-loaded reactor core in pool.
Deviations:
None.
Differences:
None. to 0CAN121102
 
Page 86 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document NEI 99-01:  HG2 Other conditions exist which in the judgment of the Emergency Director warrant declaration of a
 
General Emergency.
 
Operating Mode Applicability:
All  Example Emergency Action Level:
: 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 IMMINENT substantial core
 
degradation or melting with potential for loss of containment integrity or HOSTILE ACTION
 
that results in an actual loss of physical control of the facility. Releases can be reasonably
 
expected to exceed EPA Protective Action Guideline exposure levels offsite for more than
 
the immediate site area.
 
ANO:  HG2 Other conditions exist which in the judgment of the SM / TSC Director / EOF Director warrant
 
declaration of a General Emergency.
 
Operating Mode Applicability:
All  Emergency Action Level(s):
: 1. Other conditions exist which in the judgment of the SM / TSC Director / EOF Director indicate that events are in progress or have occurred which involve actual or IMMINENT
 
substantial core degradation or melting with potential for loss of containment integrity or
 
HOSTILE ACTION that results in an actual loss of physical control of the facility. Releases
 
can be reasonably expected to exceed EPA Protec tive Action Guideline exposure levels offsite for more than the immediate site area.
Deviations:
None.
Differences:
None. to 0CAN121102
 
Page 87 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document NEI 99-01:  SU1 Loss of all Off-site AC power to emergency busses for 15 minutes or longer.
 
Operating Mode Applicability:
Power Operation Startup Hot Standby
 
Hot Shutdown Example Emergency Action Level:
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.
 
ANO:  SU1 Loss of all offsite AC power to Vital 4.16 KV busses  15 minutes Operating Mode Applicability:
Power Operations (Mode 1)
Startup (Mode 2)
 
Hot Standby (Mode 3)
 
Hot Shutdown (Mode 4)
Emergency Action Level(s):
Note: The SM 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 offsite AC power to Vital 4.16 KV busses  15 minutes.
Deviations:
None.
Differences:
The site-specific term "Vital 4.16 KV" is used in the IC and EAL to define emergency busses. to 0CAN121102
 
Page 88 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document NEI 99-01:  SU3 UNPLANNED loss of safety system annunciation or indication in the control room for 15 minutes
 
or longer.
 
Operating Mode Applicability:
Power Operation Startup Hot Standby
 
Hot Shutdown Example Emergency Action Level:
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  b. (Site-specific control room safety system indication) to 0CAN121102
 
Page 89 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document ANO:  SU6 UNPLANNED loss of safety system annunciati on or indication in the Control Room  15 minutes Operating Mode Applicability: Power Operations (Mode 1)
Startup (Mode 2)
 
Hot Standby (Mode 3)
 
Hot Shutdown (Mode 4)
Emergency Action Level(s):
Note: The SM 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 > approximately 75% of the following  15 minutes:
: a. Control Room annunciators associated with safety systems.
OR  b. Control Room safety system indication.
Deviations:
None.
Differences:
NEI 99-01 SU3 is renumbered to ANO SU6 for formatting purposes based on site preference for order of ICs alone. In an attempt to group "families" of emergency classes together with the
 
same last digit Arabic numeral designation (such as SU1, SA1, SS1, etc.), ICs labeled SU2, SU3, SU4, and SU5 are not provided. All NE I 99-01 NOUE ICs for the System Malfunction category are still addressed in the ANO EALs.
 
Additional information is provided in the bases to define those systems associated with safety system indication. to 0CAN121102
 
Page 90 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document NEI 99-01:  SU5 RCS leakage
 
Operating Mode Applicability:
Power Operation Startup Hot Standby
 
Hot Shutdown Example Emergency Action Levels: (1 or 2)
: 1. Unidentified or pressure boundary leakage greater than 10 gpm.
: 2. Identified leakage greater than 25 gpm.
 
ANO:  SU7 RCS leakage
 
Operating Mode Applicability: Power Operations (Mode 1)
Startup (Mode 2)
 
Hot Standby (Mode 3)
 
Hot Shutdown (Mode 4)
Emergency Action Level(s):
(1 or 2) 
: 1. Unidentified or pressure boundary leakage > 10 gpm.
 
OR  2. Identified leakage
> 25 gpm.
Deviations:
None.
Differences:
Because prompt, procedurally driven action can be taken to isolate some sources of potential
 
RCS leakage (such as isolating the Letdown system), the ANO basis document further defines
 
this leakage to be a loss of RCS inventory due to a leak in the RCS or a supporting system that
 
is not or cannot be isolated within 10 minutes. This clarification meets the intent of NEI 99-01.
 
NEI 99-01 SU5 is renumbered to ANO SU7 for formatting purposes based on site preference for order of ICs alone.
 
Basis information is added noting that steam generator tube leakage is considered as identified
 
leakage. to 0CAN121102
 
Page 91 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document NEI 99-07:  SU6 Loss of All On-site or Off-site communications capabilities
 
Operating Mode Applicability:
Power Operation Startup Hot Standby
 
Hot Shutdown Example Emergency Action Levels: (1 or 2)
: 1. Loss of all of the following on-site communication methods affecting the ability to perform routine operations.
(site-specific list of communications methods)
 
2 Loss of all of the following off-site communication methods affecting the ability to perform offsite notifications.
(site-specific list of communications methods)
 
ANO:  SU8 Loss of all onsite or offsite communications capabilities
 
Operating Mode Applicability: Power Operations (Mode 1)
Startup (Mode 2)
 
Hot Standby (Mode 3)
 
Hot Shutdown (Mode 4)
Emergency Action Level(s):
(1 or 2)  1. Loss of all Table M1 onsite communications methods affecting the ability to perform routine operations.
OR    to 0CAN121102
 
Page 92 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document ANO:  SU8 (Cont'd)
: 2. Loss of all Table M2 offsite communications methods affecting the ability to perform offsite notifications.
Table M1 Onsite Communications Methods  Table M2 Offsite Communications Methods Station radio system  All telephone lines (commercial and microwave) Plant paging system  ENS In-plant telephones  Gaitronics Deviations:
None.
Differences:
NEI 99-01 SU6 is renumbered to ANO SU8 for formatting purposes based on site preference for order of ICs alone.
 
Onsite and offsite communications methods are presented in tables and the tables are
 
referenced causing a minor difference in EAL language from that in NEI 99-01. to 0CAN121102
 
Page 93 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document NEI 99-01: SU4 Fuel clad degradation
 
Operating Mode Applicability:
Power Operation Startup Hot Standby
 
Hot Shutdown Example Emergency Action Levels:
(1 or 2) 
: 1. (Site-specific radiation monitor readings indicating fuel clad degradation greater than Technical Specification allowable limits.)
: 2. (Site-specific coolant sample activity value indicating fuel clad degradation greater than Technical Specification allowable limits.)
 
ANO:  SU9 Fuel clad degradation
 
Operating Mode Applicability:
Power Operations (Mode 1)
Startup (Mode 2)
 
Hot Standby (Mode 3)
 
Hot Shutdown (Mode 4)
Emergency Action Level(s):
: 1. Failed Fuel Iodine radiation monitor reading indicates fuel clad degradation > Technical Specification allowable limits:
Unit 1: RI-1237S reads > 1.3 x 10 5 counts per minute Unit 2: 2RITS-4806B reads > .65 x 10 5 counts per minute OR    to 0CAN121102
 
Page 94 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document ANO:  SU9 (Cont'd)
: 2. RCS sample activity value indicating fuel clad degradation > Technical Specification allowable limits:
  > 1.0 uCi/gm Dose Equivalent I-131 (DEI) for more than 48 hours OR  Unit 1:  60 uCi/gm Dose Equivalent I-131 Unit 2: > 60 uCi/gm Dose Equivalent I-131 OR  Unit 1: > 2200 &#xb5;Ci/gm Dose Equivalent Xe-133 for more than 48 hours Unit 2: > 3100 &#xb5;Ci/gm Dose Equivalent Xe-133 for more than 48 hours Deviations:
None.
Differences:
NEI 99-01 SU4 is renumbered to ANO SU9 for formatting purposes based on site preference for order of ICs alone.
 
An EAL value and associated basis information are provided for the nominal operating limit for
 
DEI for RCS activity in addition to the transient iodine spiking limit described in NEI 99-01. This value is also provided for Dose Equivalent Xe-133 in accordance with the station's Technical
 
Specifications. to 0CAN121102
 
Page 95 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document NEI 99-01:  SU8 Inadvertent criticality
 
Operating Mode Applicability:
Hot Standby Hot Shutdown Example Emergency Action Level:
(1 or 2) 
: 1. UNPLANNED sustained positive period observed on nuclear instrumentation.
[BWR]
: 1. UNPLANNED sustained positive startup rate observed on nuclear instrumentation.
[PWR]
 
ANO:  SU10 Inadvertent criticality
 
Operating Mode Applicability: Hot Standby (Mode 3)
Hot Shutdown (Mode 4)
 
Emergency Action Level(s):
: 1. UNPLANNED sustained positive startup rate observed on nuclear instrumentation.
 
Deviations:
None.
Differences:
NEI 99-01 SU8 is renumbered to ANO SU10 for formatting purposes based on site preference for order of ICs alone. to 0CAN121102
 
Page 96 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document NEI 99-01:  SU2 Inability to reach required shutdown within Technical Specification limits
 
Operating Mode Applicability:
Power Operation Startup Hot Standby
 
Hot Shutdown Example Emergency Action Level:
: 1. Plant is not brought to required operating mode within Technical Specifications Limiting Condition for Operation (LCO) Action Statement Time.
 
ANO:  SU11 Inability to reach required operating mode within Technical Specification limits
 
Operating Mode Applicability:
Power Operations (Mode 1)
Startup (Mode 2)
 
Hot Standby (Mode 3)
 
Hot Shutdown (Mode 4)
Emergency Action Level(s):
: 1. Plant is not brought to required operating mode within Technical Specifications LCO action statement time.
Deviations:
None.
Differences:
NEI 99-01 SU2 is renumbered to ANO SU11 for formatting purposes based on site preference for order of ICs alone.
 
The NEI IC is changed to refer to required "operating mode" vice "shutdown" to provide clarity
 
and agreement with the associated NEI EAL and bases. This change does not alter the
 
meaning or intent of the IC. to 0CAN121102
 
Page 97 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document NEI 99-01:  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 Mode Applicability:
Power Operation Startup Hot Standby
 
Hot Shutdown Example Emergency Action Level:
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.
 
to 0CAN121102
 
Page 98 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document ANO:  SA1 AC power capability to Vital 4.16 KV busses reduced to a single power source  15 minutes such that any additional single failure would result in station blackout
 
Operating Mode Applicability:
Power Operations (Mode 1)
Startup (Mode 2)
 
Hot Standby (Mode 3)
 
Hot Shutdown (Mode 4)
Emergency Action Level(s):
Note: The SM / TSC Director / EOF 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 Vital 4.16 KV busses reduced to a single power source  15 minutes.
AND  b. Any additional single power source failure will result in station blackout.
Deviations:
None.
Differences:
NEI 99-01 SA5 is renumbered to ANO SA1 for formatting purposes based on site preference for order of ICs alone.
 
The site-specific term "Vital 4.16 KV" is used in the IC and EAL to define emergency busses. to 0CAN121102
 
Page 99 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document NEI 99-01:  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 Mode Applicability:
Power Operation Startup  Example Emergency Action Level:
: 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).
 
ANO:  SA3 Automatic trip fails to shutdown the reactor and the manual actions taken from the reactor
 
control console are successful in shutting down the reactor
 
Operating Mode Applicability:
Power Operations (Mode 1)
Startup (Mode 2)
Emergency Action Level(s):
: 1. a. An automatic trip failed to shutdown the reactor as indicated by reactor power  5%. AND  b. Manual actions taken at the reactor control console successfully shutdown the reactor as indicated by reactor power < 5%.
Deviations:
None.
Differences:
NEI 99-01 SA2 is renumbered to ANO SA3 for formatting purposes based on site preference for order of ICs alone.
 
The site-specific % reactor power is used to indicate both when the automatic reactor trip is not
 
successful and when the manual reactor trip is successful.
 
Additional information is provided in the bases to define a failure of the manual trip function. to 0CAN121102
 
Page 100 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document NEI 99-01:  SA4 UNPLANNED Loss of safety system annunciation or i ndication in the control room with either (1) a SIGNIFICANT TRANSIENT in progress, or (2) compensatory indicators unavailable.
 
Operating Mode Applicability:
Power Operation Startup Hot Standby
 
Hot Shutdown Example Emergency Action Level:
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. 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. to 0CAN121102
 
Page 101 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document ANO:  SA6 UNPLANNED loss of safety system annunciation or i ndication in the Control Room with either (1) a SIGNIFICANT TRANSIENT in progress, or (2) compensatory indicators unavailable
 
Operating Mode Applicability: Power Operations (Mode 1)
Startup (Mode 2)
 
Hot Standby (Mode 3)
 
Hot Shutdown (Mode 4)
Emergency Action Levels(s):
Note: The SM / TSC Director / EOF 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. UNPLANNED loss of > approximately 75% of the following  15 minutes:
Control Room annunciators associated with safety systems OR  Control Room safety system indication AND  b. Either of the following:
A SIGNIFICANT TRANSIENT is in progress OR  Compensatory indications are unavailable Deviations:
None.
Differences:
NEI 99-01 SA4 is renumbered to ANO SA6 for formatting purposes based on site preference for order of ICs alone.
 
An "and" and an additional "or" are used in the EAL for clarity.
 
Additional information is provided in the bases with the term "SPDS" to define computer based information.
 
Additional information is provided in the bases to define those systems associated with safety system indication. to 0CAN121102
 
Page 102 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document NEI 99-01:  SS1 Loss of all Off-site and all On-Site AC power to emergency busses for 15 minutes or longer.
 
Operating Mode Applicability:
Power Operation Startup Hot Standby
 
Hot Shutdown Example Emergency Action Level:
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.
 
ANO:  SS1 Loss of all offsite and all onsite AC power to Vital 4.16 KV busses  15 minutes Operating Mode Applicability: Power Operations (Mode 1)
Startup (Mode 2)
 
Hot Standby (Mode 3)
 
Hot Shutdown (Mode 4)
Emergency Action Level(s):
Note: The SM / TSC Director / EOF 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 offsite and all onsite AC power to Vital 4.16 KV busses  15 minutes.
Deviations:
None.
Differences:
The site-specific term "Vital 4.16 KV" is used in the IC and EAL to define emergency busses. to 0CAN121102
 
Page 103 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document NEI 99-01:  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 Mode Applicability:
Power Operation Startup  Example Emergency Action Level:
: 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).
 
ANO:  SS3 Automatic trip fails to shutdown the reactor and manual actions taken from the reactor control
 
console are not successful in shutting down the reactor.
 
Operating Mode Applicability:
Power Operations (Mode 1)
Startup (Mode 2)
Emergency Action Level(s):
: 1. a. An automatic trip failed to shutdown the reactor.
 
AND  b. Manual actions taken at the reactor control console do not shutdown the reactor as indicated by reactor power  5%. Deviations:
None.
Differences:
NEI 99-01 SS2 is renumbered to ANO SS3 for formatting purposes based on site preference for order of ICs alone.
 
Additional information is provided in the bases to define a failure of the manual trip function. to 0CAN121102
 
Page 104 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document NEI 99-01:  SS3 Loss of all vital DC power for 15 minutes or longer.
 
Operating Mode Applicability:
Power Operation Startup Hot Standby
 
Hot Shutdown Example Emergency Action Level:
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) on all (site-specific Vital DC busses) for 15 minutes or longer.
 
ANO:  SS4 Loss of all Vital DC power  15 minutes Operating Mode Applicability:
Power Operations (Mode 1)
Startup (Mode 2)
 
Hot Standby (Mode 3)
 
Hot Shutdown (Mode 4)
Emergency Action Level(s):
Note: The SM / TSC Director / EOF 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. < 105 volts on all Vital DC busses  15 minutes.
Deviations:
None.
Differences:
NEI 99-01 SS3 is renumbered to ANO SS4 for formatting purposes based on site preference for order of ICs alone. to 0CAN121102
 
Page 105 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document NEI 99-01:  SS6 Inability to Monitor a SIGNIFICANT TRANSIENT in Progress
 
Operating Mode Applicability:
Power Operation Startup Hot Standby
 
Hot Shutdown Example Emergency Action Level:
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.
to 0CAN121102
 
Page 106 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document ANO:  SS6 Inability to monitor a SIGNIFICANT TRANSIENT in progress
 
Operating Mode Applicability:
Power Operations (Mode 1)
Startup (Mode 2)
 
Hot Standby (Mode 3)
 
Hot Shutdown (Mode 4)
Emergency Action Level(s):
Note: The SM / TSC Director / EOF 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 > approximately 75% of the following  15 minutes:
Control Room annunciators associated with safety systems OR  Control Room safety system indication AND  b. A SIGNIFICANT TRANSIENT is in progress.
AND  c. Compensatory indications are unavailable.
Deviations:
None.
Differences:
Additional information is provided in the bases with the term "SPDS" to define computer based information.
 
Additional information is provided in the bases to define those systems associated with safety system indication. to 0CAN121102
 
Page 107 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document NEI 99-01:  SG1 Prolonged loss of all Off-site and all On-Site AC power to emergency busses
 
Operating Mode Applicability:
Power Operation Startup Hot Standby
 
Hot Shutdown Example Emergency Action Level:
: 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.)    to 0CAN121102
 
Page 108 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document ANO:  SG1 Prolonged loss of all offsite and all onsite AC power to Vital 4.16 KV busses
 
Operating Mode Applicability: Power Operations (Mode 1)
Startup (Mode 2)
 
Hot Standby (Mode 3)
 
Hot Shutdown (Mode 4)
Emergency Action Level(s):
: 1. a. Loss of all offsite and all onsite AC power to Vital 4.16 KV busses.
 
AND  b. Either of the following:
Restoration of at least one Vital 4.16 KV bus in < 4 hours is not likely OR  Continuing degradation of core cooling based on Fission Product Barrier monitoring as indicated by CETs  700 &deg;F. Deviations:
None.
Differences:
The site-specific term "Vital 4.16 KV" is used in the IC and EAL to define emergency busses.
 
An "or" is used in the EAL for clarity. to 0CAN121102
 
Page 109 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document NEI 99-01:  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.
 
Operating Mode Applicability:
Power Operation Startup  Example Emergency Action Level:
: 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.)    to 0CAN121102
 
Page 110 of 110
 
NEI 99-01 Revision 5 EAL Deviation-Differences Document ANO:  SG3 Automatic trip and all manual actions fail to shutdown the reactor and indication of an extreme
 
challenge to the ability to cool the core exists.
 
Operating Mode Applicability:
Power Operations (Mode 1)
Startup (Mode 2)
Emergency Action Level(s):
: 1. a. An automatic trip failed to shutdown the reactor.
 
AND  b. All manual actions do not shutdown the reactor as indicated by reactor power  5%. AND  c. Either of the following exist or have occurred due to continued power generation:
CET temperatures at or approaching 1200 &deg;F OR  Feedwater flow rate less than:
Unit 1:  430 gpm Unit 2:  485 gpm Deviations:
None.
Differences:
SG2 is renumbered to ANO IC SG3 for formatting purposes based on site preference for order of ICs alone.
 
An "or" is used in the EAL for clarity.
 
Attachment 2 to 0CAN121102 Proposed Technical Basis Document (Markup)    to 0CAN121102
 
Page 1 of 112
 
ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT AU1  Initiating Condition - NOTIFICATION OF UNUSUAL EVENT Any release of gaseous or liquid radioactivity to the environment greater than
> 2 times the Radiological Effluent Technical Specifications/
ODCM limits for > 60 minutes or longer. Operating Mode Applicability:
All  Example Emergency Action Level (s): (1 or 2 or 3 or 4 or 5)  Note: The SM    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 Channel 7 on ANY any of the following radiation monitors
> greater than the reading shown for
> 60 minutes or longer
: (site specific monitor list and threshold values)
MONITORS - UNIT 1 LIMIT RX-9820 Containment Purge 5.90E-2 &#xb5;Ci/cc RX-9825 Radwaste Area 5.36E-2 &#xb5;Ci/cc RX-9830 Fuel Handling Area 4.54E-2 &#xb5;Ci/cc RX-9835 Emergency Penetration Room 9.56E-1 &#xb5;Ci/cc MONITORS - UNIT 2 LIMIT 2RX-9820 Containment Purge 4.46E-2 &#xb5;Ci/cc 2RX-9825 Radwaste Area 3.32E-2 &#xb5;Ci/cc 2RX-9830 Fuel Handling Area 4.46E-2 &#xb5;Ci/cc 2RX-9835 Emergency Penetration Room 8.84E-1 &#xb5;Ci/cc 2RX-9840 Post Accident Sampling Building 4.42E-1 &#xb5;Ci/cc 2RX-9845 Aux. Building Extension 1.26E-1 &#xb5;Ci/cc 2RX-9850 Low Level Radwaste Storage Bldg.
1.77E-1 &#xb5;Ci/cc OR    to 0CAN121102
 
Page 2 of 112
 
ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT AU1  2. VALID reading on any of the following radiation monitors effluent monitor reading> greater than 2 times the alarm setpoint established by a current release radioactivity discharge permit for
> 60 minutes or longer. EFFLUENT MONITORS - Unit 1 RX-9820 Containment Purge (Channel 7 or 9)
RE-4830 Waste Gas Radiation Monitor RE-4642 Liquid Radwaste Monitor EFFLUENT MONITORS - Unit 2 2RX-9820 Containment Purge (Channel 7 or 9) 2RE-2429 Waste Gas Decay Tank Vent Line Radiation Monitor 2RE-2330 BMS Liquid Discharge Monitor 2RE-4423 Regenerative Waste Discharge Monitor 2RE-4425 SG Blowdown to Flume Radiation Monitor OR  3. Confirmed grab sample analyses for gaseous or liquid releases indicates concentrations or release rates
> greater than 2 times the applicable values of the ODCM (site specific RETS values) for > 60 minutes or longer. OR  4. VALID reading on perimeter radiation moni toring system reading greater than 0.10 mR/hr above normal* background for 60 minutes or longer. [for sites having telemetered perimeter monitors]  4 5VALID indication on automa tic 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 val ue. Basis:  [Refer to Appendix A for a detailed basis of the radiological effluent IC/EALs.
] The SM 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.
This IC addresses a potential reductiondecrease in the level of safety of the plant as indicated by a radiological release that exceeds regulatory commitments for an extended period of time.
ANO Nuclear power plants incorporate s features intended to control the release of radioactive effluents to the environment. Further, there are adm inistrative controls established to prevent unintentional releases, or control and monitor intentional releases.
[These controls are locate d    to 0CAN121102
 
Page 3 of 112
 
ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT AU1  in the Off
-site Dose Calculation Manual (ODCM), and for plants that have not implemented Generic Letter 89
-01, in the Radiological Effluent Technical Specifications (RETS).]
The occurrence of extended, uncontrolled radioactive releases to the environment is indicative of a degradation in these features and/or controls. 
[Some sites may find it advantageous to address gaseous and liquid releases with separate EALs.]The ODCM RETS multiples are specified in AU1 and AA1 only to distinguish between non-emergency conditions, and from each other. While these multiples obviously correspond to an offsite dose or dose rate, the emphasis in classifying these events is the degradation in the
 
level of safety of the plant, NOTnot the magnitude of the associated dose or dose rate. 
[Releases should not be prorated or averaged over 60 minutes. For example, a release exceeding 4 times x ODCM limits for 30 minutes does not meet the threshold for this IC
.]  This Initiating ConditionEAL includes any release for which a radioactivity discharge permit was not prepared, or a release that exceeds the c onditions (e.g., minimum dilution flow, maximum discharge flow, alarm setpoints, etc.) on the applicable permit.
EAL #1  This EAL addresses radioactivity releases, that for whatever reason, cause effluent radiation monitor readings to exceed the threshold identified in the EAL.
This EAL is intended for sites that have established effluent monitoring on non-routine release pathways for which a discharge permit would not normally be prepared. 
[The ODCM establishes a methodology for determining effluent radiation monitor setpoints. The ODCM specifies default source terms and, for gaseous releases, prescribes the use of pre
-determined annual average meteorology in the most limiting dow nwi nd sector for showing compliance with the regulatory commitments. This EAL should be determined using this methodology.
]  EAL #2  This EAL addresses radioactivity releases, that for whatever reason, cause effluent radiation monitor readings to exceed the threshold identified in this the I nitiating C ondition established by the release radioactivity discharge permit. This value may be associated with a planned batch release, or a continuous release path. 
[In either case, the value is established by the ODCM to warn of a release that is not in compliance with the RETS. Indexing the EAL to the ODCM setpoints in this manner insures that the EAL will never be less than the setpoint established by a specific discharge permit.
]    to 0CAN121102
 
Page 4 of 112
 
ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT AU1  EAL #3  This EAL addresses uncontrolled releases that are detected by sample analyses, particularly on unmonitored pathways, e.g., spills of radioactive liquids into storm drains, heat exchanger
 
leakage in river water systems, lake, etc. EAL s #4 and #5  The 0.10 mR/hr value in EAL #4, and t he site specific value for EAL #5,is based on a release rate not exceeding 500 mrem per year.
[As provided in the ODCM / RETS, prorated over 8766 hours, multiplied by two, and rounded.
(500 &#xf7; 8766 x 2 = 0.114).
] EAL #1 and #2 directly correlate with the IC since annual average meteorology is required to be used in showing compliance with the ODCM and is used in calculating the alarm setpoints.
EAL s #4 and #5 is are a function of actual meteorology, which will likely be different from the limiting annual av erage value. Thus, there will likely be a numerical inconsistency.
The underlying basis of this EAL involves the degradation in the level of safety of the plant implied by the uncontrolled release. Exceeding EAL #4 or #5  is an indication of an uncontroll ed release. Reference Documents:
: 1. 1604.051, "Eberline Radiation Monitor System" 2. Offsite Dose Calculation Manual    to 0CAN121102
 
Page 5 of 112
 
ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT AU2  Initiating Condition - NOTIFICATION OF UNUSUAL EVENT UNPLANNED rise in plant radiation levels  Operating Mode Applicability:
All  Example Emergency Action Level (s): (1 or 2) 
: 1.      a. UNPLANNED lowering of water level drop in the a reactor refueling canal or spent fuel pool pathway as indicated by (site specific level or indication)
:  Personnel observation, refueling crew report, indication on area security camera, borated water source (BWST or RWT) level drop due to makeup demands.
AND  b. VALID Area Radiation Monitor reading rise on any of the following: (site specific list). Unit 1 RE-8009 Spent Fuel Area RE-8017 Fuel Handling Area Unit 2 2RE-8914 Spent Fuel Area 2RE-8915 Spent Fuel Area 2RE-8916 Spent Fuel Area 2RE-8912 Containment Incore Instrumentation OR 2. UNPLANNED VALID Area Radiation Monitor readi ngs or survey results indicate a rise by a factor of 1000 over normal* levels.
NOTE:  For area radiation monitors with ranges incapable of measuring 1000 times normal* levels, classification shall be based on VALID full scale indication unless surveys confirm that area radiation levels are below 1000 times normal* within 15 minutes of the Area Radiation Monitor indications going to full scale indication.
*Normal can be considered as the highest reading in the past twenty-four hours excluding the
 
current peak value.
 
Basis:  This IC addresses elevatincreas ed radiation levels as a result of lowered water level d rop ecrease s above irradiated fuel or events that have resulted, or may result, in UNPLANNED risincreas es in radiation dose rates within plant buildings. These radiation riincrea ses represent a loss of control over radioactive material and represent a potential degradation in the level of safety of the plant. to 0CAN121102
 
Page 6 of 112
 
ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT AU2  EAL #1  [Site specific indications may include instrumentation such as water level and local area radiation monitors, and personnel (e.g., refueling crew) reports. If available, video cameras may allow remote observation. Depending on available level instrumentation, the declaration threshold may need to be based on indications of water makeup rate or decrease in water storage tank level.
]  [In light of Reactor Cavi ty Seal failure incidents at two different PWRs and loss of water in the Spent Fuel Pit/Fuel Transfer Canal at a BWR, explicit coverage of these types of events via threshold #1 is appropriate given their potential for increased doses to plant staff.
] The refueling pathway is a site specific combination of cavities, tubes, canals and pools. While a radiation monitor could detect a risen increa se in dose rate due to a drop in the water level, it might not be a reliable indication of whether or not the fuel is covered.
[For example, a refueling bridge ARM reading may rise due to planned evolutions such as head lift, or even a fuel assembly being raised in the manipulator mast. Also, a monitor could in fact be properly responding to a known event involving transfer or relocation of a source, stored in or near the fuel pool or responding to a planned evolution such as removal of the reactor head.
Generally, elevated radiation monitor indications will need to be combined with another indicator (or personnel report) of water loss
.] [Application of this EAL requires understanding of the actual radiological conditions present in the vicinity of the monitor. Information Notice No. 90
-08, "KR-85 Hazards from Decayed Fuel" should be considered in establishing radiation monitor EALs.
] For refueling events where the water level drops below the RPV flange classification would be via CU2. This event escalates to an Alert per AA2 if irradiated fuel outside the reactor vessel is
 
uncovered. For events involving irradiated fuel in the reactor vessel, escalation would be via the
 
Fission Product Barrier MatrixTable for events in operating modes 1-4.
EAL #2  This EAL addresses risincreas es 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.
This EAL excludes radiation level risincreas es that result from planned activities such as use of radiographic sources and movement of radioactive waste materials. A specific list of ARMs is not required as it would restrict the applicability of the Threshold. The intent is to identify loss of
 
control of radioactive material in any monitored area. to 0CAN121102
 
Page 7 of 112
 
ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT AA1  Initiating Condition - ALERT Any release of gaseous or liquid radioactivity to the environment
> greater than 200 times the Radiological Effluent Technical Specifications/
ODCM limits for  > 15 minutes or longer
. Operating Mode Applicability:
All  Example Emergency Action Level (s): (1 or 2 or 3 or 4 or 5)  Note: The SM / TSC Director / EOF Director 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 Channel 7 on any ANY of the following radiation monitors
>greater than the reading shown for
> 15 minutes or longer
:  (site specific monitor list and threshold values)
MONITORS - UNIT 1 LIMIT RX-9820 Containment Purge 5.90E0 &#xb5;Ci/cc RX-9825 Radwaste Area 5.36E0 &#xb5;Ci/cc RX-9830 Fuel Handling Area 4.54E0 &#xb5;Ci/cc RX-9835 Emergency Penetration Room 9.56E+1 &#xb5;Ci/cc MONITORS - UNIT 2 LIMIT 2RX-9820 Containment Purge 4.46E0 &#xb5;Ci/cc 2RX-9825 Radwaste Area 3.32E0 &#xb5;Ci/cc 2RX-9830 Fuel Handling Area 4.46E0 &#xb5;Ci/cc 2RX-9835 Emergency Penetration Room 8.84E+1 &#xb5;Ci/cc 2RX-9840 Post Accident Sampling Building 4.42E+1 &#xb5;Ci/cc  2RX-9845 Aux. Building Extension 1.26E+1 &#xb5;Ci/cc 2RX-9850 Low Level Radwaste Storage Bldg.
1.77E+1 &#xb5;Ci/cc OR    to 0CAN121102
 
Page 8 of 112
 
ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT AA1  2. EITHER VALID reading on any of the following radiation monitors effluent monitor
>reading greater than 200 times the alarm setpoint established by a current release radioactivity discharge permit for
> 15 minutes or longer OR VALID reading greater than the value listed for
> 15 minutes. MONITORS - UNIT 1 LIMIT RX-9820 Containment Purge (Channel 7 or 9)
N/A RE-4830 Waste Gas Radiation Monitor 9.5E7 cpm RE-4642 Liquid Radwaste Monitor 9.5E7 cpm MONITORS - UNIT 2 LIMIT 2RX-9820 Containment Purge (Channel 7 or 9)
N/A 2RE-2429 Waste Gas Monitoring System 9.5E5 cpm 2RE-2330 BMS Liquid Discharge Monitor 9.5E5 cpm 2RE-4423 Regenerative Waste Discharge Monitor 9.5E5 cpm 2RE-4425 SG Blowdown to Flume Radiation Monitor 9.5E5 cpm OR 3. Confirmed grab sample analyses for gaseous or liquid releases indicates concentrations or release rates
> greater than 200 times the applicable values of the ODCM (site specific RETS values) for > 15 minutes or longer. OR 4. VALID reading on perimeter radiation moni toring system reading greater than 10.0 mR/hr above normal* background for 15 minutes or longer. [for sites having telemetered perime ter monitors]
4 5VALID 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 twent y-four hours excluding the current peak value.  [Refer to Appendix A for a detailed basis of the radiological effluent IC/EALs.
] Basis:  The SM / TSC Director / EOF Director 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.
This IC addresses an actual or substantial potential reductiondecrease in the level of safety of the plant as indicated by a radiological release that exceeds regulatory commitments for an extended period of time.
ANONuclear power plants incorporate s features intended to control the release of radioactive effluents to the environment. Further, there are administrative controls established to prevent unintentional releases, or control and monitor intentional releases. 
 
[These controls are located in the Off
-site Dose Calculation Manual (ODCM), and for plants that have not implemented Generic Letter 89
-01, in the Radiological Effluent Technical Specific ations (RETS).]
The occurrence of extended, uncontrolled radioactive releases to the environment is indicative of a degradation in these features and/or controls.
[Some sites may find it advantageous to address gaseous and liquid releases with separate EA Ls.]    to 0CAN121102
 
Page 9 of 112
 
ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT AA1  The ODCM RETS multiples are specified in AU1 and AA1 only to distinguish between non-emergency conditions, and from each other. While these multiples obviously correspond to an offsite dose or dose rate, the emphasis in classifying these events is the degradation in the level
 
of safety of the plant, NOTnot the magnitude of the associated dose or dose rate. 
[To ensure a realistic near
-linear escalation path, a value should be selected roughly half
-way between the AU1 value and the value calculated for AS1 value. The value will be based on radiation monitor readings to exceed 200 times the Technical Specification limit and releases are not terminated within 15 minutes.
The ODCM establishes a methodology for determining effluent radiation monitor setpoints. The ODCM specifies default source terms and, for gaseous releases, prescribes the use of pre
-determined annual average meteorology in the most limiting downwind sector for showing compliance with the regulatory commitments. This EAL can be determined using this methodology if appropriate.
] [Releases should not be prorated or averaged. For example, a release exceeding 600 times ODCM limits for 5 minutes does not meet the threshold for this IC.]
This Initiating ConditionEAL includes any release for which a release 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. 
 
EAL #1 This EAL addresses radioactivity releases, that for whatever reason, cause effluent radiation
 
monitor readings to exceed the threshold identified in the Initiating ConditionIC. This EAL is intended for sites that have established effluent monitoring on non-routine release
 
pathways for which a discharge permit would not normally be prepared.
  [The ODCM establishes a methodology for determining effluent radiation monitor setpoints. The ODCM specifies default source terms and, for gaseous releases, prescribes the use of pre
-determined annual average meteorology in the most limiting downwind sector for showing compliance with the regulatory commitments. This EAL should be determined using this methodology.
]  EAL #2  This EAL addresses radioactivity releases, that for whatever reason, cause effluent radiation monitor readings to exceed the threshold identified in th ise I nitiating C ondition established by the radioactivity discharge permit. This value may be associated with a planned batch release, or a continuous release path.
The limit values provided are for those cases in which the maximum monitor range is less than the release permit value multiplied by 200.
[In either case, the value is established by the ODCM to warn of a release that is not in compliance with the RETS. Indexing the EAL to the ODCM setpoints in this manner insures that the EAL will never be less than the setpoint established by a specific discharge permit.
]    to 0CAN121102
 
Page 10 of 112
 
ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT AA1  EAL #3  This EAL addresses uncontrolled releases that are detected by sample analyses, particularly on unmonitored pathways, e.g., spills of radioactive liquids into storm drains, heat exchanger
 
leakage in river water systems, lake, etc. EAL s #4 and #5 The 10.0 mR/hr value in EAL #4 , and the site specific value for EAL #5,is based on a release rate not exceeding 500 mrem per ye ar.  [As provided in the ODCM / RETS, prorated over 8766 hours, multiplied by 200, and rounded.
(500 &#xf7; 8766 x 200 = 11.4)
]. EAL #1 and #2 directly correlate with the IC since annual average meteorology is required to be used in showing compliance with the ODCM and is used in calculating the alarm setpoints.
 
EAL s #4 and #5 is are a function of actual meteorology, which will likely be different from the limiting annual average value. Thus, there will likely be a numerical inconsistency.
The underlying basis of this EAL involves the degradation in the level of safety of the plant implied by the uncontrolled release. Exceeding EAL #4 or #5 is an indication of an uncontrolled release. The underlying basis of this EAL involves the degradation in the level of safe ty of the plant implied by the uncontrolled release. Exceeding EAL #4 or #5  is an indication of an uncontrolled release. Reference Documents:
: 1. 1604.051, "Eberline Radiation Monitor System" 2. Offsite Dose Calculation Manual to 0CAN121102
 
Page 11 of 112
 
ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT AA2  Initiating Condition - ALERT 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 Mode Applicability:
All  Example Emergency Action Level (s): (1 or 2) 
: 1. A water level drop in the reactor refueling canalcavity, or 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 any of the following radiation monitors due to damage to irradiated fuel or loss of water level.  (site specific radiation monitors)
Unit 1 RX-9820 Containment Purge (Channel 7 or 9) RX-9825 Radwaste Area (Channel 7 or 9) RX-9830 Fuel Handling Area (Channel 7 or 9)
RE-8060 Containment High Range Radiation Monitors RE-8061 Containment High Range Radiation Monitors RE-8009 Spent Fuel Area RE-8017 Fuel Handling Unit 2 2RX-9820 Containment Purge (Channel 7 or 9) 2RX-9825 Radwaste Area (Channel 7 or 9) 2RX-9830 Fuel Handling Area (Channel 7 or 9) 2RE-8905 Containment Equipment Hatch Area 2RE-8909 Containment Personnel Access Area 2RE-8925-1 Containment High Range Radiation Monitors 2RE-8925-2 Containment High Range Radiation Monitors 2RE-8914 Spent Fuel Area 2RE-8915 Spent Fuel Area 2RE-8916 Spent Fuel Area 2RE-8912 Containment Incore Inst.
Basis:  This IC addresses risincreas es in radiation dose rates within plant buildings, and may be a precursor to a radioactivity release to the env ironment. These events represent a loss of control over radioactive material and represent an actual or substantial potential degradation in the level
 
of safety of the plant. to 0CAN121102
 
Page 12 of 112
 
ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT AA2  [These events escalate from AU2 in that fuel activity has been released, or is anticipated due to fuel heatup. This IC applies to spent fuel requiring water coverage and is not intended to address spent fuel which is licensed for dry storage. ]
EAL #1  [Indications may include instrumentation such as water level and local area radiation monitors, and personnel (e.g., refueling crew) reports. Depending on available level indication, the declaration may be based on indications of water makeup rate or drop in applicable borated water storage tank level. Video cameras (Security or outage-related) may allow remote observation of level.]
  [In light of R eactor Cavity Seal failure incidents at two different PWRs and loss of water in the Spent Fuel Pit/Fuel Transfer Canal at a BWR, explicit coverage of these types of events via threshold #1 is appropriate given their potential for increased doses to plant s taff.]  EAL #2 This EAL addresses radiation monitor indications of fuel uncovery and/or fuel damage.
ElevatIncreas ed ventilation monitor readings may be indication of a radioactivity release from the fuel, confirming that damage has occurred.
ElevatIncr eas ed background at the ventilation monitor due to water level dropdecrease may mask elevatincreas ed ventilation exhaust airborne activity and needs to be considered.
While a radiation monitor could detect a risen increase in dose rate due to a drop in the water level, it might not be a reliable indication of whether or not the fuel is covered. 
[For example, a refueling bridge ARM reading may rise due to planned evolutions such as head lift, or even a fuel assembly being raised in the manipulator mast. Also, a monitor could in fact be properly responding to a known event involving transfer or relocation of a source, stored in or near the fuel pool or responding to a planned evolution such as removal of the reactor head.
Generally, elevated radiation monitor indications will need to be combined with another indicator (or personnel report) of water loss.]
[Application of this EAL requires understanding of the actual radiological conditions present in the vicinity of the monitor. Information Notice No. 90
-08, " KR-85 Hazards from Decayed Fuel" should be considered in establishing radiation monitor EALs.
] Escalation of this emergency classification level, if appropriate, would be based on AS1 or AG1. to 0CAN121102
 
Page 13 of 112
 
ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT AA3  Initiating Condition - ALERT Rise in radiation levels within the facility that impedes operation of systems required to maintain plant safety functions  Operating Mode Applicability:
All  Example Emergency Action Level (s): (1 or 2)  Dose rate
>greater than 15 mR/hr in ANY any of the following areas requiring continuous occupancy to maintain plant safety functions:  (site specific area list)
Unit 1 Control Room  Unit 2 Control Room  Central Alarm Station Basis:  This IC addresses elevatincreas ed radiation levels that impact continued operation in areas requiring continuous occupancy to maintain safe operation or to perform a safe shutdown.
The cause and/or magnitude of the riseincrease in radiation levels is not a concern of this IC.
The SM/TSC Director/EOF DirectorEmergency Director must consider the source or cause of the elevatincreas ed radiation levels and determine if any other IC may be involved.
[At multiple
-unit sites, the EALs could result in declaration of an Alert at one unit due to a radioactivity release or radiation shine resulting from a major accident at the other unit. This is appropriate if the increase impairs operations at the operating unit.
] [This IC is not meant to apply to rises in the containment dome radiation monitors as these are events which are addressed in the fission product barrier matrix EALs.]
[The value of 15mR/hr is derived from the GDC 19 value of 5 rem in 30 days with adjustment for expected occupancy times. Although Section III.D.3 of NUREG
-0737, "Clarification of TMI Action Plan Requirements", provides that the 15 mR/hr value can b e averaged over the 30 days, the value is used here without averaging, as a 30 day duration implies an event potentially more significant than an Alert
.] Areas requiring continuous occupancy include the Cc ontrol Rr ooms and the Central Alarm Station. and, a s appropriate to the site, any other control stations that are staffed continuously, such as a radwaste control room, or a security alarm station.
[Typically these areas are the Control Room and the Central Alarm Station (CAS).]
to 0CAN121102
 
Page 14 of 112
 
ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT AS1  Initiating Condition -- SITE AREA EMERGENCY Offsite dose resulting from an actual or IMMINENT release of gaseous radioactivity
> greater than 100 m Rrem TEDE or 500 m Rrem child tT hyroid CDE for the actual or projected duration of the release Operating Mode Applicability:
All  Example Emergency Action Level (s): (1 or 2 or 3 or 4)  Note: The SM/TSC Director/EOF Director 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, the classification declaration should be based on EAL #2 dose assessment instead of EAL #1radiation monitor values. Do not delay declaration awaiting dose assessment results. 
: 1. VALID reading on Channel 9 on any ANY of the following radiation monitors
> greater than the reading shown for
> 15 minutes or longer
:  (site specific monitor list and threshold values)
MONITORS - UNIT 1 LIMIT RX-9820 Containment Purge 5.90E+1 &#xb5;Ci/cc RX-9825 Radwaste Area 5.36E+1 &#xb5;Ci/cc RX-9830 Fuel Handling Area 4.54E+1 &#xb5;Ci/cc RX-9835 Emergency Penetration Room 9.56E+2 &#xb5;Ci/cc MONITORS - UNIT 2 LIMIT 2RX-9820 Containment Purge 4.46E+1 &#xb5;Ci/cc 2RX-9825 Radwaste Area 3.32E+1 &#xb5;Ci/cc 2RX-9830 Fuel Handling Area 4.46E+1 &#xb5;Ci/cc 2RX-9835 Emergency Penetration Room 8.84E+2 &#xb5;Ci/cc 2RX-9840 Post Accident Sampling Building 4.42E+2 &#xb5;Ci/cc 2RX-9845 Aux. Building Extension 1.26E+2 &#xb5;Ci/cc 2RX-9850 Low Level Radwaste Storage Bldg.
1.77E+2 &#xb5;Ci/cc OR  2. Dose assessment using actual meteorology indicates doses
> greater than 100 m Rrem TEDE or 500 m Rrem child thyroid CDE at or beyond the site boundary. 
: 3. VALID perimeter radiation monitoring system reading greater than 100 mR/hr for 15 minutes or longer. [for sites ha ving telemetered perimeter monitors]
to 0CAN121102
 
Page 15 of 112
 
ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT AS1  OR 34. 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 child thyroid CDE > greater than 500 m Rrem for one hour of inhalation, at or beyond the site boundary.
Basis:  [Refer to Appendix A for a detailed basis of the radiological effluent IC/EALs.
]  This IC addresses radioactivity releases that result in doses at or beyond the site boundary that exceed 10% of the EPA Protective Action Guides (PAGs). Releases of this magnitude are
 
associated with the failure of plant systems needed for the protection of the public. 
[While these failures are addressed by other ICs, this IC provides appropriate divers ity and addresses events which may not be able to be classified on the basis of plant status alone. It is important to note that for the more severe accidents the release may be unmonitored or there may be large uncertainties associated with the source term and/or meteorology.
]  [The EPA PAGs are expressed in terms of the sum of the effective dose equivalent (EDE) and the committed effective dose equivalent (CEDE), or as the thyroid committed dose equivalent (CDE). For the purpose of these IC/EALs, the dose quantity total effective dose equivalent (TEDE), as defined in 10 CFR 20, is used in lieu of "-sum of EDE and CEDE.-" The EPA PAG guidance provides for the use adult thyroid dose conversion factors. However, some states have decided to calculate child thy roid CDE. Utility IC/EALs need to be consistent with those of the states involved in the facility's emergency planning zone
.]  [The TEDE dose is set at 10% of the EPA PAG, while the 500 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.
]  EAL #1 The site specific monitor list in EAL #1 should include s effluent monitors on all potential release pathways (plant stack, primary-secondary leak, fuel handling accident). 
[The monitor reading EALs should be determined using a dose assessment method that back calculates from the dose values specified in the IC. Since doses are generally not monitored in real-time, it is suggested that a release duration of one hour be assumed, and that the EALs be based on a site specific boundary (or beyond) dose of 100 mrem whole body or 500 mrem thyroid in one hour, whichever is more limiting (as was done for EALs #2 and #4). If individual site analyses indicate a longer or shorter duration for the period in which the su bstantial portion of the activity is released, the longer duration should be used.
]  [The meteorology used should be the same as those used for determining AU1 and AA1 monitor reading EALs. The same source term (noble gases, particulates, and halogens) may also be used as long as it maintains a realistic and near linear escalation between the EALs for the four classifications. If proper escalations do not result from the use of the same source term, if the calculated values are unrealistically high, or if correlation between the values and    to 0CAN121102
 
Page 16 of 112
 
ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT AS1  dose assessment values does not exist, then consider using an accident source term for AS1 and AG1 calculations.
]  EAL #2 Since dose assessment in EAL #2 is based on actual meteorology, whereas the monitor reading s in EAL #1 areis not, the results from these assessments may indicate that the classification is not warranted, or may indicate that a higher classification is warranted. For this reason, emergency implementing procedures should call for the ti mely performance of dose assessments using actual meteorology and release information. If the results of these dose assessments are
 
available when the classification is made (e.g., initiated at a lower classification level), the dose
 
assessment results override the monitor reading EAL
: s. EAL #3 Field team surveys in EAL #3 should be perform ed at or beyond the SITE BOUNDARY and at the most accurate indicator of the condition. Field data are independent of release elevation and meteorology. The assumed release duration is one hour. Expected post accident source terms would be dominated by noble gases providing the dose rate value. Sampling of radioiodine by adsorption on a charcoal cartridge should determine the iodine value.
Reference Documents:
: 1. 1604.051, "Eberline Radiation Monitor System" 2. Offsite Dose Calculation Manual to 0CAN121102
 
Page 17 of 112
 
ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT AG1  Initiating Condition -- GENERAL EMERGENCY Offsite dose resulting from an actual or IMMINENT release of gaseous radioactivity
> greater than 1000 m Rrem TEDE or 5000 m Rrem child tT hyroid CDE for the actual or projected duration of the release using actual meteorology  Operating Mode Applicability:
All  Example Emergency Action Level (s): (1 or 2 or 3 or 4)  Note: The SM/TSC Director/EOF Director 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, the classification declaration should be based on EAL #2 dose assessment instead of EAL #1 radiation monitor values. Do not delay declaration awaiting dose assessment results. 
: 1. VALID reading on Channel 9 on any ANY of the following radiation monitors
> greater than the reading shown for
> 15 minutes or longer
:  (site specific monito r list and threshold values)
MONITORS - UNIT 1 LIMIT RX-9820 Containment Purge 5.90E+2 (Ci/cc) RX-9825 Radwaste Area 5.36E+2 (Ci/cc) RX-9830 Fuel Handling Area 4.54E+2 (Ci/cc) RX-9835 Emergency Penetration Room 9.56E+3 (Ci/cc)  MONITORS - UNIT 2 LIMIT 2RX-9820 Containment Purge 4.46E+2 (Ci/cc) 2RX-9825 Radwaste Area 3.32E+2 (Ci/cc) 2RX-9830 Fuel Handling Area 4.46E+2 (Ci/cc) 2RX-9835 Emergency Penetration Room 8.84E+3 (Ci/cc) 2RX-9840 Post Accident Sampling Building 4.42E+3 (Ci/cc) 2RX-9845 Aux. Building Extension 1.26E+3 (Ci/cc) 2RX-9850 Low Level Radwaste Storage Building 1.77E+3 (Ci/cc)  OR  2. Dose assessment using actual meteorology indicates doses
> greater than 1000 m Rrem TEDE or 5000 m Rrem child thyroid CDE at or beyond the site boundary. 
: 3. VALID perimeter radiation monitoring system reading greater than 1000 mR/hr for 15 minutes or longer. [for sites having telemetered perimeter monitors]
OR    to 0CAN121102
 
Page 18 of 112
 
ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT AG1  34. 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 child thyroid CDE > greater than 5000 m Rrem for one hour of inhalation, at or beyond the site boundary.
Basis:  [Refer to Appendix A for a detailed basis o f the radiological effluent IC/EALs.
]  This IC addresses radioactivity releases that result in doses at or beyond the site boundary that exceed the EPA Protective Action Guides (PAGs).
Public protective actions will be necessary.
Releases of this magnitude are associated with the failure of plant systems needed for the
 
protection of the public and likely involve fuel damage.
[While these failures are addressed by other ICs, this IC provides appropriate diversity and addresses events which may not be able t o be classified on the basis of plant status alone. It is important to note that for the more severe accidents the release may be unmonitored or there may be large uncertainties associated with the source term and/or meteorology.
] [The EPA PAGs are expressed in terms of the sum of the effective dose equivalent (EDE) and the committed effective dose equivalent (CEDE), or as the thyroid committed dose equivalent (CDE). For the purpose of these IC/EALs, the dose quantity total effective dose equivalent (TEDE), as defined in 10 CFR 20, is used in lieu of "-sum of EDE and CEDE.-" The EPA PAG guidance provides for the use adult thyroid dose conversion factors. However, some states have decided to calculate child thyroid CDE. Utility IC/EALs need to be consistent w ith those of the states involved in the facilities emergency planning zone.
]  [The TEDE dose is set at the EPA PAG, while the 5000 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.
] EAL #1  The site specific monitor list in EAL #1 should include s effluent monitors on all potential release pathways (plant stack, primary-secondary leak, fuel handling accident). [The monitor reading EALs should be determined using a dose assessment method that back calculates from the dose values specified in the IC. Since doses are generally not monitored in real-time, it is suggested that a release duration of one hour be assumed, and that the EALs be based on a site specific boundary (or beyond) dose of 1000 mrem whole body or 5000 mrem thyroid in one hour, whichever is more limiting (as was done for EALs #2 and #4). If individual site analyses indicate a longer or shorter duration for the period in which the substantial portion of the activity is released, the longer duration should be used.
]  [The meteorology used should be the same as those used for determining AU1 and AA1 monitor reading EALs. The same source term (noble gases, particulates, and halogens) may also be used as long as it maintains a realistic and near linear escalation between the EALs for the four classifications. If proper escalations do not result from the use of the same source term, if the calculated values are unrealistically high, or if correlation between the values and    to 0CAN121102
 
Page 19 of 112
 
ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT AG1  dose assessment values does not exist, then consider using an accident source term for AS1 and AG1 calculations.
] EAL #2  Since dose assessment in EAL #2 is based on actual meteorology, whereas the monitor reading s in EAL #1 areis not, the results from these assessments may indicate that the classification is not warranted , or may indicate that a higher classification is warranted. For this reason, emergency implementing procedures should call for the ti mely performance of dose assessments using actual meteorology and release information. If the results of these dose assessments are
 
available when the classification is made (e.g., initiated at a lower classification level), the dose
 
assessment results override the monitor reading EAL
: s. EAL #3 Field team surveys in EAL #3 should be perform ed at or beyond the SITE BOUNDARY and at the most accurate indicator of the condition. Field data are independent of release elevation and meteorology. The assumed release duration is one hour. Expected post accident source terms would be dominated by noble gases providing the dose rate value. Sampling of radioiodine by adsorption on a charcoal cartridge should determine the iodine value.
Reference Documents:
: 1. 1604.051, "Eberline Radiation Monitor System" 2. Offsite Dose Calculation Manual    to 0CAN121102
 
Page 20 of 112
 
COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION CU1  Initiating Condition -- NOTIFICATION OF UNUSUAL EVENT RCS leakage Operating Mode Applicability:
Cold Shutdown (Mode 5)  Example Emergency Action Level (s):  Note: The SMEmergency 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 o r longer. [
BWR] 1. RCS leakage results in the inability to maintain or restore level within Pressurizer or RCS level target band(site specific pressurizer or RCS/RPV level target band) for > 15 minutes or longer. [PWR]  Basis:  This IC is considered to be a potential degradation of the level of safety of the plant. The inability to maintain or restore level is indicative of loss of RCS inventory.
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.
Prolonged loss of RCS Inventory may result in escalation to the Alert emergency classification level via either CA1 or CA 34 [The difference between CU 1 and CU2 deals with the RCS conditions that exist between cold shutdown and refueling modes. In the refueling mode the RCS is not intact and RPV level and inventory are monitored by different means. In cold shutdown the RCS will normally be intact and standard RCS inventory and level monitoring means are available.
] . to 0CAN121102
 
Page 21 of 112
 
COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION CU2  Initiating Condition -- NOTIFICATION OF UNUSUAL EVENT UNPLANNED loss of RCS/
reactor vesselRPV inventory  Operating Mode Applicability:
Refueling (Mode 6) Example Emergency Action Level (s): (1 or 2)
Note: The SMEmergency 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/
reactor vessel RPV level drop as indicated by either of the following: 
: a. RCS/reactor vesselRPV water level drop below the reactor vessel RPV flange for
> 15 minutes  or longer when the RCS/
reactor vessel RPV level band is established above the reactor vesselRPV flange  OR  b. RCS/reactor vesselRPV water level drop below the RCS
/reactor vessel level band for
> 15 minutes or longer when the RCS/
reactor vessel RPV level band is established below the reactor vesselRPV flange.
OR  2. RCS/reactor vesselRPV level cannot be monitored with a loss of RCS/
reactor vessel RPV inventory as indicated by an unexplained level rise in (as applicable) the Reactor Building Sump, Reactor Drain Tank, Aux. Building Equipment Drain Tank, Aux. Building Sump, or Quench Tank(site specific sump or tank). Basis:  This IC is a precursor of more serious conditions and considered to be a potential degradation of the level of safety of the plant.
Refueling evolutions that lowerdecrease RCS water level below the reactor vesselRPV flange are carefully planned and procedurally controlled. An UNPLANNED event that results in water level droppdecreas ing below the reactor vessel RP V flange, or below the planned RCS water level for the given evolution (if the planned RCS water level is already below the reactor vessel RPV flange), warrants declaration of a n N O UE 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. to 0CAN121102
 
Page 22 of 112
 
COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION CU2  Continued loss of RCS Inventory will result in escalation to the Alert emergency classification level via either CA1 or CA
: 34.  [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 This EAL involves a dropdecrease in RCS level below the top of the reactor vessel RPV flange that continues for 15 minutes due to an UNPLANNED event. This EAL is not applicable to dropdecrease s in flooded reactor cavity level, which is addressed by AU2 EAL1, until such time as the level dropdec rease s to the level of the vessel flange.
  [For BWRs] if RPV level continues to decrease and reaches the Low
-Low ECCS Actuation Setpoint then escalation to CA1 would be appropriate.
 
[For PWRs] If reactor vesselRPV level continues to dropdecrease and reaches the Bottom ID of the RCS Loop then escalation to CA1 would be appropriate.
EAL #2  This EAL addresses conditions in the refueling mode when normal means of core temperature indication and RCS level indication may not be available. Redundant means of reactor vesselRPV 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 reactor vesselRPV inventory loss was occurring by observing sump and tank level changes. Sump and tank level risincreas es 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 the Alert emergency classification level would be via either CA1 or CA
: 34. to 0CAN121102
 
Page 23 of 112
 
COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION CU 53  Initiating Condition -- NOTIFICATION OF UNUSUAL EVENT AC power capability to Vital 4.16 KVemergency busses reduced to a single power source
> for 15 minutes or l onger such that any additional single failure would result in station blackout Operating Mode Applicability:
Cold Shutdown (Mode 5)
Refueling (Mode 6)  Example Emergency Action Level (s):  Note: The SMEmergency 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 Vital 4.16 KV (site specific emergency busses) reduced to a single power source
> f or 15 minutes or longer. AND  b. Any additional single power source failure will result in station blackout.
Basis:  The condition indicated by this IC is the degradat ion of the offsite and onsite AC power systems such that any additional single failure would result in a station blackout. This condition could
 
occur due to a loss of offsite power with a concurrent failure of all but one emergency generator
 
to supply power to its emergency busses. The subsequent loss of this single power source would
 
escalate the event to an Alert in accordance with CA
: 53. Fifteen minutes was selected as a threshold to exclude transient or momentary losses of power.
The EAL allows credit for operation of the Alternate AC Diesel Generator.
  [At multi-unit stations, the EALs should allow credit for operation of installed design features, such as cross
-ties or swing diesels, provided that abnormal or emergency operating procedures address their use. However, these stations must also consider the impact of this conditio n on other shared safety functions in developing the site specific EAL.
]  [Plants that have a proceduralized capability to cross
-tie AC power from an off
-site power supply of a companion unit may take credit for the redundant power source in the associated EAL for this IC.]  Reference Documents:
: 1. 1202.007, "
Degraded Powe r" 2. 1202.008, "
Blackout" 3. 2202.007, "
Loss of Off-Site Power"
: 4. 2202.008, "
Station Blackout"
: 5. 2104.037, "
Alternate AC Diesel Generator Operations" to 0CAN121102
 
Page 24 of 112
 
COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION CU 34  Initiating Condition -- NOTIFICATION OF UNUSUAL EVENT UNPLANNED loss of decay heat removal capability with irradiated fuel in the reactor vessel RPV
. Operating Mode Applicability:
Cold Shutdown (Mode 5)            Refueling (Mode 6)  Example Emergency Action Level (s): (1 or 2)
Note:  The SMEmer gency 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 200 &deg;Fthe Te chnical Specification cold shutdown temperature limit. OR  2. Loss of all RCS temperature and RCS/
reactor vesselRPV level indication for
> 15 minutes or longer. Basis:  This IC is 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. 
[Entry into cold shutdown conditions may be attained within hours of operating at pow er. Entry into the refueling mode procedurally may not occur for typically 100 hours (site specific) 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 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 reactor vesselRPV will normally be maintained above the reactor vesselRPV flange. Refueling evolutions that lower decrease water level below the reactor vesselRPV flange are carefully planned and procedurally controlled. Loss of forced decay heat removal at reduced inventory may result in more rapid risincreas es in RCS/
reactor vesselRPV temperatures depending on the time since shutdown. to 0CAN121102
 
Page 25 of 112
 
COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION CU 34  N[Unlike the cold shutdown mode,] n ormal means of core temperature indication and RCS level indication may not be available in the refueling mode. Redundant means of reactor vesselRPV 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 of refueling modes, EAL 2 would result in declaration of a n NUE NOUE if both temperature and 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 CA 34 based on exceeding its temperature criteria. to 0CAN121102
 
Page 26 of 112
 
COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION CU 86  Initiating Condition -- NOTIFICATION OF UNUSUAL EVENT Loss of all oO nsite or oO ffsite communications capabilities  Operating Mode Applicability:
Cold Shutdown (Mode 5)            Refueling (Mode 6)            Defueled Example Emergency Action Level (s): (1 or 2) 
: 1. Loss of all Table C2of the followi ng onsite communication methods affecting the ability to perform routine operations
.  (site specific list of communications methods)
OR  2. Loss of all Table C3 of the following offsite communication methods affecting the ability to perform offsite notifications
.  (site specific list of communications methods)
Table C2 Onsite Communications Methods  Table C3 Offsite Communications Methods Station radio system All telephone lines (commercial and microwave)
Plant paging system ENS In-plant telephones Gaitronics Basis:  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 issues with offsite authorities. The loss of off-
 
site communications ability is expected to be significantly more comprehensive than the condition
 
addressed by 10 CFR 50.72.
The availability of one method of ordinary offsite communications is sufficient to inform federal, state, and local authorities of plant issues. This EAL is intended to be used only when
 
extraordinary means (e.g., relaying of information from radio transmissions, individuals being sent
 
to offsite locations, etc.) are being utilized to make communications possible.
[Site specific list for on
-site communications loss must encompass the loss of all means of routine communications (e.g., commercial telephones, sound powered phone systems, page party system and radios / walkie talk ies). Site specific list for off
-site communications loss must encompass the loss of all means of communications with off
-site authorities. This should include the ENS, commercial telephone lines, telecopy transmissions, and dedicated phone systems.
]    to 0CAN121102
 
Page 27 of 112
 
COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION CU 67  Initiating Condition -- NOTIFICATION OF UNUSUAL EVENT Loss of required DC power
> for 15 minutes or longer
. Operating Mode Applicability:
Cold Shutdown (Mode 5)            Refueling (Mode 6)  Example Emergency Action Level (s):  Note: The SM Emergency D irector  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.  < 105 volts Less than (site specific bus voltage indication) on required Vital DC bus (site specific Vital DC busses)
> for 15 minutes or longer. Basis:  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.
] [Plants will routinely perform maintenance on a Train relat ed basis during shutdown periods The required busses are the minimum allowed by Technical Specifications for the mode of operation.]
It is intended that the loss of the operating (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 CA
: 34.  [(Site specific) 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 le ast 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. For a 60 cell string of batteries the cell voltage is typically 1.75 Volts per cell. For a 58 string battery set the minimum voltage is typically 1.81 Volts per cell.
] Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. to 0CAN121102
 
Page 28 of 112
 
COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION CU 78  Initiating Condition -- NOTIFICATION OF UNUSUAL EVENT Inadvertent criticality  Operating Mode Applicability:
Cold Shutdown (Mode 5)            Refueling (Mode 6)  Example Emergency Action Level (s):  1. UNPLANNED sustained positive period observed on nuc lear instrumentation. (BWR)
: 1. UNPLANNED sustained positive startup rate observed on nuclear instrumentation. (PWR)  Basis:  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 mis-loading events and inadvertent dilution events. This IC indicates a potential degradation of the level of safety of the plant, warranting a n N O UE classification. 
[This condition can be identified using the startup rate meter. The term "sustained" is used in order to allow exclusion of expected short term positive startup rates from planned fuel bundle or control rod movements during core alteration. These short term positive startup rates are the result of the rise in neutron population due to subcritical multiplication.]
Escalation would be by SMEmergency Director jJ udgment. to 0CAN121102
 
Page 29 of 112
 
COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION CA1  Initiating Condition - ALERT Loss of RCS/
reactor vessel/RPV inventory
. Operating Mode Applicability:
Cold Shutdown (Mode 5)          Refueling (Mode 6)    Example Emergency Action Level (s): (1 or 2)
Note: The SM/TSC Director/EOF DirectorEmergency 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/
reactor vesselRPV inventory as indicated by
: level.less than (site specific level)
Unit 1:  RVLMS Levels 1 through 8 indicate DRY Unit 2:  RVLMS Levels 1 through 5 indicate DRY OR Unit 1:  Reactor vessel level < 368 ft., 0 in. (bottom of the hot leg)
Unit 2:  Reactor vessel level < 369 ft., 1.5 in. (bottom of the hot leg)
[Low-Low ECCS actuation setpoint / Level 2 (BWR)
]  [Bottom ID of the RCS loop (PWR)
]  OR  2. RCS/reactor vesselRPV level cannot be monitored for
> 15 minutes or longer with a loss of RCS/reactor vesselRPV inventory as indicated by an unexplained level rise in the Reactor Building Sump, Reactor Drain Tank, Aux. Building Equipment Drain Tank, Aux. Building Sump, or Quench Tank.(site specific sump or tank). Basis:  These 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 reactor vesselRPV level loweringdecrease and potential core uncovery. This condition will result in a minimum emergency classification level of an Alert.
EAL #1  [The BWR Low
-Low ECCS Actuation Setpoint/Level 2 was chosen because it is a standard setpoint at which some available injection systems automatically start. The PWR Bottom ID of the RCS Loop Setpoint was chosen because at this level remote RCS level indication may be lost and loss of suction to decay heat removal systems has occurred. The Bottom ID of the RCS Loop    to 0CAN121102
 
Page 30 of 112
 
COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION CA1  Setpoint should be the level equal to the bottom of the RPV loop penetration (not the low point of the loop).
]  The bottom of the RCS hot leg penetration into the reactor vessel is approximately RLVMS Level 8 (Unit 1) or RVLMS Level 5 (Unit 2). Howe ver, RVLMS may not be available in mode 6.
Redundant means level indication is provided in this mode and included in EAL #1. The bottom of the RCS hot leg penetration into the reactor vessel is 368 ft., 0 in. (Unit 1) or 369 ft., 1.5 in. (Unit 2). Below this level, reactor vessel level indication will be lost and loss of suction to decay heat removal systems will occur.
The inability to restore and maintain level after reaching this setpoint would be indicative of a failure of the RCS barrier.
EAL #2  [In the cold shutdown mode, normal RCS level and reactor vessel level instrumentation systems will usually be available. In the refueling mode, normal means of reactor vessel level indication may not be available. Redundant means of reactor vessel level indication will usually 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 reactor vessel inventory loss was occurring by observing sump and tank level changes. Sump and tank level rises 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. Significant fuel damage is not expected to occur until the core has been uncovered for greater than 1 hour per the analysis referenced in the CG1 basis. Therefore this EAL meets the definition for an Alert.
]  If reactor vesselRPV level continues to lower then escalation to Site Area Emergency will be via CS1. to 0CAN121102
 
Page 31 of 112
 
COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION CA 53  Initiating Condition - ALERT Loss of all oO ffsite and all oO n sS ite AC power to Vital 4.16KVemergency busses > for 15 minutes or longer
. Operating Mode Applicability:
Cold Shutdown (Mode 5)            Refueling (Mode 6)            Defueled Example Emergency Action Level (s):  Note: The SM/TSC Director/EOF DirectorEmergency 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 oO ff sS ite and all oO n-sS ite AC pP ower to Vital 4.16KV busses (site specific emergency busses)
>for 15 minutes or longer. Basis:  Loss of all AC power compromises all plant sa fety systems requiring electric power including DHR/shutdown coolingRHR , emergency core coolingECCS , cC ontainment coolingHeat Removal , sS pent fF uel pool coolingHeat Removal and the uU ltimate hH eat sS ink. The event can be classified as an Alert when in cold shutdown, refueling, or defueled mode because of the significantly reduced decay heat and lower temperature and pressure, which allow raiincrea sing 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 iation Levels / Radiological Effluent (TAB A) ICs. Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.
[The companion IC is SS1]
. to 0CAN121102
 
Page 32 of 112
 
COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION CA 34  Initiating Condition - ALERT Inability to maintain plant in Cc old Ss hutdown  Operating Mode Applicability:
Cold Shutdown (Mode 5)            Refueling (Mode 6)  Example Emergency Action Level (s): (1 or 2) 
: 1. An UNPLANNED event results in RCS temperature
> 200&deg;Fgreater than (site specific Technical Specificatio n cold shutdown temperature limit
) > for greater than the specified duration io n Tt able C1. Table C1 RCS Reheat Duration Thresholds RCS  Containment Closure Duration  Intact (but not RCS Lowered Inventory
[PWR])  N/A  60 minutes*  Established  20 minutes*
Not intact or RCS Lowered Inventory (PWR)  Not Established  0 minutes
* If an RCS heat removal system is in operation within this time frame and RCS temperature is being reduced, the EAL is not applicable.
OR  Note: EAL #2 does not apply in solid plant conditions.
: 2. An UNPLANNED event results in RCS pressure riseincrease
> greater than 10 psi due to a loss of RCS cooling. (PWR-This EAL does not apply in Solid Plant conditions.)
Basis:  EAL #1  TFor EAL 1, t he RCS Reheat Duration Threshold table addresses complete loss of functions required for core cooling for greater than 60 minutes during refueling and cold shutdown modes when RCS integrity is established.
[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 CONTAINMENT CLOSURE in this condition is immaterial given that the RCS is providing a high pressure barrier to fission product release to t he environment.
] The 60 minute time frame should allow sufficient time to restore cooling without there being a substantial degradation in plant safety.
The RCS Reheat Duration Threshold table also addresses the complete loss of functions required for core cooling for greater than 20 minutes during refueling and cold shutdown modes
 
when CONTAINMENT CLOSURE is established but RCS integrity is not established or RCS    to 0CAN121102
 
Page 33 of 112
 
COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION CA 34  inventory is reduced
[(e.g., mid-loop operation in PWRs)].  [As discussed above, 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
.]    Finally, the EAL addresses complete loss of functions required for core cooling during refueling and cold shutdown modes when neither CONTAINMENT CLOSURE nor RCS integrity are established. Fin ally, complete loss of functions required for core cooling during refueling and cold shutdown modes when neither CONTAINMENT CLOSURE 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 because the evaporated reactor coolant that may be released into the Containment during this heatup condition could also be directly released to the environment.]  The note (*) indicates that this EAL is not applicable if actions are successful in restoring an RCS heat removal system to operation and RCS temperature is being reduced within the specified time frame. 
 
EAL #2  TIn EAL 2, t he 10 psi pressure riseincrease addresses situations where, due to high decay heat loads, the time provided to restore temperature control, should be less than 60 minutes. The RCS pressure setpoint chosen should be 10 psi or the lowest pressure that the site can read on
 
installed Control Board instrumentation that is equal to or greater than 10 psi.
Escalation to Site Area Emergency would be via CS1 should boiling result in significant reactor vesselRPV level loss leading to core uncovery.
[For PWRs, this IC and its associated EALs are based on concerns raised by Generic Letter 88
-17, "Loss of Decay Heat Removal." A number of phenomena such as pressurization, vortexing, steam generator U
-tube draining, RCS level differences when operating at a mid
-loop condition, decay heat removal system design, and level instrumentation problems can lead to conditions where decay heat removal is lost and core uncovery can occur. NRC analyses show that there are sequences that can cause core uncovery in 15 to 20 minutes and severe core damage w ithin an hour after decay heat removal is lost.
]  A loss of Technical Specification components alone is not intended to constitute an Alert. The same is true of a momentary UNPLANNED excursion above the Technical Specification cold
 
shutdown temperature limit when the heat removal function is available.
The SM/TSC Director/EOF Director Emergency Director must remain alert to events or conditions that lead to the conclusion that exceeding the EAL is IMMINENT. If, in the judgment of the SM/TSC Director/EOF DirectorEmergency Director , an IMMINENT situation is at hand, the classification should be made as if the threshold has been exceeded. to 0CAN121102
 
Page 34 of 112
 
COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION CS1  Initiating Condition - SITE AREA EMERGENCY Loss of RCS/
reactor vesselRPV inventory affecting core decay heat removal capability
. Operating Mode Applicability:
Cold Shutdown (Mode 5)            Refueling (Mode 6)  Example Emergency Action Level (s): (1 or 2 or 3)  Note:  The SM/TSC Director/EOF Director 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 not established:
RCS/RPV level less than (site specific level). Unit 1:  RVLMS Levels 1 through 9 indicate DRY Unit 2:  RVLMS Levels 1 through 6 indicate DRY
[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, core exit thermocouples indicate superheat.
R CS/RPV level less than (site specific level for TOAF).
OR  3. RCS/reactor vesselRPV level cannot be monitored for
> 30 minutes or longer with a loss of RCS/reactor vessel RPV inventory as indicated by any ANY of the following:  Containment High Range Radiation Monitor reading  > 10R/hr (Site specific radiation monitor) reading greater than (site specific value)
Erratic sS ource rR ange mM onitor iI ndication  Unexplained level rise in Reactor Building Sump, Reactor Drain Tank, Quench Tank, Aux. Building Equipment Drain Tank, or Aux. Building Sump.(site specific sump or tank)
Basis:  Under the conditions specified by this IC, continued loweringdecrease in RCS/reactor vesselRPV level is indicative of a loss of inventory control. Inventory loss may be due to an RCS breach, pressure boundary leakage, or continued boiling in the RPV. Thus, declaration of a Site Area
 
Emergency is warranted.
Escalation to a General Emergency is via CG1 or AG1. to 0CAN121102
 
Page 35 of 112
 
COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION CS1  EAL #1  [6" below the bottom ID of the RCS Loop should be the level equal to 6" below the bottom of the RPV loop penetration (not the low point of the loop). PWRs unable to measure this level should choose the first observable point below the bottom ID of the loop as the EAL value. If a water level instrument is not available such that the PWR EAL value cannot be determined, then EAL 3 should be used to determine if the IC has been met
.]  [Since BWRs have RCS penetrations below the EAL value, continued level decrease may be indicative of pressure boundary leakage.
]  EAL #3  [In the cold shutdown mode, normal RCS level and reactor vessel level instrumentation systems will usually be available. In the refueling mode, normal means of reactor vessel level indication may not be available. Redundant means of reactor vessel level indication will usually 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 reactor vessel inventory loss was occurring by observing sump and tank level changes. Sump and tank level rises 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 30-minute duration allows sufficient time for actions to be performed to recover inventory control equipment.
As water level in the reactor vesselRPV lowers, the dose rate above the core will riseincrease. The dose rate due to this core shine should result in site specific monitor indication and possible alarm.  [This EAL should conservatively estimate a site specific dose rate setpoint indicative of core uncovery (i.e., level at TOAF). For BWRs that do not have installed radiation monitors capable of indicating core uncovery, alternate site specific level indications of core uncovery should be used.]  [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.
]    to 0CAN121102
 
Page 36 of 112
 
COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION CG1 36 Initiating Condition - GENERAL EMERGENCY Loss of RCS/
reactor vesselRPV inventory affecting fuel clad integrity with containment challenged  Operating Mode Applicability:
Cold Shutdown (Mode 5)            Refueling (Mode 6)  Example Emergency Action Level (s): (1 or 2)  Note:  The SM/TSC Director/EOF DirectorEmergency 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. Core exit thermocouples indicate superheat RCS/RPV level less than (site specific level for TOAF) for > 30 minutes or longer
. AND b. Any of the following containment challenge indications:
ANY containment challenge indication (see Table):
CONTAINMENT CLOSURE not established  Explosive mixture inside containment  UNPLANNED rise in containment pressure OR  2. a. RCS/
reactor vesselRPV level cannot be monitored with core uncovery indicated by any ANY of the following for
> 30 minutes
: or longer Containment High Range Radiation Monitor reading > 10R/hr (Site specific radiation monitor) reading greater than (site specific setpoint).
Erratic source range monitor indication  UNPLANNED level rise in Reactor Building Sump, Reactor Drain Tank, Quench Tank, Aux. Building Equipment Drain Tank, or Aux. Building Sump (site specific sump or tank).      [Other site specific indications]
AND    to 0CAN121102
 
Page 37 of 112
 
COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION CG1 37 b. Any of the following containment challenge indications:
ANY containment challenge indication (see Table):
Table: Containment Challenge Indications
  ~CONTAINMENT CLOSURE not established.
~(Site specific explosive mixture) inside containment.
~UNPLANNED rise in containment pressure.
Se condary containment radiation monitor reading above (site specific value). [
BWR only]    CONTAINMENT CLOSURE not established  Explosive mixture inside containment  UNPLANNED rise in containment pressure Basis:  This IC represents the inability to restore and maintain reactor vesselRPV level to above the top of active fuel with containment challenged. Fuel damage is probable if reactor vesselRPV level cannot be restored, as available decay heat will cause boiling, further reducing the reactor vesselRP V level. With the 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.
  [These EALs are based on concerns raised by Generic Letter 88
-17, Loss of Decay Heat Removal, SECY 91
-283, Evaluation of Shutdown and Lo w 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.
]  A number of variables can have a significant impact on heat removal capability challenging the
 
fuel clad barrier. Examples include
: [BWRs] initial vessel level, shutdown heat removal system design  [PWRs] mid-loop, reduced level/flange level, head in place, cavity flooded, RCS venting strategy, decay heat removal system design, vortexing pre-disposition, and steam generator U-tube draining. Analysis indicates that core damage may occur within an hour following continued core uncovery
 
therefore, 30 minutes was conservatively chosen.
If CONTAINMENT CLOSURE is re-established prior to exceeding the 30 minute core uncovery time limit then escalation to GE would not occur. to 0CAN121102
 
Page 38 of 112
 
COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION CG1 38  [Site shutdown contingency plans typically provide for re
-establishing CONTAINMENT CLOSURE following a loss of heat removal or RCS inven tory functions.
] [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 Containment. However, 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
.]  [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 "maximum safe values" because these values are easily recognizable and have an emergency basis.
]  EAL #2  Sump and tank level risesincreases 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. 
[In the cold shutdown mode, normal RCS level and reactor vessel level instrumentation systems will usually be available. In the refueling mode, normal means of reactor vessel level indication may not be available. Redundant means of reactor vessel level indication will usually 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 reactor vessel inventory loss was occurring by observing sump and tank level changes. Sump and tank level rises 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
.]  As water level in the reactor vesselRPV lowers, the dose rate above the core will riseincrease. The dose rate due to this core shine should result in site specific monitor indication and possible alarm.  [This EAL should conservatively estimate a site specific dose rate setpoint indicative of core uncovery (ie., level at TOAF). For BWRs that do not have installed radiation monitors capable of indicating core uncovery, alternate site specific level indications of core uncovery should be used.]  [Post-TMI studies indicated that the installed nuclear instrument ation will operate erratically when the core is uncovered and that this should be used as a tool for making such determinations.]
Reference Documents
: 1. ULD-1-SYS-24, "Unit 1 Inadequate Core Cooling"
: 2. ULD-2-SYS-24, "Unit 2 Inadequate Core Cooling" to 0CAN121102
 
Page 39 of 112
 
ISFSI MALFUNCTION E-HU1 39 Initiating Condition - NOTIFICATION OF UNUSUAL EVENT Damage to a loaded cask CONFINEMENT BOUNDARY  Operating Mode Applicability:
All No t applicable Example Emergency Action Level (s):  1. Damage to a loaded cask CONFINEMENT BOUNDARY. Basis:  An NUEA NOUE in this IC is categorized on the basis of the occurrence of an event of sufficient magnitude that a loaded cask CONFINEMENT BOUNDARY is damaged or violated. This includes classification based on a loaded fuel storage cask CONFINEMENT BOUNDARY loss
 
leading to the degradation of the fuel during storage or posing an operational safety problem with
 
respect to its removal from storage.
  [The results of the ISFSI Safety Analysis Report (SAR) per NUREG 1536 or SAR referenced in the cask('s) Certificate of Complianc e and the related NRC Safety Evaluation Report identify natural phenomena events and accident conditions that could potentially effect the CONFINEMENT BOUNDARY
]This EAL addresses a dropped cask, a tipped over cask, EXPLOSION, PROJECTILE damage, FIRE damage or natural phenomena affecting a cask (e.g., seismic event, tornado, etc.). to 0CAN121102
 
Page 40 of 112
 
FISSION PRODUCT BARRIERS 40 General BasesNOTES The logic used for these initiating conditions reflects the following considerations:  The Fuel Clad Barrier and the RCS Barrier are weighted more heavily than the Containment Barrier (See Sections 3.4 and 3.8).
NUENOUE ICs associated with RCS and Fuel Clad Barriers are addressed under System Malfunction (S) ICs. At the Site Area Emergency level, there must be some ability to dynamically assess how far present conditions are from the threshold for a General Emergency. For example, if Fuel Clad
 
and RCS Barrier "Loss" EALs existed, that, in addition to off-site dose assessments, would
 
require continual assessments of radioactive inventory and containment integrity.
 
Alternatively, if both Fuel Clad and RCS Barrier "Potential Loss" EALs existed, the SM/TSC Director/EOF DirectorEmergency Director would have more assurance that there was no immediate need to escalate to a General Emergency.
The ability to escalate to higher emergency classesclassification levels as an event deteriorates must be maintained. For example, RCS leakage steadily increasing would represent an increasing risk to public health and safety.
The Containment Barrier should not be declared lost or potentially lost based on exceeding Technical Specification action statement criteria, unless there is an event in progress
 
requiring mitigation by the Containment barrier. When no event is in progress (Loss or
 
Potential Loss of either Fuel Clad and/or RCS) the Containment Barrier status is addressed
 
by Technical Specifications.
to 0CAN121102
 
Page 41 of 112
 
FISSION PRODUCT BARRIERS PWR TABLE 5
-F-3 FUEL CLAD 41 FUEL Fuel CLAD Clad BARRIER Barrier Emergency Action LevelsTHRESHOLDS
: (1 or 2 or 3 or 4 or 6 or 7 or 8)
FCB1 OR FCB2 OR FCB3 OR FCB4 OR FCB5 OR FCB6 The Fuel Clad barrier consists of the zircalloy or stainless steel fuel bundle tubes that contain the fuel pellets. 
: 1. Critical Safety Function Status
[These thresholds are for PWRs using Critical Safety Function Status Tree (CSFST) monitoring and functional restoration procedures. For more information, please refer to Section 3.9 of this document.]  Loss: Core Cooling Red Entry Conditions Met.
Potential Loss:
Core Cooling
-Orange Entry Conditions Met OR Heat Sink-Red Entry Conditions Met. Loss Thre shold A Core Cooling
- RED indic ates significant superheating and core uncovery and is considered to indicate loss of the Fuel Clad Barrier.
Potential Loss Threshold A Core Cooling
- ORANGE indicates subcooling has been lost and that some clad damage may occur. Potential Loss T hreshol d B Heat Sink
- RED when heat sink is required indicates the ultimate heat sink function is under extreme challenge.
: 12. Primary Coolant Activity Level (FCB1) Loss:    1. Coolant activity
>greater than (site specific value) 300 &#xb5;Ci/gm dose equivalent I-131 activity by Chemistry sample OR  2. Radiation levels > 1000 MR/hr Unit 1:  at SA-229 Unit 2:  at 2TCD-19 to 0CAN121102
 
Page 42 of 112
 
FISSION PRODUCT BARRIERS PWR TABLE 5
-F-3 FUEL CLAD 42 Potential Loss:
NoneNot Applicable Basis:  Loss  The site specific value corresponds to 300 Ci/gm I-131 equivalent. Assessment by the EAL Task Force indicates that this amount of coolant activity is well above that expected for iodine spikes
 
and corresponds to less than 5% fuel clad damage. This amount of radioactivity indicates significant clad damage and thus the Fuel Clad Barrier is considered lost. 
 
[The value can be expressed either in mR/hr observed on the sample or as Ci/gm results from analysis.]  A reading of greater than 1000 mR/hr within at one foot from the RCS sample lines (SA-229 for Unit 1, 2TCD-19 for Unit 2) has been determined to correspond to fuel clad failure of approximately 2-5%, and thus the fuel clad barrier is considered lost. This reading is well above that expected for iodine spikes and thus indicates significant clad damage and thus the fuel clad barrier is considered lost.
 
Potential Loss There is no Potential Loss EALthreshold associated with this item.
Reference Documents
: 1. ANO Calculation 03-E-0002-01, "Radiation Monitor EAL Setpoints for Fission Product Barrier Degradation"
: 23. Core Exit Thermocouple Readings (FCB2) Loss:  Core e xit t hermocouple s reading > 1200 &deg;F CET temperaturegreater than (site specific degree F). Potential Loss:
Core e xit t hermocouple s read ing greater than (site specific degree F)
. Unit 1: ICC exists as evidenced by CETs indicating superheated conditions Unit 2:  Average CETs indicate superheat for current RCS pressure Basis:  [Core Exit Thermocouple Readings are included in addition to the Critical Safety Functions to include conditions when the CSFs may not be in use (initiation after SI is blocked) or plants which do not have a CSF scheme.
]      to 0CAN121102
 
Page 43 of 112
 
FISSION PRODUCT BARRIERS PWR TABLE 5
-F-3 FUEL CLAD 43 Loss Th reshold A The Loss EAL of > 1200 &deg;F is consistent with NEI 99-01 and corresponds The s ite specific reading should correspond to significant superheating of the coolant. 
[This value typically corresponds to the temperature reading that indicates core cooling
- RED in Fuel Clad Barrier loss threshold 1.A which is usually about 1200 degrees F.]  Potential Loss Threshold A The site specific reading should correspond to loss of subcooling.
The Potential Loss EAL corresponds to a loss of subcooling margin.
Note that the loss or potential loss EAL for this category will occur after a loss of adequate sub-cooling margin, which represents a loss of the RCS barrier in EAL RCB1, and therefore represents the loss of two barriers, resulting in a Site Area Emergency per FS1. Any loss or potential loss of the containment barrier at that point would escalate to a General Emergency.
[This value typically corresponds to the temperature reading that indicates core cooling
- ORANGE in Fuel Clad Barrier potential loss threshold 1.A which is usually about 700 to 900 degrees F.
]  Reference Documents
: 1. Unit 1 EOP 1202.005, "Inadequate Core Cooling"
: 2. Unit 1 EOP 1202.013, "EOP Figures"
: 3. Unit 2 OP 2202.009, "Functional Recovery"
: 4. ANO Procedure OP 1302.022, "Core Damage Assessment"
: 5. CE-NPSD-241, "Development of the Comprehensive Procedure Guideline for Core Damage Assessment,"
Task 467
: 6. BWOG EOP Technical Bases Document, Vol. 3, Chapter III.F
: 34. Reactor Vessel Water Level (FCB3)  Loss: NoneNot Applicable Potential Loss:
RCS/RPV level less than (site specific level for TOAF).
Unit 1:  RVLMS Levels 1 through 9 indicate DRY Unit 2:  RVLMS Levels 1 through 7 indicate DRY Basis:  Loss  There is no Loss EALthreshold associated with this item. to 0CAN121102
 
Page 44 of 112
 
FISSION PRODUCT BARRIERS PWR TABLE 5
-F-3 FUEL CLAD 44  Potential Loss The site specific value for the Potential Loss threshold corresponds to the top of the active fuel.
The Reactor Vessel Level Monitoring Systems at AN O do not provide positive indication of core uncovery. The above core level indication provided is used to monitor the approach to and recovery from ICC conditions, but the CETs are used to identify core uncovery, and are the only positive indication of core uncovery.
Per reference document #1, the reactor vessel level indicators installed in Unit 1 extend from the top of the reactor vessel to the fuel alignment plate, and information in reference document #2 indicates that the lowest sensor is greater than 2 feet above the top of active fuel. If any of the 4 RCPs are running, flow induced turbulence produced by the pumps renders the reactor vessel level indicator readings invalid.
Per reference document #3, only the reactor vessel level indicators above the core are considered part of the ICC monitoring system. Per reference document #4, the lowest sensor above the core, RVLMS LVL 6 on the ICC monitoring panel 2C388, is 47 inches above the top of the core. If any of the 4 RCPs are running, flow induced turbulence produced by the pumps renders the reactor vessel level indicator readings invalid.
For either unit then, should CET indication be unavailable and reactor vessel level indication be unavailable due to RCP operation or any other cause, a degraded ability to monitor the barrier would exist.
[For sites using CSFSTs, the Potential Loss threshold is defined by the Core Cooling
- ORANGE path. The site specific value in this threshold should be consistent with the C SFST value.
] Reference Documents
: 1. ULD-1-SYS-24, "Unit 1 Inadequate Core Cooling System"
: 2. Calculation 84-EQ-0080-02, "Loop Error Analysis for Reactor Vessel Level Monitoring System" 3. ULD-2-SYS-24, "Unit 2 Inadequate Core Cooling Monitoring System"
: 4. Calculation 90-E-0116-01, "Unit 2 EOP Setpoint Document,"
Setpoint R.3
: 5. Not Applicable (included for numbering consistency between barrier tables)
: 46. Containment Radiation Monitoring (FCB4)  Loss:  Containment radiation monitor reading greater than (site specific value).
Containment high range radiation monitor reading > 1000 R/hr Potential Loss:
NoneNot Applicable Basis:    to 0CAN121102
 
Page 45 of 112
 
FISSION PRODUCT BARRIERS PWR TABLE 5
-F-3 FUEL CLAD 45  Loss  The 1000 R/hrsite specific reading on the containment high range radiation monitors (RE-8060 or RE-8061 for Unit 1, 2RE-8925-1 or 2RE-8925-2 for Unit 2) is a value which indicates the release of reactor coolant, with elevated activity indicative of fuel damage, into the containment.
  [The reading should be calculated assuming the instantaneous release and dispersal of the reactor coolant noble gas and iodine inventory associated with a concentration of 300 Ci/gm dose equivalent I
-131 into the containment atmosphere.]
Reactor coolant concentrations of this magnitude are several times larger than the maximum concentrations (including iodine spiking) allowed within technical specifications and are therefore
 
indicative of fuel damage.
This radiation monitor value is higher than that specified for RCS barrier Loss EAL RCB3 threshold #6. Thus, this EALthreshold indicates a loss of both the Fuel Clad barrier and RCS barrier that appropriately escalates the emergency classification level to a Site Area Emergency per FS1. Potential Loss
[Caution: it is important to recognize that in the event the radiation monitor is sensitive to shine from the reactor vessel or piping, spurious readings will be present and another indicator of fuel clad damage is necessary or compensated for in the threshold value.]
There is no Potential Loss EALthreshold associated with this item.
Reference Documents
: 1. NUREG 1228, "Source Term Estimation During Incident Response to Severe Nuclear Power Plant Accidents"
: 2. ANO Calculation 03-E-0002-01, "Radiation Monitor EAL Setpoints for Fission Product Barrier Degradation"
: 57. Other Site Specific Indications Core Damage Assessment (FCB5)
This subcategory addresses other site specific thresholds that may be included to indicate loss or potential loss of the Fuel Clad barrier
. Loss:  At least 5% fuel clad damage as determined from core damage assessment Potential Loss:
None  Basis:      to 0CAN121102
 
Page 46 of 112
 
FISSION PRODUCT BARRIERS PWR TABLE 5
-F-3 FUEL CLAD 46 Loss  This level is consistent with other fuel clad barrier loss EALs indicative of significant fuel clad damage, but uses core damage assessment eval uations by Technical Support personnel. The fuel clad barrier is considered lost.
If this determination is made from the high range containment radiation monitor readings, or if accompanied by other indications of a loss or potential loss of the RCS barrier, this EAL condition represents a Site Area Emergency per FS1. Potential Loss There is no potential loss EAL associated with this item.
Reference Documents
: 1. ANO Procedure OP-1302.022, "Core Damage Assessment"
: 68. Emergency Director Judgment (FCB6)  Loss:  Any condition in the opinion of the SM/TSC Director/EOF DirectorEmergency Director that indicates Loss or Potential Loss of the Fuel Clad bB arrier. Potential Loss:
Any condition in the opinion of the Emergency Director that indicates Potential Loss of the Fuel Clad Barrier.
Basis:  This EALThese thresholds address es any other factors that are to be used by the SM/TSC Director/EOF Director Emergency Director in determining whether the Fuel Clad barrier is lost or potentially lost. In addition, the inability to monitor the barrier should also be incorporated in this EALthreshold as a factor in SM/TSC Director/EOF Director Emergency Director judgment that the barrier may be considered lost or potentially lost. to 0CAN121102
 
Page 47 of 112
 
FISSION PRODUCT BARRIERS PWR TABLE 5
-F-3 RCS  47 RCS B arrierARRIER  EALsTHRESHOLDS
: (1 or 2 or 4 or 6 or 7 or 8)
RCB1 OR RCB2 OR RCB3 OR RCB4 The RCS Barrier includes the RCS primary side and its connections up to and including the pressurizer safety and relief valves, and other connections up to and including the primary
 
isolation valves.
: 1. Critical Safety Function Status
[These thresholds are for PWRs using Critical Safety Function Status Tree (CSFST) monitoring and functional restoration procedures. For more information, refer to Section 3.9 of this report.
]  Loss:  Not Applicable Potential Loss:
A. RCS Integrity
-Red Entry Conditions Met.
OR  B. Heat Sink
-Red Ent ry Conditions Met.
TBD OR Potential Loss Threshold A RCS Integrity
- RED indicates an extreme challenge to the safety function derived from appropriate instrument readings.
Potential Loss Threshold B Heat Sink
- RED when heat sink is required indicates th e ultimate heat sink function is under extreme challenge.
There is no Loss threshold associated with this item.
: 12. RCS Leak Rate (RCB1)  Loss:  RCS leak rate
> greater than available makeup capacity as indicated by
: a loss of RCS subcoolin g. Unit 1:  Loss of adequate subcooling margin Unit 2:  RCS subcooling (MTS) can NOT be maintained at least 30 &deg;F    to 0CAN121102
 
Page 48 of 112
 
FISSION PRODUCT BARRIERS PWR TABLE 5
-F-3 RCS  48  Potential Loss:
RCS leak rate indicated greater than (site specific capacity of one charging pump in the normal charging mode) with Letdown isolated
. Unit 1: UNISOLABLE RCS leak > 50 gpm with Letdown isolated Unit 2: UNISOLABLE RCS leak > 44 gpm with Letdown isolated Basis:  Loss  Threshold A This EALthreshold addresses conditions where leakage from the RCS is greater than available inventory control capacity such that a loss of subcooling has occurred. The loss of subcooling is the fundamental indication that the inventory control systems are inadequate in maintaining RCS
 
pressure and inventory against the mass loss through the leak. 
 
Potential Loss Threshold A This EALthreshold is based on the apparent inability to maintain normal liquid inventory within the Reactor Coolant System (RCS) by normal operation of the Makeup and Purification System (Unit 1) or the Chemical and Volume Control System (Unit 2).Chemical and Volume Control System which is considered to be the flow rate equivalent to one charging pump discharging to the charging header. For Unit 1 this is based on indications that leakage is greater than normal makeup capacity.
The operator could not batch in water and boric acid to the makeup system fast enough to maintain the makeup tank level during a 50 gpm RCS leak. It is not necessary to perform a detailed assessment of the RCS leakrate to implement this EAL. Any event or condition which, in the judgment of the SM/TSC Director/EOF Director, could result in RCS leakage in excess of Unit 1 normal makeup capacity would meet the intent of this EAL; for example:  Need to open the BWST suction for the operating makeup pump due to decreasing lowering makeup tank level  Full or partial HPI is needed to maintain the RCS pressure or pressurizer level  Two out of three seal stages failed on any RCP  RCS pressure decreasing lowering due to failure of a primary relief valve to reseat For Unit 2, this is considered as the capacity of one charging pump discharging to the charging header (44 gpm). Any event or condition which, in the judgment of the SM/TSC Director/EOF Director, could result in RCS leakage in excess of Unit 2 normal makeup capacity would meet the intent of this EAL; for example:
A second charging pump being required is indicative of a substantial RCS leak  Three out of four seal stages failed on any RCP  RCS pressure decreasing lowering due to failure of a primary relief valve to reseat    to 0CAN121102
 
Page 49 of 112
 
FISSION PRODUCT BARRIERS PWR TABLE 5
-F-3 RCS  49  Isolating letdown is a standard abnormal operating procedure action and may prevent unnecessary classifications when a non-RCS leakage path such as a Makeup and Purification System or CVCS leak exists. The intent of this condition is met if attempts to isolate Letdown are NOT successful. Additional charging pumps being required is indicative of a substantial RCS leak. 
[For plants with low capacity charging pumps, a 50 gpm indicated leak rate value may be used to indicate the Potential Loss.
] Reference Documents
: 1. Unit 1 EOP 1202.013, Figure 1, "Saturation and Adequate SCM"
: 2. Unit 1 EOP Setpoint Document, Calculation  90-E-0116-07, Setpoint B.19
: 3. Unit 2 EOP 2202.009, "Functional Recovery"
: 4. Unit 2 EOP Setpoint Document, Calculation 90-E-0116-01
: 5. Unit 2 SAR Table 9.3-14, Charging Pumps Design Data
: 3. Not Applicable (included for numbering consistency between barrier tables)
: 24. SG Tube Rupture (RCB2) Loss:  SGTR Ruptured SG that results in an ECCS (SI) actuation Potential Loss:
NoneNot Applicable Basis:  Loss  This EALthreshold addresses the full spectrum of Steam Generator (SG) tube rupture events in conjunction with Containment barrier Loss EALsthresholds. It addresses RUPTURED SG(s) for which the leakage is large enough to cause actuation (either automatic or manual) of ECCS (SI).
This is consistent to the RCS leak rate barrier Potential Loss EALthreshold
.  [For plants that have implemented Westinghouse Owners Group emergency response guides, this condition is described by "entry into E
-3 required by EOPs".
]
By itself, this EALthreshold will result in the declaration of an Alert. However, if the SG is also FAULTED (i.e., two barriers failed), the declaration escalates to a Site Area Emergency per Containment barrier Loss EAL CNB3thresholds. Potential Loss There is no Potential Loss EALthreshold associated with this item.
: 5. Not Applicable (included for numbering consistency between barrier tables) to 0CAN121102
 
Page 50 of 112
 
FISSION PRODUCT BARRIERS PWR TABLE 5
-F-3 RCS  50 36. Containment Radiation Monitoring (RCB3) Loss:  Containment radiation monitor reading greater than (site specific value). Containment high range rad monitor reading > 100 R/hr.
Potential Loss:
NoneNot Applicable Basis:  Loss  The 100 R/hrsite specific reading on the containment high range radiation monitors (RE-8060 or RE-8061 for Unit 1, 2RE-8925-1 or 2RE-8925-2 for Unit 2) is a value which indicates the release of reactor coolant to the containment.
 
[The reading should be calculated assuming the instantaneous release and dispersal of the reactor coolant noble gas and iodine inventory associated with normal opera ting concentrations (i.e., within T/S) into the containment atmosphere.]
This reading is will be less than that specified for Fuel Clad barrier EAL FCB4 threshold
: 6. Thus, this EALthreshold is would be indicative of a RCS leak only. If the radiation monitor reading roseincreased to that specified by Fuel Clad barrier EALthreshold , fuel damage would also be indicated.
  [However, if the site specific physical location of the containment radiation monitor is such that radiation from a cloud of released RCS gases could not be distinguished from radiation from adjacent piping and components containing elevated reactor coolant activity, this threshold should be omitted and other site specific indications of RCS leakage substituted.
]
During the initial fifteen minutes after a thermal event inside containment, the high range radiation monitor readings are considered invalid due to possibility of a transient thermally-induced current.
Potential Loss There is no Potential Loss EALthreshold associated with this item.
Reference Documents
: 1. ANO Calculation 03-E-0002-01 , "Radiation Monitor EAL Setpoints for Fission Product Barrier Degradation"
: 7. Other Site Specific Indications This subcategory addresses other site specific thresholds that may be included to indica te loss or potential loss of the RCS barrier.
: 48. Emergency Director Judgment (RCB4)    to 0CAN121102
 
Page 51 of 112
 
FISSION PRODUCT BARRIERS PWR TABLE 5
-F-3 RCS  51  Loss:  Any condition in the opinion of the SM/TSC Director/EOF DirectorEmergency Director that indicates Loss or Potential Loss of the RCS Barrier.
Potential Loss:
Any condition in the opinion of the Emergency Director that indicates Potential Loss of the Fuel Clad Barrier.
Basis:  This EALThese thresholds address es any other factors that are to be used by the SM/TSC Director/EOF Director Emergency Director in determining whether the RCS barrier is lost or potentially lost. In addition, the inability to monitor the barrier should also be incorporated in this EALthreshold as a factor in SM/TSC Director/EOF Director Emergency Director judgment that the barrier may be considered lost or potentially lost. to 0CAN121102
 
Page 52 of 112
 
FISSION PRODUCT BARRIERS PWR TABLE 5
-F-3 CONTAINMENT 52 C ontainmentONTAINMENT B arrierARRIER EALsTHRESHOLDS
: (1 or 2 or 3 or 4 or 5 or 6 or 7 or 8)  CNB1 OR CNB2 OR CNB3 OR CNB4 OR CNB5 OR CNB6 OR CNB7 The Containment Barrier includes the containment building and connections up to and including the outermost containment isolation valves. This barrier also includes the main steam, feedwater, and blowdown line extensions outside the containment building up to and including the outermost
 
secondary side isolation valve.
: 1. Critical Safety Function Status
  [These thresholds are for PWRs using Critical Safety Function Status Tree (CSFST) monitoring and functional restoration procedures. For more information, refer to Section 3.9 of this report.]
Loss:  Not Applicable Potential Loss:
A. Containment
-Red Entry Conditions Met.
RED path indicates an extreme challenge to the safety function derived from appropriate instrument readings and/or sampling results, and thus represents a potential loss of containment.
Conditions leading to a C ontainment RED path result from RCS barrier and/or Fuel Clad Barrier Loss. Thus, this threshold is primarily a discr iminator between Site Area Emergency and General Emergency representing a potential loss of the third barrier.
There is no Loss threshold associated with this item.
: 2. Containment Pressure (CNB)  Loss:    1.A. A containment pressure rise followed by a Rr apid unexplained drop in containment pressure following an initial rise in containment pressure OR  2.B. Containment pressure or sump level response not consistent with LOCA conditions Potential Loss:
1.A. Unit 1: Containment pressure
> 73.7 PSIA (59 PSIG)greater than (site specific value) and rising Unit 2: Containment pressure
> 73.7 PSIA (59 PSIG)greater than (site specific value
) and rising  OR    to 0CAN121102
 
Page 53 of 112
 
FISSION PRODUCT BARRIERS PWR TABLE 5
-F-3 CONTAINMENT 53  2.B. Explosive mixture exists inside containment.
OR  3.C. a. Containment Pressure> greater than containment spray  depressurization actuation setpoint UNIT 1:  44.7 PSIA (30 PSIG)
UNIT 2:  23.3 PSIA (8.6 PSIG)
AND  b. LESS THANL ess than one full train of spray depressurization equipment operating Basis: Loss Thresholds A and B Rapid unexplained loss of pressure (i.e., not attributable to containment spray or condensation effects) following an initial pressure riseincrease from a primary or secondary high energy line break indicates a loss of containment integrity. Containment pressure and sump levels should riseincrease as a result of mass and energy release into containment from a LOCA. Thus, sump level or pressure not risincr eas ing indicates containment bypass and a loss of containment integrity.
This indicator relies on operator recognition of an unexpected response for the condition and
 
therefore does not have a specific value associated with it. The unexpected response is important
 
because it is the indicator for a containment bypass condition.
Potential Loss
: 1. Threshold A The site specific pressure is based on the containment design pressure.
Potential Loss
: 2. Threshold B Existence of an explosive mixture means a hy drogen and oxygen concentration of at least the lower deflagration limit curve exists
. The hydrogen concentration of 4% has been recognized by the NRC staff as a well-established lower flammability limit in air or steam-air atmospheres that is adequately conservative for protecting against an H 2 explosion.
Hydrogen control systems at ANO are designed and operated as to maintain the containment hydrogen concentration below this level, so that indications of hydrogen concentrations above this are considered a potential challenge to the containment integrity.The indications of potential loss under this EAL corresponds to some of those leading to the RED path in potential loss threshold 1.A above and may be declared by those sites using CSFSTs.      to 0CAN121102
 
Page 54 of 112
 
FISSION PRODUCT BARRIERS PWR TABLE 5
-F-3 CONTAINMENT 54 Potential Loss
: 3. Threshold C This EALthreshold represents a potential loss of containment in that the containment heat removal/depressurization system (e.g., containment sprays, ice condenser fans, etc., but not including containment venting strategies) are ei ther lost or performing in a degraded manner, as indicated by containment pressure greater than the setpoint at which the equipment was
 
supposed to have actuated.
 
Reference Documents
: 1. Unit 1 OP-1105.003, "Engineering Safeguards Actuation System"
: 2. Unit 1 SAR Sections 1.4.43, 5.2.1.2.1,  14.2.2.5.5.1 (reactor building design pressure)
: 3. Unit 1 SAR Section 6.6 (Post-Loss of Coolant Accident Hydrogen Control)
: 4. Unit 1 TS Table 3.3.5-1
: 5. Unit 2 SAR Section 6.2.5 (Combustible Gas Control In Containment)
: 6. Unit 2 SAR Section 3.8.1.3.1.D (Containment Design Pressure)
: 7. Unit 2 TS Table 3.3-4
: 8. Regulatory Guide 1.7, "Control of Combustible Gas Concentrations in Containment Following a Loss-of-Coolant Accident, Rev. 2 1978"
: 23. Core Exit Thermocouple Readings (CNB2) Loss:  NoneNot App licable Potential Loss:
A.1. a. CETs Core Exit Thermocouples in excess of indicate > 1200 (site specific)
&#xba; F AND  b. Restoration procedures not effective within 15 minutes.
OR  B.2..a. a. CETsCore exit thermocouples indicate > 700 i n excess of (site
-specific) &#xba; F        AND  2.      b. b. RVLMS indicates:Reactor vessel level below (site specific level).
Unit 1:  Levels 1 through 9 DRY Unit 2:  Levels 1 through 7 DRY AND      c.      c. Restoration procedures not effective within 15 minutes.
Basis:    to 0CAN121102
 
Page 55 of 112
 
FISSION PRODUCT BARRIERS PWR TABLE 5
-F-3 CONTAINMENT 55  Loss [Core Exit Thermocouple Readings are included in addition to the Critical Safety Functions to include conditions when the CSFs may not be in use (initiation after SI is blocked) or plants which do not have a CSF scheme.
]  There is no Loss EALthreshold associated with this item.
Potential Loss The conditions in these EALthreshold s represent an IMMINENT core melt sequence which, if not corrected, could lead to vessel failure and a n higherincreased potential for containment failure. In conjunction with the Core Cooling and RCS Leakage criteria in the Fuel and RCS barrier columns, this threshold would result in the declaration of a General Emergency -- loss of two
 
barriers and the potential loss of a third. If the function restoration procedures are ineffective, there is no "success" path. 
 
The function restoration procedures are those emergency operating procedures that address the
 
recovery of the core cooling critical safety functions. The procedure is considered effective if the
 
temperature is droppingdecreasing or if the vessel water level is risingincreasing.  [For units using the CSF status trees, a direct correlation to those status trees can be made if the effectiveness of the restoration procedures is also evaluated as stated below.
] [Severe accident analyses (e.g., NUREG
-1150) have concluded that function restoration procedures can arrest core degradation within the reactor vessel in a significant fraction of the core damage scenarios, and that the likelihood of containment failure is very small in these events. Given this, it is appropriate to provide a reasonable period to allow function restoration procedures to arrest the core melt sequence.
] Whether or not the procedures will be effective should be apparent within 15 minutes. The SM/TSC Director/EOF DirectorEmergency Director should make the declaration as soon as it is determined that the procedures have been, or will be ineffective. 
 
Potential Loss Threshold B
[The reactor vessel level chosen should be consistent with the emergency response guides applicable to the facility.
]  34. SG Secondary Side Release With Primary
-to-Secondary Leakage (CNB3)  Loss:    to 0CAN121102
 
Page 56 of 112
 
FISSION PRODUCT BARRIERS PWR TABLE 5
-F-3 CONTAINMENT 56  Potential Loss:
NoneNot Applicable Basis:  Th ise loss EALthreshold recognizes that SG tube leakage can represent a bypass of the cC ontainment barrier as well as a loss of the RCS barrier.
Users should realize that the two loss threshold s could be considered redundant. This was recognized during the development process. The inclusion of an threshold that uses Emergency Procedure common ly used terms like "RUPTURED and FAULTED" adds to the ease of the classification process and has been included based on this human factor concern.
 
This EAL threshold results in a NUE NOUE for smaller breaks that; (1) do not exceed the Normal Makeup Capacity for Unit 1 or the capacity of one charging pump in the normal charging lineup for Unit 2 the normal charging capacity EAL thre shold in RCS leak rate barrier Potential Loss threshold , or (2) do not result in ECCS actuation in RCS SG tube rupture barrier Loss threshold. For larger breaks, RCS barrier threshold criteria would result in an Alert. For SG tube ruptures which may involve multiple steam generators or UNISOLABLEunisolable secondary line breaks, this condition threshold would exist in conjunction with RCS barrier conditionthreshold s and would result in a Site Area Emergency.
Escalation to General Emergency would be based on "Potential Loss" of the Fuel Clad Barrier.
Loss 1.Threshold A This EALthreshold addresses the condition in which a RUPTURED steam generator is also FAULTED. This condition represents a bypass of the RCS and containment barriers and is a subset of the second threshold.
In conjunction with RCS leak rate barrier loss EAL RCB2threshold , this would always result in the declaration of a Site Area Emergency.
Loss 2.Threshold B This EALthreshold addresses SG tube leaks that exceed 10 gpm in conjunction with an UNISOLABLE release path to the environment from the affected steam generator. The threshold for establishing the UNISOLABLE secondary side release is intended to be a prolonged release
 
of radioactivity from the RUPTURED steam generat or directly to the environment. This could be expected to occur when the main condenser is unavailable to accept the contaminated steam (i.e., SG tube rupture with concurrent loss of off-site power and the RUPTURED steam generator
 
is required for plant cooldown or a stuck open relief valve).
The time it takes to isolate a SG with A. RUPTURED SG is also FAULTED outside of containment
 
B.. 1a. Primary-to-sS econdary leakrate
> greater than 10 gpm  AND  2b. UNISOLABLE steam release from affected steam generatorSG to the environment    to 0CAN121102
 
Page 57 of 112
 
FISSION PRODUCT BARRIERS PWR TABLE 5
-F-3 CONTAINMENT 57 tube leakage > 10 gpm in accordance with plant specific EOPs is not considered a prolonged release. In this case the SG with tube leakage > 10 gpm with a concurrent loss of offsite power is normally steamed to the environment in a contro lled manner to achieve and maintain a RCS Hot Leg temperature below that which corresponds to the Main Steam Safety Valve relief settings.
However, if the SG cannot be isolated or if both SGs have tube leakage > 10 gpm, a prolonged release will likely be necessary to support plant cooldown.
If the main condenser is available, there may be releases via air ejectors, gland s eal exhausters, and other similar controlled, and often monitored, pathways. These pathways do not meet the intent of an UNISOLABLE release
 
path to the environment.
These minor releases are assessed using Abnormal Rad iation Levels /
Radiological Effluent ICs (TAB A). Potential Loss
[The leakage threshold for this threshold has been increased with Revision 3. In the earlier revision, the threshold was leakage greater than T/S allowable. Since the prior revision, many plants have implemented reduced steam generator T/S limits (e.g., 150 gpd) as a defense in depth associated with alternate steam generator plugging criteria. The 150 gpd threshold is deemed too low for use as an emergency threshold. A pressure boundary leakage of 10 gpm was used as the threshold in IC SU5, RCS Leakage, and is deemed appropriate for this threshold.]  There is no Potential Loss EAL associated with this item.
: 45. Containment Isolation Failure or Bypass (CNB4)  Loss:  A. 1a. A.Failure of a ll valves in any one line to close UNISOLABLE breach of containment . AND        2b. Direct downstream pathway to the envir onment exists after containment isolation signal  Potential Loss:
NoneNot Applicable Basis:  Loss This EALthreshold addresses incomplete containment isolation that allows a direct release to the environment.
A breach of containment has also occurred if an inboard and outboard pair of isolation valves fails to close on an automatic actuation signal or from a manual action in the Control Room and opens a release path to the environment.
to 0CAN121102
 
Page 58 of 112
 
FISSION PRODUCT BARRIERS PWR TABLE 5
-F-3 CONTAINMENT 58 The breach is not isolable from the Control Room if an attempt for isolation from the Control Room has been made and was unsuccessful. An attempt for isolation should be made prior to the accident classification. If isolable upon identification then this Initiating Condition is not applicable.
The use of the modifier "direct" in defining the release path discriminates against release paths
 
through interfacing liquid systems. The existence of an in-line charcoal filter does not make a
 
release path indirect since the filter is not effective at removing fission product noble gases. 
 
Typical filters have an efficiency of 95-99% removal of iodine. Given the magnitude of the core
 
inventory of iodine, significant releases could still occur. 
 
In addition, since the fission product release would be driven by boiling in the reactor vessel, the
 
high humidity in the release stream can be expected to render the filters ineffective in a short
 
period. Potential Loss There is no Potential Loss EALthreshold associated with this item. 
: 56. Containment Radiation Monitoring (CNB5)  Loss:  NoneNot Applicable Potential Loss:
A. Containment radiation monitor reading greater than (site specific value).
Containment high range rad monitor reading > 4000 R/hr Basis:  Loss  There is no Loss EALthreshold associated with this item.
Potential Loss The 4000 R/hrsite specific reading on the containment high range radiation monitors (RE-8060 or RE-8061 for Unit 1, 2RE-8925-1 or 2RE-8925-2 for Unit 2) is a value which indicates significant fuel damage well in excess of the EALthreshold s associated with both loss of Fuel Clad and loss of RCS barriers.
AAs stated in Section 3.8, a major release of radioactivity requiring off-site protective actions from core damage is not possible unless a major failure of fuel cladding allows radioactive material to be released from the core into the reactor coolant.
Regardless of whether containment is challenged, this amount of activity in containment, if released, could have such severe consequences that it is prudent to treat this as a potential loss
 
of containment, such that a General Emergency declaration is warranted.
 
to 0CAN121102
 
Page 59 of 112
 
FISSION PRODUCT BARRIERS PWR TABLE 5
-F-3 CONTAINMENT 59 [NUREG-1228, "Source Estimations During Incident Response to Severe Nuclear Power Plant Accidents," indicates that such conditions do not exist when the amount of clad damage is less than 20%. Unless there is a (site specific) analysis justifying a higher value, it is recommended that a radiation monitor reading corresponding to 20% fuel clad damage be specified here
.]  Because the monitor reading exceeds the readings for Fuel Clad Barrier loss in FCB4 and RCS Barrier loss in RCB3 , the SM/TSC Director/EOF Director should declare a General Emergency when this value on the Containment High Range Rad Monitor is exceeded as a loss of two barriers (fuel clad and RCS) and potential loss of the third (containment).
Reference Documents:
: 1. ANO Calculation 03-E-0002-01, "Radiation Monitor EAL Setpoints for Fission Product Barrier Degradation"
: 2. NUREG 1228, "Source Term Estimation During Incident Response to Severe Nuclear Power Plant Accidents"
: 67. Other Site Specific Indications (CNB6) Elevated readings on the following radiation monitors that indicate loss or potential loss of the Containment barrier:
MONITORS - UNIT 1 RX-9820 Containment Purge RX-9825 Radwaste Area RX-9830 Fuel Handling Area RX-9835 Emergency Penetration Room MONITORS - UNIT 2 2RX-9820 Containment Purge 2RX-9825 Radwaste Area 2RX-9830 Fuel Handling Area 2RX-9835 Emergency Penetration Room 2RX-9840 Post Accident Sampling Building 2RX-9845 Aux. Building Extension to 0CAN121102
 
Page 60 of 112
 
FISSION PRODUCT BARRIERS PWR TABLE 5
-F-3 CONTAINMENT 60 Basis:  This EAL covers other indications that may unambiguously indicate the loss or potential loss of the containment barrier.[
This EAL shoul d cover other (site
-specific) indications that may unambiguously indicate loss or potential loss of the containment barrier, including indications from area or ventilation monitors in containment annulus or other contiguous buildings. If site emergency operating procedures provide for venting of the containment during an emergency as a means of preventing catastrophic failure, a Loss EAL should be included for the containment barrier. This EAL should be declared as soon as such venting is IMMINENT. Containm ent venting as part of recovery actions is classified in accordance with the radiological effluent ICs.
]  78. Emergency Director Judgment (CNB7)  Loss:  Any condition in the opinion of the SM/TSC Director/EOF DirectorEmergency Director that indicates Loss or Potential Loss of the Containment Barrier.
Potential Loss:
Any condition in the opinion of the Emergency Director that indicates Potential Loss of the Fuel Clad Barrier.
Basis:  This EALThese thresholds address es any other factors that are to be used by the SM/TSC Director/EOF Director Emergency Director in determining whether the Containment barrier is lost or potentially lost. In addition, the inability to monitor the barrier should also be incorporated in this EAL threshold as a factor in SM/TSC Director/EOF Director Emergency Director judgment that the barrier may be considered lost or potentially lost.
 
The Containment barrier should not be declared lost or potentially lost based on exceeding
 
Technical Specification action statement criteria, unless there is an event in progress requiring
 
mitigation by the Containment barrier. When no event is in progress (Loss or Potential Loss of
 
either Fuel Clad and/or RCS) the Containment barrier status is addressed by Technical
 
Specifications. to 0CAN121102
 
Page 61 of 112
 
HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU 61  61 Initiating Condition - NOTIFICATION OF UNUSUAL EVENT Natural or destructive phenomena affecting the PROTECTED AREA Operating Mode Applicability:
All  Example Emergency Action Level: 
(1 or 2 or 3 or 4 or 5 or 6)  1. Seismic event identified by any AN Y 2 of the following:
* Seismic event confirmed by annunciation of the 0.01g acceleration alarm (site specific indication or method)
* Earthquake felt in plant
* National Earthquake Center OR  2. Tornado striking within PROTECTED AREA boundary or high winds
> greater than 67(site specific mph). OR 3. Internal flooding that has the potential to affect safety related equipment required by Technical Specifications for the current operating mode in any ANY of the structures or areas in Table H1.following areas
:  (site specific area list) to 0CAN121102
 
Page 62 of 112
 
HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU 61  62  Table H1 Unit 1 Unit 2 CA-1 & HP Office Area Condensate Demineralizer Room Corridor 98 Fire Area C Lower North Electrical Penetration Room (LNEPR) Lower South Electrical Equipment Room (LSEER)
/ Air Compressor Room Lower South Electrical Penetration Room (LSEPR) Lower South Piping Penetration Room (LSPPR)
Main Steam Isolation Violation (MSIV) Room North Engineered Safeguards (ES) SWGR Room (A4) South ES SWGR Room Turbine Building  A1, A2, H1, H2 SWGR area  354' Bowling Alley north end west of Breathing Air compressor room  368' West Heater Deck from LSEER (orange door) along east wall of ES SWGR Rooms to Corridor 98 door.
Upper North Electrical Penetration Room (UNEPR) / Hot Tool Room / Decon Room Upper South Electrical Penetration Room (USEPR) Upper South Piping Penetration Room (USPPR) 2A3 Room 2A4, 2D02, & East Battery Room 2B53 Room 2B63 Room 2B9/2B10 Room 2Y11/13 Equipment Room Auxiliary Building 317' General Access Auxiliary Building 335' Auxiliary Building 354'
'B' Engineered Safeguards Features (ESF)
Room Corridor Behind Door 340 Turbine Building  2A1, 2A2, 2H1, 2H2 Area  354' West wall of Demineralizer area  368' West Heater Deck north of north Switchgear (SWGR) Room (2A3) and East of LNEPR Intake Structure  354' or 366' LNEPR LSEPR Motor-Generator (MG) Set Room Steam Pipe Area Hot Machine Shop UNEPR, UNPPR, LNPPR, USPPR OR 4. Turbine failure resulting in casing penetration or damage to turbine or generator seals.
OR 5. (S ite specific occurrences affecting the PROTECTED AREA).
: 5. Lake Dardanelle level < 335 feet.
OR    to 0CAN121102
 
Page 63 of 112
 
HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU 61  63 6. Lake Dardanelle level > 345 feet.
Basis:  These EALs are categorized on the basis of the occurrence of an event of sufficient magnitude to be of concern to plant operators.
EAL #1  Damage may be caused to some portions of the site, but should not affect ability of safety functions to operate.
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 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. 
[For most plants with seismic instrumentation, the seismic switches are set at an acceleration of about 0.01g. This EAL should be developed on site specific basis. The method of detection can be based on instrumentation, validated by a reliable source, or operator assessment.
]  The National Earthquake Center can confirm if an earthquake has occurred in the area of the plant. EAL #2  This EAL is based on a tornado striking (touching down) or high winds within the PROTECTED AREA.  [The high wind value should be based on site specific FSAR design basis as long as it is within the range of the instrumentation available for wind speed.
]  The high wind value in EAL #2 is conservatively based on the SAR design basis for Unit 1 of 67 mph. Unit 2 Design basis is 80 mph.
Escalation of this emergency classification level, if appropriate, would be based on VISIBLE
 
DAMAGE, or by other in plant conditions, via HA
: 61. EAL #3  This EAL addresses the effect of internal flooding caused by events such as component failures, equipment misalignment, or outage activity mishaps. to 0CAN121102
 
Page 64 of 112
 
HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU 61  64 [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. The plant's IPEEE may provide insight into areas to be considered when developing this EAL.
]  Escalation of this emergency classification level, if appropriate, would be based VISIBLE DAMAGE via HA 61 , or by other plant conditions.
EAL #4  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 HU 42 and HU 53. This EAL is consistent with the definition of a n NUE NOUE while maintaining the anticipatory nature desired and recognizing the risk to non-safety related equipment.
Escalation of this emergency classification level, if appropriate, would be to HA 61 based on damage done by PROJECTILES generated by the failure or by the radiological releases for a BWR, or in conjunction with a steam generator tube rupture , for a PWR. These latter events would be classified by the radiological (A) ICs or Fission Product Barrier (F) ICs. EAL #5  This EAL addresses other site specific phenomena (such as hurricane, flood, or seiche) that can also be precursors of more serious events.
[S ites subject to severe weather as defined in the NUMARC station black out initiatives should include an EAL based on activation of the severe weather mitigation procedures (e.g., precautionary shutdowns, diesel testing, staff call
-outs, etc.).
]  EALs #5 and #6 EALs #5 and #6 are based on the levels of Lake Dardanelle at which the site will take specific action to reduce the impact of the lake level on plant safety by initiating plant shutdown.
Reference Documents:
: 1. OP-1203.025, "Natural Emergencies"
: 2. OP-2203.008, "Natural Emergencies"
: 3. Unit 1 FSAR 4. Unit 2 FSAR    to 0CAN121102
 
Page 65 of 112
 
HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU 42  65 Initiating Condition - NOTIFICATION OF UNUSUAL EVENT FIRE within the PROTECTED AREA not extinguished within 15 minutes of detection or EXPLOSION within the PROTECTED AREA  Operating Mode Applicability:
All  Example Emergency Action Level (s): (1 or 2)
Note: The SMEmergency 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 in any Table H1 structure or area not extinguished within
: 1) within 15 minutes of Cc ontrol Rr oom notification or
: 2) within 15 minutes of verification of a Cc ontrol Rr oom FIRE alarm in ANY of the following areas:
  (site specific area list) to 0CAN121102
 
Page 66 of 112
 
HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU 42  66  Table H1 Unit 1 Unit 2 CA-1 & HP Office Area Condensate Demineralizer Room Corridor 98 Fire Area C Lower North Electrical Penetration Room (LNEPR) Lower South Electrical Equipment Room (LSEER)
/ Air Compressor Room Lower South Electrical Penetration Room (LSEPR) Lower South Piping Penetration Room (LSPPR)
Main Steam Isolation Violation (MSIV) Room North Engineered Safeguards (ES) SWGR Room (A4) South ES SWGR Room Turbine Building  A1, A2, H1, H2 SWGR area  354' Bowling Alley north end west of Breathing Air compressor room  368' West Heater Deck from LSEER (orange door) along east wall of ES SWGR Rooms to Corridor 98 door.
Upper North Electrical Penetration Room (UNEPR) / Hot Tool Room / Decon Room Upper South Electrical Penetration Room (USEPR) Upper South Piping Penetration Room (USPPR) 2A3 Room 2A4, 2D02, & East Battery Room 2B53 Room 2B63 Room 2B9/2B10 Room 2Y11/13 Equipment Room Auxiliary Building 317' General Access Auxiliary Building 335' Auxiliary Building 354'
'B' Engineered Safeguards Features (ESF)
Room Corridor Behind Door 340 Turbine Building  2A1, 2A2, 2H1, 2H2 Area  354' West wall of Demineralizer area  368' West Heater Deck north of north Switchgear (SWGR) Room (2A3) and East of LNEPR Intake Structure  354' or 366' LNEPR LSEPR Motor-Generator (MG) Set Room Steam Pipe Area Hot Machine Shop UNEPR, UNPPR, LNPPR, USPPR OR 2. EXPLOSION within the PROTECTED AREA.
Basis:  This ICEAL 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. to 0CAN121102
 
Page 67 of 112
 
HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU 42  67 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 Cc ontrol Rr oom 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 15 minute duration is to size the FIRE and to discriminate against small FIRES that are readily extinguished (e.g., smoldering waste paper basket). 
[The site specific list should be limited and applies to buildings and areas in actual contact with or immediately adjacent to VITAL AREAS or other significant buildings or areas. The intent of this IC is not to include buildings (i.e., warehouses) or areas that are n ot in actual contact with or immediately adjacent to VITAL AREAS. This excludes FIRES within administration buildings, 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.
] 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 SMEmergency 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 HA
: 42. to 0CAN121102
 
Page 68 of 112
 
HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU 53  68 Initiating Condition - NOTIFICATION OF UNUSUAL EVENT Release of toxic, corrosive, asphyxiant, or flammable gases deemed detrimental to NORMAL PLANT OPERATIONS. Operating Mode Applicability:
All  Example Emergency Action Level (s): (1 or 2) 
: 1. Toxic, corrosive, asphyxiant or flammable gases in amounts that have or could adversely affect NORMAL PLANT OPERATIONS.
OR 2. Report by Ll ocal, Cc ounty or Ss tate officials for evacuation or sheltering of site personnel based on an offsite event.
Basis:  This ICEAL is based on the release of toxic, corrosive, asphyxiant or flammable gases of sufficient quantity to affect NORMAL PLANT OPERATIONS.
The fact that SCBA s may be worn does not eliminate the need to declare the event.
This IC 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 HA
: 53. to 0CAN121102
 
Page 69 of 112
 
HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU 14  69 Initiating Condition - NOTIFICATION OF UNUSUAL EVENT Confirmed SECURITY CONDITION or threat which indicates a potential degradation in the level of safety of the plant
. Operating Mode Applicability:
All  Example Emergency Action Level (s):  (1 or 2 or 3) 
: 1. A SECURITY CONDITION that does notNOT involve a HOSTILE ACTION as reported by the ANO Security Shift Supervision(site specific security shift supervision). OR  2. A credible site specific security threat notification.
OR 3. A validated notification from NRC providing information of an aircraft threat.
Basis:  NOTENote: Timely and accurate communication between Security Shift Supervision and the Control Room is crucial for the implem entation of effective Security EALs.
Security events which do not represent 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 HA 14 , HS 14 and HG1.
A higher initial classification could be made based upon the nature and timing of the security threat and potential consequences.
CThe licensee shall c onsider ation shall be given to upgrading the emergency response status and emergency classification level in accordance with the site's Safeguards Contingency Plan and Emergency Plan.
EAL #1  The Security Shift Supervisor isReference is made to site specific security shift supervision because these individuals are the designated individualpe rsonnel 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 EALthresho ld is based on the Safeguards Contingency Plansite specific security plans. The Safeguards Contingency Plan isSite specific Saf eguards Contingency Plans are based on guidance provided inby NEI 03-12. to 0CAN121102
 
Page 70 of 112
 
HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU 14  70 EAL #2  This EALthreshold is included to ensure that appropriate notifications for the security threat are made in a timely manner. This includes information of a credible threat. Only the plant to which the specific threat is made need declare the NUENotification of an UnusualEvent. 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 O ffsite R esponse O rganization s 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. Only the plant to which the specific threat is made need declare the NUEUnusual Event. 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 via HA1 emergency classification level would be via HA4 would be appropriate if the threat involves an airliner within 30 minutes of the plant. to 0CAN121102
 
Page 71 of 112
 
HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU 25  71 Initiating Condition - NOTIFICATION OF UNUSUAL EVENT Other conditions exist which in the judgment of the SMEmergency Director warrant declaration of a n NUENOUE. Operating Mode Applicability:
All  Example Emergency Action Level (s):  1. Other conditions exist which in the judgment of the SMEmergency Director indicate that events are in progress or have occurred which indicate a potential degradation of the level of safety of the plant or indicate a security threat to facility protection has been initiated. No
 
releases of radioactive material requiring offsite response or monitoring are expected unless
 
further degradation of safety systems occurs.
Basis:  This EAL addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the SMEmergency Director to fall under the NUENOUE emergency classification level. to 0CAN121102
 
Page 72 of 112
 
HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HA 61  72 Initiating Condition - ALERT Natural or destructive phenomena affecting VITAL AREAS  Operating Mode Applicability:
All  Example Emergency Action Level (s): (1 or 2 or 3 or 4 or 5 or 6) 
: 1. a. Seismic event
>greater than Operating Basis Earthquake (OBE) as indicated by annunciation of the 0.1g acceleration alarm(site specific seismic instrumentation) r eading (site specific OBE limit). AND b. Earthquake confirmed by any ANY of the following:  Earthquake felt in plant  National Earthquake Center  Control Room indication of degraded perform ance of systems required for the safe shutdown of the plant OR  2. Tornado striking or high winds
>greater than 67 mph(site specific mph) resulting in VISIBLE DAMAGE to any ANY of the following structures
/equipment containing safety systems or components or OR Cc ontrol Rr oom indication of degraded performance of those safety systems:  (site specific structure list)
Reactor Building  Intake Structure  Ultimate Heat Sink  BWST/RWT  Auxiliary Building  Turbine Building  QCST  Control Room  Startup Transformers  Diesel Fuel Vault OR      to 0CAN121102
 
Page 73 of 112
 
HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HA 61  73 3. Internal flooding in any ANY of the following areas resulting in an electrical shock hazard that precludes access to operate or monitor safety equipment or OR Cc ontrol Rr oom indication of degraded performance of those safety systems: 
(site specific area list)
Intake Structure  Ultimate Heat Sink  BWST/RWT  Auxiliary Building  Turbine Building  QCST  Control Room  Startup Transformers  Diesel Fuel Vault OR  4. Turbine failure-generated PROJECTILES resulting in VISIBLE DAMAGE to or  penetration of any ANY of the following structures
/equipment in Table H2 containing safety systems or components or OR Cc ontrol Rr oom indication of degraded performance of those safety systems:  (site specific structure list)
Table H2 Reactor Building Turbine Building Intake Structure QCST Ultimate Heat Sink Control Room BWST/RWT Startup Transformers Auxiliary Building Diesel Fuel Vault OR  5. Lake Dardanelle level < 335 feet and Emergency Cooling Pond inoperable.
OR  65. Vehicle crash resulting in VISIBLE DAMAGE to any ANY of the following structures
/equipment in Table H2 containing safety systems or components or OR Cc ontrol Rr oom indication of degraded performance of those safety systems: 
(site specific structure list) to 0CAN121102
 
Page 74 of 112
 
HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HA 61  74 Table H2 Reactor Building Turbine Building Intake Structure QCST Ultimate Heat Sink Control Room BWST/RWT Startup Transformers Auxiliary Building Diesel Fuel Vault
: 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)  Basis:  These EALs escalate from HU 61 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 system s in those structures evidenced by Cc ontrol Rr oom 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.
Escalation of this emergency classification level, if appropriate, would be based on System Malfunction (S) ICs. EALs #2 - #5  [These EALs should specify site specific structures or areas that contain safety system, or component and functions required for safe shutdown of the plant. Site specific Safe Shutdown Analysis should be consulted for equipment and plant areas required to establish or maintain safe shutdown.
]  EAL #1  Seismic events of this magnitude 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. 
[This threshold should be based on site specific FSAR design basis. See EPRI
-sponsored "Guidelines for Nuclear Plant Response to an Earthquake", dated October 1989, for information on seismic event categories.
]    to 0CAN121102
 
Page 75 of 112
 
HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HA 61  75 The National Earthquake Center can confirm if an earthquake has occurred in the area of the plant. EAL #2  This EAL is based on a tornado striking (touching down) or high winds that have caused VISIBLE DAMAGE to structures containing functions or systems required for safe shutdown of the plant.
The high wind value in EAL #2 is conservatively based on the SAR design basis for Unit 1 of 67 mph. Unit 2 Design basis is 80 mph. 
  [The high wind value should be based on site specific FSAR design basis as long as it is within the range of the instrumentation available for wind speed.
]  EAL #3  This EAL addresses the effect of internal flooding caused by events such as component failures, equipment misalignment, or outage activity mishaps. It is based on the degraded performance
 
of systems, or has created industrial safety hazards (e.g., electrical shock) that preclude
 
necessary access to operate or monitor safety equipment. The inability to access, operate or
 
monitor safety equipment represents an actual or substantial potential degradation of the level of
 
safety of the plant.
Flooding as used in this EAL describes a condition where water is entering the room faster than installed equipment is capable of removal, resulting in a rise of water level within the room.
 
Classification of this EAL should not be delayed while corrective actions are being taken to isolate
 
the water source. 
[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. The plant's IPEEE may provide insight into areas to be considered when developing this EAL.
]  EAL #4  This EAL addresses the threat to safety re lated 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. 
[The site specific list of areas should include all areas containing safety structure, system, or component, their controls, and their power supplies.
]  EAL #5  This EAL addresses vehicle crashes within the PROTECTED AREA that result s in VISIBLE DAMAGE to VITAL AREAS or indication of damage to safety structures, systems, or components containing functions and systems required for safe shutdown of the plant. to 0CAN121102
 
Page 76 of 112
 
HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HA 61  76 EAL #6  This EAL addresses other site specific phenomena tha t result in VISIBLE DAMAGE to VITAL AREAS or results in indication of damage to safety structures, systems, or components containing functions and systems required for safe shutdown of the plant (such as hurricane, flood, or seiche) that can also be precur sors of more serious events.
[S ites subject to severe weather as defined in the NUMARC station blackout initiatives should include an EAL based on activation of the severe weather mitigation procedures (e.g., precautionary shutdowns, diesel testing, staff call-outs, etc.).
] EAL #6 addresses site specific phenomena which has the potential for the loss of primary and secondary heat sink.
Reference Documents:
: 1. OP-1203.025, "Natural Emergencies"
: 2. OP-2203.008, "Natural Emergencies"
: 3. Unit 1 FSAR 4. Unit 2 FSAR to 0CAN121102
 
Page 77 of 112
 
HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HA 42  77 Initiating Condition - ALERT FIRE or EXPLOSION affecting the operability of plant safety systems required to establish or maintain safe shutdown  Operating Mode Applicability:
All  Example Emergency Action Level (s):  1. FIRE or EXPLOSION resulting in VISIBLE DAMAGE to any ANY Table H1 of the following structure s or area containing safety systems or components or OR Cc ontrol Rr oom indication of degraded performance of those safety systems
.  (site specific structure list)
Table H1 Unit 1 Unit 2 CA-1 & HP Office Area Condensate Demineralizer Room Corridor 98 Fire Area C Lower North Electrical Penetration Room (LNEPR) Lower South Electrical Equipment Room (LSEER)
/ Air Compressor Room Lower South Electrical Penetration Room (LSEPR) Lower South Piping Penetration Room (LSPPR)
Main Steam Isolation Violation (MSIV) Room North Engineered Safeguards (ES) SWGR Room (A4) South ES SWGR Room Turbine Building  A1, A2, H1, H2 SWGR area  354' Bowling Alley north end west of Breathing Air compressor room  368' West Heater Deck from LSEER (orange door) along east wall of ES SWGR Rooms to Corridor 98 door.
Upper North Electrical Penetration Room (UNEPR) / Hot Tool Room / Decon Room Upper South Electrical Penetration Room (USEPR) Upper South Piping Penetration Room (USPPR) 2A3 Room 2A4, 2D02, & East Battery Room 2B53 Room 2B63 Room 2B9/2B10 Room 2Y11/13 Equipment Room Auxiliary Building 317' General Access Auxiliary Building 335' Auxiliary Building 354'
'B' Engineered Safeguards Features (ESF)
Room Corridor Behind Door 340 Turbine Building  2A1, 2A2, 2H1, 2H2 Area  354' West wall of Demineralizer area  368' West Heater Deck north of north Switchgear (SWGR) Room (2A3) and East of LNEPR Intake Structure  354' or 366' LNEPR LSEPR Motor-Generator (MG) Set Room Steam Pipe Area Hot Machine Shop UNEPR, UNPPR, LNPPR, USPPR    to 0CAN121102
 
Page 78 of 112
 
HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HA 42  78 Basis:  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 or areas 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 SM/TSC Director/EOF DirectorEmergency Director with the resources needed to perform detailed damage assessments.
The SM/TSC Director/EOF DirectorEmergency Director also needs to consider any security aspects of the EXPLOSION. 
[This EAL should specify site specific structures or areas that contain safety system, or component and functions required for safe shutdown of the plant. Site specific Safe Shutdown Analysis should be consulted for equipment and plant areas required to establish or maintain safe shutdown.]  Escalation of this emergency classification level, if appropriate, will be based on System Malfunction (S)s , Fission Product Barrier Degradation (F) or Abnormal Rad iation Levels /
Radiological Effluent (A) ICs. to 0CAN121102
 
Page 79 of 112
 
HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HA 53  79 Initiating Condition - ALERT 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 Mode Applicability:
All  Example Emergency Action Level (s):  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.
Basis:  Gases in a VITAL AREA can affect the ability to safely operate or safely shutdown the reactor.
The fact that SCBA s may be worn does not eliminate the need to declare the event.
Declaration should not be delayed for confirmation from atmospheric testing if the atmosphere poses an immediate threat to life and health or an immediate threat of severe exposure to gases.
 
This could be based upon documented analysis, indication of personal ill effects from exposure, or operating experience with the hazards.
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 shut down beyond that already allowed by Technical Specifications at the time of the event.
An asphyxiant is a gas capable of reducing the level of oxygen in the body to dangerous levels.
Most commonly, asphyxiants work by merely displacing air in an enclosed environment. This
 
reduces the concentration of oxygen below the normal level of around 19%, which can lead to
 
breathing difficulties, unconsciousness or even death.
An uncontrolled release of flammable gasses within a facility structure has the potential to affect safe operation of the plant by limiting either operator or equipment operations due to the potential
 
for ignition and resulting equipment damage/personnel injury. Flammable gasses, such as
 
hydrogen and acetylene, are routinely used to maintain plant systems (hydrogen) or to repair
 
equipment/components (acetylene - used in welding). This EAL assumes concentrations of
 
flammable gasses which can ignite/support combustion.
Escalation of this emergency classification level, if appropriate, will be based on System Malfunction (S)s , Fission Product Barrier Degradation (F) or Abnormal Rad iation Levels /
Radioactive Effluent (A) ICs. to 0CAN121102
 
Page 80 of 112
 
HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HA 14  80 Initiating Condition - ALERT HOSTILE ACTION within the OWNER CONTROLLED AREA or airborne attack threat  Operating Mode Applicability:
All  Example Emergency Action Level (s):  (1 or 2)  1. A HOSTILE ACTION is occurring or has occurred within the OWNER CONTROL L ED AREA as reported by ANOthe  Security Shift Supervision(site specific security shift supervision). OR 2. A validated notification from NRC of an airliner attack threat within 30 minutes of the site.
Basis:  NOTENote: Timely and accurate communication between Security Shift Supervision and the Control Room is crucial for the implem entation 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, hunt ers, or physical disputes between employees within the O WNER C ONTROLLED A REA. 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 I ndependent S pent F uel S torage I nstallation' s that may be outside the PROTECTED AREA but still with in the OWNER CONTROLLED AREA. 
[Although nuclear plant security officers are well trained and prepared to protect against HOSTILE ACTION, it is appropriate for OROs to be notified and encouraged to begin activation (if they do not norm ally) 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    to 0CAN121102
 
Page 81 of 112
 
HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HA 14  81 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 O ffsite R esponse O rganization s 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. Only the plant to which the specific threat is
 
made need declare the Alert.
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. to 0CAN121102
 
Page 82 of 112
 
HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HA 35  82 Initiating Condition - ALERT Control room evacuation has been initiated  Operating Mode Applicability:
All  Example Emergency Action Level (s):  1.      Alternate Shutdown procedure(Site
-specific procedure) requires Cc ontrol Rr oom evacuation
:  Unit 1:
1203.002, "Alternate Shutdown" Unit 2:
2203.014, "Alternate Shutdown" Basis:  With the Cc ontrol Rr oom evacuated, additional support, monitoring and direction through the Technical Support Center and/or other emergency response facilities may be necessary.
Inability to establish plant control from outside the Cc ontrol Rr oom will escalate this event to a Site Area Emergency. to 0CAN121102
 
Page 83 of 112
 
HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HA 26  83 Initiating Condition - ALERT Other conditions exist which in the judgment of the SM/TSC Director/EOF DirectorEmergency Director warrant declaration of an Alert  Operating Mode Applicability:
All  Example Emergency Action Level (s):  1. Other conditions exist which in the judgment of the SM/TSC Director/EOF DirectorEmergency Director indicate that events are in progress or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant or a security event that involves probable life threatening risk to site personnel or damage to site equipment because
 
of HOSTILE ACTION. Any releases are expected to be limited to small fractions of the EPA
 
Protective Action Guideline exposure levels.
Basis:  This EAL addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the SM/TSC Director/EOF DirectorEmergency Di rector to fall under the Alert emergency classification level. to 0CAN121102
 
Page 84 of 112
 
HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HS 32  84 Initiating Condition - SITE AREA EMERGENCY Control Rr oom evacuation has been initiated and plant control cannot be established  Operating Mode Applicability:
All  Example Emergency Action Level (s):  1. a. Control room evacuation has been initiated  AND  b. Control of the plant cannot be established in accordance with the following procedures within 15(site specific minutes):  Unit 1:  1203.002, "Alternate Shutdown" Unit 2:  2203.014, "Alternate Shutdown" Basis:  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 tr ansfer of control of safety systems has not occurred (although fission product barrier damage may not yet be indicated).
The intent of the EAL is to establish control of important plant equipment and knowledge of important plant parameters in a timely m anner. Primary emphasis should be placed on those components and instruments that supply protection for and information about safety functions such as reactivity control (ability to shutdown the reactor and maintain it shutdown), RCS inventory (ability to cool the core), and decay heat removal (ability to maintain a heat sink).Typically, these safety fun ctions are reactivity control (ability to shutdown the reactor and maintain it shutdown), reactor water level (ability to cool the core), and decay heat removal (ability to maintain a heat sink) for a BWR. The equivalent functions for a PWR are reactivity control, RCS inventory, and secondary heat removal.
The determination of whether or not control is established at the remote shutdown panel is based on SM/TSC Director/EOF DirectorEmergency Director (ED) judgment. The SM/TSC Director/EOF DirectorEmerge ncy Director is expected to make a reasonable, informed judgment within 15 minutesthe site specific time for transfer that the plant stafflicensee has control of the plant from the remote shutdown panel. 
[The site specific time for transfer is based on analysis or assessments as to how quickly control must be reestablished without core uncovering and/or core damage. This time should not exceed 15 minutes without additional justification
.]  Escalation of this emergency classification level, if appropriate, would be by Fission Product Barrier Degradation (F) or Abnormal Rad iation Levels/Radiological Effluent (A) EALs. to 0CAN121102
 
Page 85 of 112
 
HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HS 23  85 Initiating Condition - SITE AREA EMERGENCY Other conditions exist which in the judgment of the SM/TSC Director/EOF DirectorEmergency Director warrant declaration of a Site Area Emergency  Operating Mode Applicability:
All  Example Emergency Action Level (s):  1. Other conditions exist which in the judgment of the SM/TSC Director/EOF DirectorEmergency 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.
Basis:  This EAL addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the SM/TSC Director/EOF DirectorEmergency Director to fall under the emergency classification level description for Site Area Emergency. to 0CAN121102
 
Page 86 of 112
 
HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HS 14  86 Initiating Condition - SITE AREA EMERGENCY HOSTILE ACTION within the PROTECTED AREA  Operating Mode Applicability:
All  Example Emergency Action Level (s):  1. A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by ANOthe  Security Shift Supervision(site security shift supervision). Basis:  This condition represents an escalated threat to plant safety above that contained in the Alert in that a HOSTILE FORCE has progressed from the OWNER CONTROLLED AREA to the
 
PROTECTED AREA.
This EAL addresses the contingency for a very rapid progression of events, such as that experienced on September 11, 2001. It is not prem ised 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 O ffsite R esponse O rganization readiness and preparation for the implementation of protective measures.
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, hunt ers, or physical disputes between employees within the PROTECTED AREA. Those events are adequately addressed by other EALs. 
[Although nuclear plant security officers are well trained and prepared to protect against HOSTILE ACTION, it is appropriate for OROs to be notified and encouraged to begin preparations for public protective actions (if they do not normally) to be bett er prepared should it be necessary to consider further actions.
]  [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.
]  Escalation of this emergency classification level, if appropriate, would be based on actual plant status after impact or progression of attack. to 0CAN121102
 
Page 87 of 112
 
HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HG1  87 Initiating Condition - GENERAL EMERGENCY HOSTILE ACTION resulting in loss of physical control of the facility  Operating Mode Applicability:
All  Example Emergency Action Level (s):  (1 or 2)  1. A HOSTILE ACTION has occurred such that plant personnel are unable to operate equipment required to maintain safety functions.
OR  2. A HOSTILE ACTION has caused failure of Spent Fuel Cooling Systems and IMMINENT fuel damage is likely for a freshly off-loaded reactor core in pool.
Basis:  EAL #1  This EAL encompasses conditions under which a HOSTILE ACTION has resulted in a loss of physical control of VITAL AREAS (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. These safety functions are reactivity control (ability to shut down the reactor and keep it shutdown) RCS inventory (ability to cool the core), and secondary heat removal (ability to maintain a heat sink). 
  [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 maint ain a heat sink) for a BWR. The equivalent functions for a PWR are reactivity control, RCS inventory, and secondary heat removal.
]  [Loss of physical control of the Control Room or remote shutdown/alternate shutdown capability alone may not prevent the ability to maintain safety functions per se. Design of the remote shutdown/alternate 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 threshold is not met.
EAL #2  This EAL addresses failure of spent fuel 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. At ANO, the term "freshly off-loaded reactor core" refers to fuel that has been discharged from the core and stored in the spent fuel pool for a period of LESS THAN one year. 
[A f reshly off
-loaded reactor core is defined by site specific criteria. to 0CAN121102
 
Page 88 of 112
 
HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HG2  88 Initiating Condition - GENERAL EMERGENCY Other conditions exist which in the judgment of the SM/TSC Director/EOF DirectorEmergency Director warrant declaration of a General Emergency  Operating Mode Applicability:
All  Example Emergency Action Level (s):  1. Other conditions exist which in the judgment of the SM/TSC Director/EOF DirectorEmergency Director indicate that events are in progress or have occurred which involve actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity or HOSTILE ACTION that results in an actual loss of physical control of the facility.
 
Releases can be reasonably expected to exc eed EPA Protective Action Guideline exposure levels offsite for more than the immediate site area.
Basis:  This EAL addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the SM/TSC Director/EOF DirectorEmergency Director to fall under the emergency classification level description for General Emergency. to 0CAN121102
 
Page 89 of 112
 
SYSTEM MALFUNCTION SU1  89 Initiating Condition - NOTIFICATION OF UNUSUAL EVENT Loss of all oO ffsite AC power to Vital 4.16 KVemergency busses > for 15 minutes or longer Operating Mode Applicability:
Power Operation s (Mode 1)
Startup (Mode 2)            Hot Standby (Mode 3)            Hot Shutdown (Mode 4)  Example Emergency Action Level (s):  Note: The SM 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 offsite AC power to Vital 4.16 KV (site specific emergency busses) for > 15 minutes or longe r. Basis:  Prolonged loss of offsite 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
 
to emergency busses.
Fifteen minutes was selected as a threshold to exclude transient or momentary losses of off-site power.        Reference Documents:
: 1. 1202.007, "
Degraded Powe r" 2. 1202.008, "
Blackout" 3. 2202.007, "
Loss of Off-Site Power"
: 4. 2202.008, "
Station Blackout"
  [At multi-unit stations, the EALs should allow credit for operation of installed design features, such as cross
-ties or swing diesels, provided that abnormal or emergency operating procedures address their use. However, these stations must also consider the impact of this condition on other shared safety functions in developing the site specific EAL.
]  [Plants that have a proceduralized capability to cross
-tie AC power from an off
-site power supply of a companion unit may take credit for the redundant power source in the associated EAL for this IC.]    to 0CAN121102
 
Page 90 of 112
 
SYSTEM MALFUNCTION SU 112  90 Initiating Condition - NOTIFICATION OF UNUSUAL EVENT Inability to reach required operating modeshutdown within Technical Specification limits Operating Mode Applicability:
Power Operation s (Mode 1)
Startup (Mode 2)            Hot Standby (Mode 3)            Hot Shutdown (Mode 4)  Example Emergency Action Level (s):  1. Plant is not brought to required operating mode within Technical Specifications LCO Action Statement tT ime. Basis:  Limiting Conditions of Operation (LCOs) require the plant to be brought to a required operating mode when the Technical Specification required configuration cannot be restored. Depending on
 
the circumstances, this may or may not be an emergency or precursor to a more severe
 
condition. In any case, the initiation of plant shutdown required by the site Technical
 
Specifications requires a four hour report under 10 CFR 50.72 (b) Non-emergency events. The
 
plant is within its safety envelope when being shut down within the allowable action statement
 
time in the Technical Specifications. An immediate N O UE is required when the plant is not brought to the required operating mode within the allowable action statement time in the Technical Specifications. Declaration of a n  N O UE is based on the time at which the LCO-specified action statement time period elapses under the site Technical Specifications and is not related to how long a condition may have existed.  [Other required Technical Specification shutdowns that involve precursors to more serious events are addressed by other System Malfunction, Hazards, or Fission Product Barrier Degradation ICs.]  Reference Documents:
: 1. ANO2 Technical Specifications 2. ANO1 Technical Specifications    to 0CAN121102
 
Page 91 of 112
 
SYSTEM MALFUNCTION SU 63  91 Initiating Condition - NOTIFICATION OF UNUSUAL EVENT UNPLANNED loss of safety system annunciation or indication in the Cc ontrol Rr oom > for 15 minutes or longer Operating Mode Applicability:
Power Operation s (Mode 1)
Startup (Mode 2)            Hot Standby (Mode 3)            Hot Shutdown (Mode 4)  Example Emergency Action Level (s):  Note: The SMEmergency 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:  (Site specific control room safety system annunciation)
: a. Control Room annunciators associated with safety systems.
OR  b. C(Site specific c ontrol Rr oom safety system indication
.)  Basis:  This IC and its associated EAL are intended to recognize the difficulty associated with monitoring changing plant conditions without the use of a major portion of the annunciation or indication
 
equipment.
Recognition of the availability of computer based indication equipment is considered
[e.g., SPDS, plant computer, etc.
]  "Planned" loss of annunciators or indicators includes scheduled maintenance and testing activities.
Quantification is arbitrary, however, it is estimat ed that if approximately 75% of the safety system annunciators or indicators are lost, there is an increased risk that a degraded plant condition
 
could go undetected. It is not intended that plant personnel perform a detailed count of the
 
instrumentation lost but use the value as a judgment threshold for determining the severity of the
 
plant conditions. to 0CAN121102
 
Page 92 of 112
 
SYSTEM MALFUNCTION SU 63  92 It is further recognized that most plant designs provide redundant safety system indication powered from separate uninterruptible power supplies. While failure of a large portion of
 
annunciators is more likely than a failure of a large portion of indications, the concern is included
 
in this EAL due to difficulty associated with assessment of plant conditions. The loss of specific, or several, safety system indicators should re main a function of that specific system or component operability status. This will be addressed by the specific Technical Specification. The
 
initiation of a Technical Specification imposed plant shutdown related to the instrument loss will
 
be reported via 10 CFR 50.72. If the shutdown is not in compliance with the Technical
 
Specification action, the N O UE is based on SU 112 "Inability to rR each rR equired operating mode Shutdown wW ithin Technical Specification lL imits."  Indicators associated with safety systems are those indicators for reactivity control, core cooling, maintaining reactor coolant system integrity or maintaining containment integrity.
  [Site specific annunciators or indicators for this EAL must include those identified in the Abnormal Operating Procedures, in the Emergency Operating Procedures, and in other EALs (e.g., area, process, and/or effluent rad monitors, etc.).
]  Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. 
[Due to the limited number of safety systems in operation during cold shutdown, refueling, and defueled modes, no IC is indicated during these modes of operation.
]  This N O UE will be escalated to an Alert based on a concurrent loss of compensatory indications or if a SIGNIFICANT TRANSIENT is in progress during the loss of annunciation or indication (SA6). Reference Documents:
: 1. 1203.043, "Loss Control Room Annunciators"
: 2. 2203.042, "Loss of Control Room Annunciators" to 0CAN121102
 
Page 93 of 112
 
SYSTEM MALFUNCTION SU9  93 Initiating Condition - NOTIFICATION OF UNUSUAL EVENT Fuel cC lad degradation Operating Mode Applicability:
Power Operation s (Mode 1)
Startup (Mode 2)            Hot Standby (Mode 3)            Hot Shutdown (Mode 4)  Example Emergency Action Level (s):  (1 or 2) 
: 1. (Site specific radiation monitor readings indicating fuel clad degradation greater than Technical Specification allowable limits.)
: 2. (Site specific coolant sample activity value indicating fuel clad deg radation greater than Technical Specification allowable limits.)
: 1. Failed Fuel Iodine radiation monitor reading indicates fuel clad degradation > Technical Specification allowable limits:
Unit 1: RI-1237S reads > 1.3 x 10 5 counts per minute Unit 2:  2RITS-4806B reads > .65 x 10 5 counts per minute OR 2. RCS sample activity value indicating fuel clad degradation > Technical Specification allowable limits:
  > 1.0 uCi/gm Dose Equivalent I-131 for more than 48 hours OR  Unit 1: > 60 uCi/gm Dose Equivalent I-131 Unit 2: > 60 uCi/gm Dose Equivalent I-131 OR  Unit 1: > 2200 &#xb5;Ci/gm Dose Equivalent Xe-133 for more than 48 hours Unit 2: > 3100 &#xb5;Ci/gm Dose Equivalent Xe-133 for more than 48 hours    to 0CAN121102
 
Page 94 of 112
 
SYSTEM MALFUNCTION SU9  94 Basis:  This ICEAL is included because it is a precursor of more serious conditions and, as result, is considered to be a potential degradation of the level of safety of the plant.
EAL #1  This threshold addresses the Letdown site
-specific Rr adiation Mm onitor readings that provide indication of a degradation of fuel clad integrity. 
[Such as BWR air ejector monitors, PWR failed fuel monitors, etc.
]  EAL #2  This EALthreshold addresses coolant samples exceeding coolant technical specifications for transient iodine spiking limits and coolant samples exceeding coolant Technical Specifications for nominal operating limits for the time period specified in the Technical Specifications
. Escalation of this ICEAL to the Alert level is via the Fission Product Barriers (F). Reference Documents:
: 1. ANO1 Technical Specifications 2. ANO2 Technical Specifications  to 0CAN121102
 
Page 95 of 112
 
SYSTEM MALFUNCTION SU 75  95 Initiating Condition - NOTIFICATION OF UNUSUAL EVENT RCS leakage  Operating Mode Applicability:
Power Operation s (Mode 1)
Startup (Mode 2)            Hot Standby (Mode 3)            Hot Shutdown (Mode 4)  Example Emergency Action Level (s):  (1 or 2) 
: 1. Unidentified or pressure boundary leakage
>greater than 10 gpm.      OR  2. Identified leakage
> greater than 25 gpm. Basis:  With respect to this IC, RCS leakage is defined as a loss of RCS inventory due to a leak in the RCS or a supporting system that is not or cannot be isolated within 10 minutes. For example, isolation of the RCS Letdown (purification) system is a standard abnormal operating procedure action and may prevent unnecessary classifications when a non-RCS leakage path leak exists.
However, the intent of this condition is met if attempts to isolate the RCS leak are NOT successful.
This IC is included as a n N O UE 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. The 10 gpm value for the unidentified or pressure boundary leakage was selected as it is observable with
 
normal Cc ontrol Rr oom indications. Lesser values must generally be determined through time-consuming surveillance tests (e.g., mass balances).
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 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.
Steam generator tube leakage is identified leakage.
In either case, escalation of this IC to the Alert level is via Fission Product Barrier Degradation (F) ICs. to 0CAN121102
 
Page 96 of 112
 
SYSTEM MALFUNCTION SU 86  96 Initiating Condition - NOTIFICATION OF UNUSUAL EVENT Loss of all oO nsite or oO ffsite communications capabilities Operating Mode Applicability:
Power Operation s (Mode 1)
Startup (Mode 2)            Hot Standby (Mode 3)            Hot Shutdown (Mode 4)  Example Emergency Action Level (s):  (1 or 2) 
: 1. Loss of all Table M1of the following onsite communication s methods affecting the ability to perform routine operations.  (site specific list of communications methods)
OR  2. Loss of all Table M2of the following offsite communication s methods affecting the ability to perform offsite notifications. (site specific list of communications methods)
Table M1 Onsite Communications Methods  Table M2 Offsite Communications Methods Station radio system All telephone lines (commercial and microwave)
Plant paging system ENS In-plant telephones Gaitronics Basis:  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 issues with offsite authorities. 
[The loss of off
-site communications ability is expected to be significantly more comprehensive than the condition addressed by 10 CFR 50.72.
]  The availability of one method of ordinary offsite communications is sufficient to inform federal, state, and local authorities of plant problems. This EAL is intended to be used only when
 
extraordinary means (e.g., relaying of information from non-routine radio transmissions, individuals being sent to off-site locations, etc.) are being used to make communications possible. to 0CAN121102
 
Page 97 of 112
 
SYSTEM MALFUNCTION SU 86  97 Reference Documents:
: 1. 1903.062, "Communications System Operating Procedure"
  [Site specific list for on
-site communications loss must encompass the loss of all means of communications (e.g., commercial telephones, sound powered phone systems, page party system (Gaitronics) and radios / walkie talkies) routinely used for operations.
]  [Site specific list for off
-site communications loss must encompass the loss of all means of communications with off
-site authorities. This should include the ENS, commercial telep hone lines, telecopy transmissions, and dedicated phone systems that are routinely used for offsite emergency notifications.
]    to 0CAN121102
 
Page 98 of 112
 
SYSTEM MALFUNCTION SU 108  98 Initiating Condition - NOTIFICATION OF UNUSUAL EVENT Inadvertent criticality Operating Mode Applicability:
Hot Standby (Mode 3)            Hot Shutdown (Mode 4)
Example Emergency Action Level (s):  1. UNPLANNED sustained positive period observed on nuclear instrumentation. [
BWR]  1. UNPLANNED sustained positive startup rate observed on nuclear instrumentation.
[PWR]  Basis:  This IC addresses inadvertent criticality events. This IC indicates a potential degradation of the level of safety of the plant, warranting a n N O UE classification. This IC excludes inadvertent criticalities that occur during planned reactivity changes associated with reactor startups (e.g., criticality earlier than estimated). 
[This condition can be identified using the startup rate meter. The term "sustained" is used in order to allow exclusion of expected short term positive startup rates from planned control rod movements for (such as shutdown bank withdrawal). These short term positive startup rates are the result of the rise in neutron population due to subcritical multiplication
.]  Escalation would be by the Fission Product Barrier Table (F), as appropriate to the operating mode at the time of the event.
Reference Documents:
: 1. 1203.012G, "Annunciator K08 Corrective Action"
: 2. 2203.012D, "Annunciator 2K04 Corrective Action" to 0CAN121102
 
Page 99 of 112
 
SYSTEM MALFUNCTION SA 32  99 Initiating Condition - ALERT Automatic tScram (T rip) fails to shutdown the reactor and the manual actions taken from the reactor control console are successful in shutting down the reactor  Operating Mode Applicability:
Power Operation s (Mode 1)
Startup (Mode 2)  Example Emergency Action Level (s):  1. a. An automatic scram (trip) failed to shutdown the reactor as indicated by reactor power >
5%. AND  b. Manual actions taken at the reactor control console successfully shutdown the reactor as indicated by reactor power < 5%(site specific indications of plant shutdo wn). Basis:  [The reactor should be considered shutdown when it producing less heat than the maximum decay heat load for which the safety systems are designed (typically 3 to 5% power). For plants using CSFSTs, this EAL equates to the criteria used to determine a valid Subcriticality Red Path.
For BWRs this EAL should be the APRM downscale trip setpoint.
]  Manual scram (trip) actions taken at the reactor control console are any set of actions by the Rr eactor Oo perator(s) which causes or should cause control rods to be rapidly inserted into the core and shuts down the reactor.
Any action taken to trip the reactor from any location other than panel C03 (Unit 1) or panels 2C03/2C14 (Unit 2) constitutes a failure of the manual trip function.
Failure of manual trip would escalate the event to a Site Area Emergency (SS3).  [If the manual scram (trip) switches/pushbuttons on the control room console panels are considered an automatic input into the Reactor Protection System, a failure to scram (trip) without any other automatic input would make this threshold applicable.
]
This condition indicates failure of the automatic protection system to scram (trip) the reactor. This condition is more than a potential degradation of a safety system in that a front line automatic protection system did not function in response to a plant transient. Thus the plant safety has been
 
compromised because design limits of the fuel may have been exceeded. An Alert is indicated
 
because conditions may exist that lead to potential loss of fuel clad or RCS and because of the
 
failure of the Reactor Protection System to automatically shutdown the plant.
If manual actions taken at the reactor control console fail to shutdown the reactor, the event
 
would escalate to a Site Area Emergency. to 0CAN121102
 
Page 100 of 112
 
SYSTEM MALFUNCTION SA 64  100 Initiating Condition - ALERT UNPLANNED lL oss of safety system annunciation or indication in the Cc ontrol Rr oom with eitherEITHER (1) a SIGNIFICANT TRANSIENT in progress, or (2) compensatory indicators unavailable  Operating Mode Applicability:
Power Operation s (Mode 1)
Startup (Mode 2)            Hot Standby (Mode 3)            Hot Shutdown (Mode 4)  Example Emergency Action Level (s):  Note: The SM/TSC Director/EOF DirectorEmergency 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. UNPLANNED loss of
>greater than approximately 75% of the following
> for 15 minutes or longer:  (Site specific control room safety system annunciation)
Control Room annunciators associated with safety systems OR    C(Site specific c ontrol Rr oom safety system indication
)  AND    b. Either EITHER of the following:  A SIGNIFICANT TRANSIENT is in progress OR. Compensatory indications are unavailable.
Basis:  This IC is intended to recognize the difficulty associated with monitoring changing plant conditions without the use of a major portion of the annunciation or indication equipment during a
 
SIGNIFICANT TRANSIENT. 
[Recognition of the availability of computer based indication equipment is considered (e.g., SPDS, plant computer, etc.).
]  "Planned" loss of annunciators or indicators includes scheduled maintenance and testing activities.
to 0CAN121102
 
Page 101 of 112
 
SYSTEM MALFUNCTION SA 64  101 Quantification is arbitrary, however, it is estimat ed that if approximately 75% of the safety system annunciators or indicators are lost, there is an increased risk that a degraded plant condition
 
could go undetected. It is not intended that plant personnel perform a detailed count of the
 
instrumentation lost but use the value as a judgment threshold for determining the severity of the
 
plant conditions. It is also not intended that the Shift ManagerSupervisor be tasked with making a judgment decision as to whether additional personnel are required to provide increased monitoring of system operation.
It is further recognized that most plant designs provide redundant safety system indication powered from separate uninterruptible power supplies. While failure of a large portion of
 
annunciators is more likely than a failure of a large portion of indications, the concern is included
 
in this EAL due to difficulty associated with assessment of plant conditions. The loss of specific, or several, safety system indicators should re main a function of that specific system or component operability status. This will be addressed by the specific Technical Specification. The
 
initiation of a Technical Specification imposed plant shutdown related to the instrument loss will
 
be reported via 10 CFR 50.72. If the shutdown is not in compliance with the Technical
 
Specification action, the N O UE is based on SU 112 "Inability to rR each rR equired operating modeShutdown wW ithin Technical Specification lL imits."  [Site-specific annunciators or indicators for this EAL must include those identified in the Abnormal Operating Procedures, in the Emergency Operating Procedures, and in other EALs (e.g., area, process, and/or effluent rad monitors, etc.).]
Indicators associated with safety systems are those indicators for reactivity control, core cooling, maintaining reactor coolant system integrity or maintaining containment integrity.
"Compensatory indications" in this context includes computer based information such as SPDS , QSPDS, COLSS, etc. [This should include all computer systems available for this use depending on specific plant design and subsequent retrofits.
] If both a major portion of the annunciation system and all computer monitoring are unavailable, the Alert is required. 
[Due to the limited number of safety systems in operation during cold shutdown, refueling and defueled modes, no IC is indicated during these modes of operation.
]  Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. 
 
This Alert will be escalated to a Site Area Emergency if the operating crew cannot monitor the transient in progress due to a concurrent loss of compensatory indications with a SIGNIFICANT
 
TRANSIENT in progress during the loss of annunciation or indication.
Reference Documents:
: 1. 1015.037, "
Post Transient Review"
: 2. 1203.043, "Loss of Control Room Annunciators"
: 3. 2203.042, "Loss of Control Room Annunciators" to 0CAN121102
 
Page 102 of 112
 
SYSTEM MALFUNCTION SA 15  102 Initiating Condition - ALERT AC power capability to Vital 4.16 KVemergency busses reduced to a single power source
> for 15 minutes or longer such that any additional single failure would result in station blackout  Operating Mode Applicability:
Power Operation s (Mode 1)
Startup (Mode 2)            Hot Standby (Mode 3)            Hot Shutdown (Mode 4)  Example Emergency Action Level (s):  Note: The SM/TSC Director/EOF DirectorEmergency 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 Vital 4.16 KV(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.
Basis:  [This IC and the associated EALs are intended to provide an escalation from IC SU1, "Loss of All Off-site AC Power To Emergency Busses for Greater Than 15 Minutes."]  The condition indicated by this IC is the degradat ion of the offsite and onsite AC power systems such that any additional single failure would result in a station blackout. This condition could
 
occur due to a loss of offsite power with a concurrent failure of all but one emergency generator
 
to supply power to its emergency busses. Another related condition could be the loss of all offsite
 
power and loss of onsite emergency generators with only one train of emergency busses being backfed from the unit main generator, or the loss of onsite emergency generators with only one train of emergency busses being backfed from offsite power. The subsequent loss of this single power source would escalate the event to a Site Area Emergency in accordance with SS1. Fifteen minutes was selected as a threshold to exclude transient or momentary losses of power.
The EAL allows credit for operation of the Alternate AC Diesel Generator.
Reference Documents:
: 1. 1202.007, "
Degraded Powe r" 2. 1202.008, "
Blackout" 3. 2202.007, "
Loss of Off-Site Power"
: 4. 2202.008, "
Station Blackout"
: 5. 2104.037, "
Alternate AC Diesel Generator Operations" to 0CAN121102
 
Page 103 of 112
 
SYSTEM MALFUNCTION SA 15  103  [At mu lti-unit stations, the EALs should allow credit for operation of installed design features, such as cross
-ties or swing diesels, provided that abnormal or emergency operating procedures address their use. However, these stations must also consider the impa ct of this condition on other shared safety functions in developing the site specific EAL.
]  [Plants that have a proceduralized capability to cross
-tie AC power from an off
-site power supply of a companion unit may take credit for the redundant power sour ce in the associated EAL for this IC.] 
. to 0CAN121102
 
Page 104 of 112
 
SYSTEM MALFUNCTION SS1  104 Initiating Condition - SITE AREA EMERGENCY Loss of all oO ffsite and all oO n sS ite AC power to Vital 4.16 KVemergency busses >for 15 minutes or longer
. Operating Mode Applicability:
Power Operation s (Mode 1)
Startup (Mode 2)            Hot Standby (Mode 3)            Hot Shutdown (Mode 4)  Example Emergency Action Level (s):  Note: The SM/TSC Director/EOF DirectorEmergency 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 oO ff s-S ite and all oO n s-S ite AC power to Vital 4.16 KV(site specific emergency busses) > for 15 minutes or longer. Basis:  Loss of all AC power to emergency busses comp romises all plant safety systems requiring electric power including Shutdown CoolingRHR , ECCS, Containment Heat Removal and the Ultimate Heat Sink. Prolonged loss of all AC power to emergency busses will lead to loss of Fuel Clad, RCS, and Containment, thus this event can escalate to a General Emergency.
Fifteen minutes was selected as a threshold to exclude transient or momentary losses of offsite power.  [At multi-unit stations, the EALs should allow credit for operation of installed design features, such as cross-ties or swing diesels, provided that abnormal or emergency operating procedures address their use. However, these stations must also consider the impact of this condition on other shared safety functio ns in developing the site specific EAL.
]  [Plants that have a proceduralized capability to cross
-tie AC power from an off
-site power supply of a companion unit may take credit for the redundant power source in the associated EAL for this IC.]  Escalation to General Emergency is via Fission Product Barrier Degradation (F) or IC SG1, "Prolonged lL oss of aA ll oO ffsite Power and aProlonged Loss of A ll oO nsite AC pP ower to Vital 4.16 KV busses
."  Reference Documents:
: 1. 1202.007, "
Degraded Power"
: 2. 1202.008, "Blackout"
: 3. 2202.007, "
Loss of Off-Site Power"
: 4. 2202.008, "
Station Blackout"
: 5. 2104.037, "
Alternate AC Diesel Generator Operations" to 0CAN121102
 
Page 105 of 112
 
SYSTEM MALFUNCTION SS 32  105 Initiating Condition - SITE AREA EMERGENCY Automatic tScram (T rip) fails to shutdown the reactor and manual actions taken from the reactor control console are not successful in shutting down the reactor Operating Mode Applicability:
Power Operation s (Mode 1)
Startup (Mode 2)                        Example Emergency Action Level (s):  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 reactor power >
5%. (site specific indications of reactor not shutdown). Basis:  Under these conditions, the reactor is producing more heat than the maximum decay heat load for which the safety systems are designed and efforts to bring the reactor subcritical are
 
unsuccessful. A Site Area Emergency is warranted because conditions exist that lead to
 
IMMINENT loss or potential loss of both fuel clad and RCS. 
[The reactor should be considered shutdown when it producing less heat than the maximum decay heat load for which the safety systems are designed (typically 3 to 5% power). For plants using CSFSTs, this EAL equates to the criteria used to determine a valid Subcriticality Red Path.
For BWRs this EAL should be the APRM downscale trip setpoint.
]  Manual scram (trip) actions taken at the reactor control console are any set of actions by the Rr eactor Oo perator(s) at which causes or should cause control rods to be rapidly inserted into the core and shuts down the reactor.
Manual scram (trip) actions are not considered successful if action away from panel C03 (Unit 1) or panels 2C03/2C14 (Unit 2)the reactor control console is required to scram (trip) the reactor.
This EAL is still applicable even if actions taken away from panel C03 (Unit 1) or panels 2C03/2C14 (Unit 2)the reactor control console are successful in shutting the reactor down because the design limits of the fuel may have been exceeded or because of the gross failure of the Reactor Protection System to shutdown the plant. 
[Although this IC may be viewed as redundant to the Fission Product Barrier Degradation IC, its inclusion is necessary to better assure timely recognition and emergency response.
]  Escalation of this event to a General Emergency would be due to a prolonged condition leading to an extreme challenge to either core-cooling or heat removal. to 0CAN121102
 
Page 106 of 112
 
SYSTEM MALFUNCTION SS 43  106 Initiating Condition - SITE AREA EMERGENCY Loss of all vital DC power
> for 15 minutes or longer. Operating Mode Applicability:
Power Operation s (Mode 1)
Startup (Mode 2)            Hot Standby (Mode 3)            Hot Shutdown (Mode 4)  Example Emergency Action Level (s):  Note: The SM/TSC Director/EOF DirectorEmergency 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 105 volts (site specific bus voltage indication) on all (site specific Vv ital DC busses) > for 15 minutes or longer. Basis:  Loss of all DC power compromises ability to monitor and control plant safety functions. Prolonged loss of all DC power will cause core uncovering and loss of containment integrity when there is
 
significant decay heat and sensible heat in the reactor system. 
[Site specific 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. For a 60 cell string of batteries the cell voltage is typically 1.75 Volts per cell. For a 58 string battery set the minimum voltage is typically 1.81 Volts per cell.
]  Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. 
 
Escalation to a General Emergency would occur by Abnormal Rad iation Levels/Radiological Effluent (A), Fission Product Barrier Degradation (F). to 0CAN121102
 
Page 107 of 112
 
SYSTEM MALFUNCTION SS6  107 Initiating Condition - SITE AREA EMERGENCY Inability to monitor a SIGNIFICANT TRANSIENT in progress  Operating Mode Applicability:
Power Operation s (Mode 1)
Startup (Mode 2)            Hot Standby (Mode 3)            Hot Shutdown (Mode 4)  Example Emergency Action Level (s):  Note: The SM/TSC Director/EOF DirectorEmergency 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)
Control Room annunciators associated with safety systems OR  C(Site specific c ontrol Rr oom safety system indication
)  AND  b. A SIGNIFICANT TRANSIENT is in progress.
AND c. Compensatory indications are unavailable.
Basis:  This IC is intended to recognize the threat to plant safety associated with the complete loss of capability of the control room staff to monitor plant response to a SIGNIFICANT TRANSIENT.  "Planned" and "UNPLANNED" actions are not differentiated since the loss of instrumentation of this magnitude is of such significance during a transient that the cause of the loss is not an
 
ameliorating factor.
Quantification is arbitrary, however, it is estimat ed that if approximately 75% of the safety system annunciators or indicators are lost, there is an increased risk that a degraded plant condition
 
could go undetected. It is not intended that plant personnel perform a detailed count of the
 
instrumentation lost but use the value as a judgment threshold for determining the severity of the
 
plant conditions. It is also not intended that the Shift ManagerSupervisor be tasked with making a    to 0CAN121102
 
Page 108 of 112
 
SYSTEM MALFUNCTION SS6  108 judgment decision as to whether additional personnel are required to provide increased monitoring of system operation.
It is further recognized that most plant designs provide redundant safety system indication powered from separate uninterruptible power supplies. While failure of a large portion of
 
annunciators is more likely than a failure of a large portion of indications, the concern is included
 
in this EAL due to difficulty associated with assessment of plant conditions. The loss of specific, or several, safety system indicators should re main a function of that specific system or component operability status. This will be addressed by the specific Technical Specification. The
 
initiation of a Technical Specification imposed plant shutdown related to the instrument loss will
 
be reported via 10 CFR 50.72. If the shutdown is not in compliance with the Technical
 
Specification action, the N O UE is based on SU 112 "Inability to rR each rR equired operating modeShutdown wW ithin Technical Specification lL imits."  A Site Area Emergency is considered to exist if the Cc ontrol Rr oom staff cannot monitor safety functions needed for protection of the public while a significant transient is in progress. 
[Site specific annunciators for this EAL should be limited to include those identified in the Abnormal Operating Procedures, in the Emergency Ope rating Procedures, and in other EALs (.g., area, process, and/or effluent rad monitors, etc.)
]  Site specific indications needed to monitor safety functions necessary for protection of the public must include Cc ontrol Rr oom indications, computer generated indications and dedicated annunciation capability.
Indicators associated with safety systems are those indicators for reactivity control, core cooling, maintaining reactor coolant system integrity or maintaining containment integrity.
[The specific indications should be those used to determine such functions as the ability to shut down the reactor, maintain the core cooled, to maintain the reactor coolant system intact, maintain the spent fuel cooled, and to maintain containment intact.
]  "Compensatory indications" in this context includes computer based information such as SPDS , QSPDS, COLSS, etc. This should include all computer systems available for this use depending on specific plant design and subsequent retrofits.
Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.
Reference Documents:
: 1. 1015.037, "Post Transient Review"
: 2. 1203.043, "Loss of Control Room Annunciators"
: 3. 2203.042, "Loss of Control Room Annunciators"
  [Due to the limited number of safety systems in operation during cold shutdown, refueling and defueled modes, no IC is indicated during these modes of operation.
]    to 0CAN121102
 
Page 109 of 112
 
SYSTEM MALFUNCTION SG1  109 Initiating Condition - GENERAL EMERGENCY Prolonged loss of all oO ffsite and all oO n s-S ite AC power to safetyemergency busses  Operating Mode Applicability:
Power Operation s (Mode 1)
Startup (Mode 2)            Hot Standby (Mode 3)            Hot Shutdown (Mode 4)  Example Emergency Action Level (s):  1. a. Loss of all offsite and all onsite AC power to safety(s ite specific emergency busses). AND    b. Either EITHER of the following:  Restoration of at least one safetyemergency bus in < less than (site specific 4 hours) is not likely.
OR  Continuing degradation of core cooling based on Fission Product Barrier monitoring as indicated by CETs > 700&deg;F.(Site specific indication of continuing degradation of core cooling based on Fission Product Barrier monitoring.)
Basis:  Loss of all AC power to emergency busses comp romises all plant safety systems requiring electric power including Shutdown CoolingRHR , ECCS, Containment Heat Removal and the Ultimate Heat Sink. Prolonged loss of all AC power to emergency busses will lead to loss of fuel clad, RCS, and containment, thus warranting declaration of a General Emergency. 
[The (site-specific hours) to restore AC power can be based on a site blackout coping analysis performed in conformance with 10 CFR 50.63 and Regulatory Guide 1.155, "Station Blackout," as available. Appropriate allowance for off
-site emergency response including evacuation of surrounding areas should be considered. Although this IC may be viewed as redundant to the Fission Product Barrier Degradation IC, its inclusion is necessary to better assure timely recognition and emergency response.
]  This IC is specified to assure that in the unlikely event of a prolonged station blackout, timely recognition of the seriousness of the event occurs and that declaration of a General Emergency
 
occurs as early as is appropriate, based on a r easonable assessment of the event trajectory.
The likelihood of restoring at least one emergency bus should be based on a realistic appraisal of the situation since a delay in an upgrade decision based on only a chance of mitigating the event
 
could result in a loss of valuable time in preparing and implementing public protective actions. to 0CAN121102
 
Page 110 of 112
 
SYSTEM MALFUNCTION SG1  110 In addition, under these conditions, fission product barrier monitoring capability may be degraded. 
 
[Although it may be difficult to predict when power can be restored, it is necessary to give the SM/TSC Director/EOF Director Emergency Director a reasonable idea of how quickly (s)he may need to declare a General Emergency based on two major considerations:
: 1. Are there any present indications that core cooling is already degraded to the point that loss or potential loss of Fission Product Barriers is IMMINENT? 
: 2. If there are no present indications of such core cooling degradation, how likely is it that power can be restored in time to assure that a loss of two barriers with a potential loss of the third barrier can be prevented?
Thus, indication of continuing core cooling degradation must be based on Fission Product Barrier monitoring with particular emphasis on SM/TSC Director/EOF Director Emergency Director judgment as it relates to IMMINENT loss or potential loss of fission product barriers and degraded ability to monitor fission product barriers
.] Reference Documents:
: 1. Unit 1 Calculation 85-E-0072-02, "Time from Loss of All AC Power to Loss of Subcooling"
: 2. Unit 2 Calculation 85-E-0072-01, "Time from Loss of All AC Power to Loss of Subcooling" to 0CAN121102
 
Page 111 of 112
 
SYSTEM MALFUNCTION SG 32  Initiating Condition - GENERAL EMERGENCY Automatic tScram (T rip) and all manual actions fail to shutdown the reactor and indication of an extreme challenge to the ability to cool the core exists  Operating Mode Applicability:
Power Operation s (Mode 1)
Startup (Mode 2)  Example Emergency Action Level (s):  1. a. An automatic scram (trip) failed to shutdown the reactor AND  b. All manual actions do not shutdown the reactor as indicated by reactor  power > 5% (site specific indications of reactor not shutdown)
. AND  c. Either EITHER of the following exist or have occurred due to continued power generation: 
(Site specific indication that core cooling is extremely challenged.)
CET temperatures at or approaching 1200&deg; F OR    (Site specific indication that heat removal is extremely challenged.)
Feedwater flow rate less than:
Unit 1:  430 gpm Unit 2:  485 gpm Basis:  Under these conditions, the reactor is producing more heat than the maximum decay heat load
 
for which the safety systems are designed and efforts to bring the reactor subcritical are
 
unsuccessful. 
[The reactor should be considered shutdown when it producing less heat than the maximum decay heat load for which the safety systems are designed (typically 3 to 5% power). For plants using CSFSTs, this EAL equates to the criteria used to determine a valid Subcriticality Red Path.
For BWRs this EAL should be the APRM downscale trip setpoint.
]  [For PWRs, the extreme challenge to the ability to cool the core is intended to mean that the core exit temperatures are at or approaching 1200 degrees F or that the reactor vessel water level is    to 0CAN121102
 
Page 112 of 112
 
SYSTEM MALFUNCTION SG 32  below the top of active fuel. For plants using CSFSTs, this EAL equates to a Core Cooling RED condition combined with a Subcriticality RED condition.
]  [For BWRs, the extreme challenge to the ability to cool the core is intended to mean that the reactor vessel water level cannot be restored and maintained above Minimum Steam Cooling RPV Water Level as described in the EOP bases.
]  [Another consideration is the inability to initially remove heat during the early stages of this sequence. For PWRs, if emergency feedwater flow is insufficient to remove the amount of heat required by design from at least one steam generator, an extreme challenge should be considered to exist. For plants using CSFSTs, this EAL equates to a Heat Sink RED condition combined with a Subcriticality RED condition.
]  [For BWRs, considerations include inability to remove heat via the main condenser, or via the suppression pool or torus (e.g., due to high pool water temperature)
.]  In the event either of these challenges exists at a time that the reactor has not been brought below the power associated with the safety syst em design a core melt sequence exists. In this situation, core degradation can occur rapidly. For this reason, the General Emergency declaration
 
is intended to be anticipatory of the fission product barrier table declaration to permit maximum
 
off-site intervention time.
 
Attachment 3 to 0CAN121102 Proposed Technical Basis Document (Clean) 1 
 
ANO EAL BASIS DOCUMENT to 0CAN121102
 
Page 2 of 110
 
ANO EAL BASIS DOCUMENT TABLE OF CONTENTS SECTION PAGE General Notes on Basis Document Use............................................................................ 4 Definitions.......................................................................................................................... 5 Abnormal Radiation Levels / Radiological Effluents.......................................................... 9 AU1................................................................................................................................ 10 AU2................................................................................................................................ 13 AA1................................................................................................................................ 15 AA2................................................................................................................................ 18 AA3................................................................................................................................ 20 AS1................................................................................................................................ 21 AG1............................................................................................................................... 23 Cold Shutdown / Refueling System Malfunction................................................................ 25 CU1............................................................................................................................... 26 CU2............................................................................................................................... 27 CU3............................................................................................................................... 29 CU5............................................................................................................................... 30 CU6............................................................................................................................... 31 CU7............................................................................................................................... 32 CU8............................................................................................................................... 33 CA1................................................................................................................................ 34 CA3................................................................................................................................ 36 CA5................................................................................................................................ 38 CS1................................................................................................................................ 39 CG1............................................................................................................................... 41 Independent Spent Fuel Storage Installation (ISFSI) Malfunction..................................... 43 E-HU1............................................................................................................................ 44 Fission Product Barrier Degradation.................................................................................. 45 General Bases............................................................................................................... 46 Fuel Clad Barrier EALs (FCB)....................................................................................... 47 RCS Barrier EALs (RCB)............................................................................................... 52 Containment Barrier EALS (CNB)................................................................................. 55    to 0CAN121102
 
Page 3 of 110
 
ANO EAL BASIS DOCUMENT TABLE OF CONTENTS SECTION PAGE Hazards and Other Conditions Affecting Plant Safety....................................................... 62 HU1............................................................................................................................... 63 HU2............................................................................................................................... 64 HU4............................................................................................................................... 66 HU5............................................................................................................................... 68 HU6............................................................................................................................... 69 HA1................................................................................................................................ 73 HA2................................................................................................................................ 75 HA3................................................................................................................................ 76 HA4................................................................................................................................ 77 HA5................................................................................................................................ 79 HA6................................................................................................................................ 80 HS1................................................................................................................................ 84 HS2................................................................................................................................ 85 HS3................................................................................................................................ 86 HG1............................................................................................................................... 87 HG2............................................................................................................................... 88 System Malfunction............................................................................................................
89 SU1................................................................................................................................ 90 SU6................................................................................................................................ 91 SU7................................................................................................................................ 93 SU8................................................................................................................................ 94 SU9................................................................................................................................ 95 SU10.............................................................................................................................. 97 SU11.............................................................................................................................. 98 SA1................................................................................................................................ 99 SA3................................................................................................................................ 100 SA6................................................................................................................................ 101 SS1................................................................................................................................ 103 SS3................................................................................................................................ 104 SS4................................................................................................................................ 105 SS6................................................................................................................................ 106 SG1............................................................................................................................... 108 SG3............................................................................................................................... 110    to 0CAN121102
 
Page 4 of 110
 
ANO EAL BASIS DOCUMENT GENERAL NOTES ON BASIS DOCUMENT USE Plant Operating Mode Usage for ANO EALs:
UNIT 1: Mode 1 = Power Operation - Keff  0.99, Reactor Power > 5%
Mode 2 = Startup - Keff  .99, Reactor Power  5% Mode 3 = Hot Standby - Keff < .99, RCS  280 &deg;F Mode 4 = Hot Shutdown - Keff < .99, 280 &deg;F > RCS >200 &deg;F
 
Mode 5 = Cold Shutdown - Keff < .99, RCS  200 &deg;F Mode 6 = Refueling - One or more reactor vessel head closure bolts less than fully tensioned
 
Defueled (D) - All reactor fuel removed from reactor pressure vessel (full core offload during refueling or extended outage). This is not an operating mode designation by Technical
 
Specifications.
 
UNIT 2: Mode 1 = Power Operation - Keff  0.99, Reactor Power > 5%, RCS  300 &deg;F Mode 2 = Startup - Keff  .99, Reactor Power  5%, RCS  300 &deg;F Mode 3 = Hot Standby - Keff < .99, Reactor Power 0, RCS  300 &deg;F Mode 4 = Hot Shutdown - Keff < .99, Reactor Power 0, 300 &deg;F > RCS > 200 &deg;F
 
Mode 5 = Cold Shutdown - Keff < .99, Reactor Power 0, RCS  200 &deg;F  Mode 6 = Refueling - Keff  .95, Reactor Power 0, RCS  140 &deg;F Reactor vessel head unbolted or removed and fuel in the vessel Defueled (D) - All reactor fuel removed from reactor pressure vessel (full core offload during refueling or extended outage). This is not an operating mode designation by Technical
 
Specifications.
This basis document serves two basic functions:
It provides background and explanatory information based on NEI 99-01 to present a basis for the origination of the ANO EALs for reviewers and users. The second function this basis document may provide is an aid to decision makers when making a determination to classify an emergency event. It is intended that decision
 
makers have all the information in Attachment 7.1 of this procedure that they need to
 
make a sound classification decision. Information that may be useful to a decision maker
 
in classifying emergency events is also contained in the Basis section for each IC in the Basis Document.
The expectation is that emergency classifications are to be made as soon as conditions are
 
present and recognizable for the classification, but within 15 minutes or less in all cases of
 
conditions present. A decision maker's use of this Basis Document for assistance is not
 
intended to delay the classification. to 0CAN121102
 
Page 5 of 110
 
ANO EAL BASIS DOCUMENT DEFINITIONS
 
The following definitions are taken from NEI 99-01 are applicable to the ANO emergency classification system:
 
AFFECTING SAFE SHUTDOWN:
Event in progress has adversely affected functions that are necessary to bring the plant to and maintain it in the applicable HOT or COLD SHUTDOWN condition. Plant condition applicability
 
is determined by Technical Specification LCOs in effect.
 
Example 1: Event causes damage that results in entry into an LCO that requires the plant to be placed in HOT SHUTDOWN. HOT SHUTDOWN is achievable, but
 
COLD SHUTDOWN is not. This event is not "AFFECTING SAFE
 
SHUTDOWN."
Example 2: Event causes damage that results in entry into an LCO that requires the plant to be placed in COLD SHUTDOWN. HOT SHUTDOWN is achievable, but
 
COLD SHUTDOWN is not. This event is "AFFECTING SAFE SHUTDOWN."
BOMB:  Refers to an explosive device suspected of hav ing sufficient force to damage plant systems or structures.
 
CIVIL DISTURBANCE:
A group of persons violently protesting station operations or activities at the site.
 
CONFINEMENT BOUNDARY:
The barrier(s) between areas containing radioactive substances and the environment.
 
CONTAINMENT CLOSURE:
The site specific procedurally defined actions taken to secure containment and its associated
 
structures, systems, and components as a functional barrier to fission product release under
 
existing plant conditions.
 
EXPLOSION:
A rapid, violent, unconfined combustion, or catastrophic failure of pressurized/energized
 
equipment that imparts energy of sufficient fo rce to potentially damage permanent structures, systems, or components.
 
EXTORTION:
An attempt to cause an action at the station by threat of force.
 
to 0CAN121102
 
Page 6 of 110
 
ANO EAL BASIS DOCUMENT FAULTED:  In a steam generator, the existence of secondary side leakage that results in an uncontrolled
 
drop in steam generator pressure or the steam generator being completely depressurized.
 
FIRE:  Combustion characterized by heat and light. Source s of smoke such as slipping drive belts or overheated electrical equipment do not constitute FIREs.
 
Observation of flame is preferred but is NOT required if large quantities of smoke and heat are
 
observed.
 
HOSTAGE:  A person(s) held as leverage against the station to ensure that demands will be met by the
 
station.
HOSTILE ACTION:
An act toward a Nuclear Power Plant or its personnel that includes the use of violent force to
 
destroy equipment, take hostages, and/or intimidate the  licensee to achieve an end. This
 
includes attack by air, land, or water using guns , explosives, projectiles, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be
 
included. HOSTILE ACTION should not be construed to include acts of civil disobedience or
 
felonious acts that are not part of a concerted attack on the NPP. Non-terrorism-based EALs
 
should be used to address such activities (i.e., this may include violent acts between individuals
 
in the OWNER CONTROLLED AREA).
 
HOSTILE FORCE:
One or more individuals who are engaged in a determined assault, overtly or by stealth and
 
deception, equipped with suitable weapons capable of killing, maiming, or causing destruction.
 
IMMINENT:
Mitigation actions have been ineffective, additional actions are not expected to be successful, and trended information indicates that the event or condition will occur. Where IMMINENT
 
timeframes are specified, they shall apply.
 
INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI):
A complex that is designed and constructed for the interim storage of spent nuclear fuel and
 
other radioactive materials associated with spent fuel storage.
 
INTRUSION:
A person(s) present in a specified area without authorization. Discovery of a BOMB in a
 
specified area is indication of INTRUSION into that area by a HOSTILE FORCE. to 0CAN121102
 
Page 7 of 110
 
ANO EAL BASIS DOCUMENT NORMAL PLANT OPERATIONS:
Activities at the plant site associated with routine testing, maintenance, or equipment
 
operations, in accordance with normal operating or administrative procedures. Entry into
 
off-normal or emergency operating procedures, or dev iation from normal security or radiological controls posture, is a departure from NORMAL PLANT OPERATIONS.
 
OWNER CONTROLLED AREA (OCA):
The external area contiguous to the designated reactor site Protected Area over which site
 
Security exercises control. The OCA extends outward to the Entergy site property lines.
 
PROJECTILE:
An object directed toward a Nuclear Power Plant that could cause concern for its continued
 
operability, reliability, or personnel safety.
 
PROTECTED AREA:
An area encompassed by physical barriers (i.e., the security fence) and to which access is
 
controlled.
 
RUPTURED:
In a steam generator, existence of primary-to-secondary leakage of a magnitude sufficient to
 
require or cause a reactor trip and safety injection. 
 
SABOTAGE:
Deliberate damage, mis-alignment, or mis-operation of plant equipment with the intent to render
 
the equipment inoperable. Equipment found tampered with or damaged due to malicious
 
mischief may not meet the definition of SABOTAGE until this determination is made by security
 
supervision.
 
SECURITY CONDITION:
Any Security Event as listed in the approved security  contingency plan that constitutes a threat/compromise to site security, threat/risk to site personnel, or a potential degradation to the
 
level of safety of the plant. A SECURITY CONDITION does not involve a HOSTILE ACTION.
 
SIGNIFICANT TRANSIENT:
An UNPLANNED event involving one or more of the following: (1) automatic turbine runback
 
> 25% thermal reactor power, (2) electrical load rejection > 25% full electrical load, (3) Reactor
 
Trip, (4) Safety Injection Activation, or (5) thermal power oscillations > 10%.
 
to 0CAN121102
 
Page 8 of 110
 
ANO EAL BASIS DOCUMENT STRIKE ACTION:
A work stoppage within the PROTECTED AREA by a body of workers to enforce compliance
 
with demands made on Entergy or its affiliates. The STRIKE ACTION must threaten to interrupt
 
NORMAL PLANT OPERATIONS.
 
UNISOLABLE:
A breach or leak that cannot be promptly isolated.
 
UNPLANNED:
A parameter change or an event that is not the result of an intended evolution and requires
 
corrective or mitigative actions.
 
VALID:  An indication, report, or condition, is considered to be VALID when it is verified by (1) an
 
instrument channel check, (2) indications on related or redundant indicators, or (3) by direct
 
observation by plant personnel, such that doubt related to the indicator's operability, the
 
condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment.
 
VISIBLE DAMAGE:
Damage to equipment or structure that is readily observable without measurements, testing, or analysis. Damage is sufficient to cause concern regarding the continued operability or reliability
 
of the affected structure, system, or com ponent. Example damage includes: deformation due to heat or impact, denting, penetration, rupture, cracking, paint blistering. Surface blemishes (e.g.,
paint chipping, scratches) should not be included.
 
VITAL AREAS:
Any area within a protected area containing any equi pment, system or device which, by result of failure, destruction or associated release, could directly or indirectly endanger the health and
 
safety of the public.
 
to 0CAN121102
 
Page 9 of 110
 
ABNORMAL RADIATION LEVELS / RADIOLOGICAL EFFLUENTS to 0CAN121102
 
Page 10 of 110
 
ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT AU1  Initiating Condition - NOTIFICATION OF UNUSUAL EVENT Any release of gaseous or liquid radioactivity to the environment > 2 times the ODCM limits for  60 minutes Operating Mode Applicability:
All  Example Emergency Action Level(s): 
(1 or 2 or 3)
Note: The SM 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 Channel 7 on any of the following radiation monitors > the reading shown for  60 minutes:
MONITORS - UNIT 1 LIMIT RX-9820 Containment Purge 5.90E-2 &#xb5;Ci/cc RX-9825 Radwaste Area 5.36E-2 &#xb5;Ci/cc RX-9830 Fuel Handling Area 4.54E-2 &#xb5;Ci/cc RX-9835 Emergency Penetration Room 9.56E-1 &#xb5;Ci/cc MONITORS - UNIT 2 LIMIT 2RX-9820 Containment Purge 4.46E-2 &#xb5;Ci/cc 2RX-9825 Radwaste Area 3.32E-2 &#xb5;Ci/cc 2RX-9830 Fuel Handling Area 4.46E-2 &#xb5;Ci/cc 2RX-9835 Emergency Penetration Room 8.84E-1 &#xb5;Ci/cc 2RX-9840 Post Accident Sampling Building 4.42E-1 &#xb5;Ci/cc 2RX-9845 Aux. Building Extension 1.26E-1 &#xb5;Ci/cc 2RX-9850 Low Level Radwaste Storage Bldg. 1.77E-1 &#xb5;Ci/cc OR    to 0CAN121102
 
Page 11 of 110
 
ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT AU1  2. VALID reading on any of the following radiation monitors > 2 times the alarm setpoint established by a current release permit for  60 minutes.
EFFLUENT MONITORS - Unit 1 RX-9820 Containment Purge (Channel 7 or 9) RE-4830 Waste Gas Radiation Monitor RE-4642 Liquid Radwaste Monitor EFFLUENT MONITORS - Unit 2 2RX-9820 Containment Purge (Channel 7 or 9) 2RE-2429 Waste Gas Decay Tank Vent Line Radiation Monitor 2RE-2330 BMS Liquid Discharge Monitor 2RE-4423 Regenerative Waste Discharge Monitor 2RE-4425 SG Blowdown to Flume Radiation Monitor OR  3. Confirmed grab sample analyses for gaseous or liquid releases indicates concentrations or release rates > 2 times the applicable values of the ODCM for  60 minutes.
Basis:  The SM 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.
 
This IC addresses a potential reduction in the level of safety of the plant as indicated by a
 
radiological release that exceeds regulatory commitments for an extended period of time.
 
ANO incorporates features intended to control the release of radioactive effluents to the
 
environment. Further, there are administrative controls established to prevent unintentional releases, or control and monitor intentional releases. The occurrence of extended, uncontrolled
 
radioactive releases to the environment is indicative of a degradation in these features and/or
 
controls.
 
The ODCM multiples are specified in AU1 and AA1 only to distinguish between non-emergency
 
conditions, and from each other. While these multiples obviously correspond to an offsite dose
 
or dose rate, the emphasis in classifying these events is the degradation in the level of safety of
 
the plant, NOT the magnitude of the associated dose or dose rate.
 
Releases should not be prorated or averaged over 60 minutes. For example, a release
 
exceeding 4 times ODCM limits for 30 minutes does not meet the threshold for this IC.
 
to 0CAN121102
 
Page 12 of 110
 
ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT AU1  This Initiating Condition includes any release for which a radioactivity discharge permit was not prepared, or a release that exceeds the condi tions (e.g., minimum dilution flow, maximum discharge flow, alarm setpoints, etc.) on the applicable permit.
 
EAL #1 This EAL addresses radioactivity releases, that for whatever reason, cause effluent radiation
 
monitor readings to exceed the threshold identified in the EAL.
 
This EAL is intended for sites that have established effluent monitoring on non-routine release
 
pathways for which a discharge permit would not normally be prepared.
 
EAL #2 This EAL addresses radioactivity releases, that for whatever reason, cause effluent radiation
 
monitor readings to exceed the threshold identified in this Initiating Condition established by the
 
release permit. This value may be associated with a planned batch release, or a continuous
 
release path.
 
EAL #3 This EAL addresses uncontrolled releases that are detected by sample analyses, particularly on
 
unmonitored pathways, e.g., spills of radioactive liquids into storm drains, heat exchanger
 
leakage in river water systems, lake, etc.
 
EAL #1 and #2 directly correlate with the IC since annual average meteorology is required to be
 
used in showing compliance with the ODCM and is used in calculating the alarm setpoints.
 
Reference Documents
:  1. 1604.051, "Eberline Radiation Monitor System"
: 2. Offsite Dose Calculation Manual to 0CAN121102
 
Page 13 of 110
 
ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT AU2  Initiating Condition - NOTIFICATION OF UNUSUAL EVENT UNPLANNED rise in plant radiation levels
 
Operating Mode Applicability:
All  Example Emergency Action Level(s): 
(1 or 2) 
: 1. a. UNPLANNED lowering of water level in the refueling canal or spent fuel pool as indicated by:
Personnel observation, refueling crew report, indication on area security camera, borated water source (BWST or RWT) level drop due to makeup demands.
AND  b. VALID Area Radiation Monitor reading rise on any of the following:
Unit 1 RE-8009 Spent Fuel Area RE-8017 Fuel Handling Area Unit 2 2RE-8914 Spent Fuel Area 2RE-8915 Spent Fuel Area 2RE-8916 Spent Fuel Area 2RE-8912 Containment Incore Instrumentation OR  2. UNPLANNED VALID Area Radiation Monitor readings or survey results indicate a rise by a factor of 1000 over normal* levels.
Note: For area radiation monitors with ranges incapable of measuring 1000 times normal* levels, classification shall be based on VALID full scale indication unless surveys confirm that area radiation levels are below 1000 times normal* within 15 minutes of the Area Radiation Monitor indications going to
 
full scale indication.
* Normal can be considered as the highest reading in the past twenty-four hours excluding the current peak value.
to 0CAN121102
 
Page 14 of 110
 
ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT AU2  Basis:  This IC addresses elevated radiation levels as a result of lowered water level above irradiated
 
fuel or events that have resulted, or may result, in UNPLANNED rises in radiation dose rates
 
within plant buildings. These radiation rises represent a loss of control over radioactive material
 
and represent a potential degradation in the level of safety of the plant.
 
EAL #1 The refueling pathway is a site specific combination of cavities, tubes, canals and pools. While
 
a radiation monitor could detect a rise in dose rate due to a drop in the water level, it might not
 
be a reliable indication of whether or not the fuel is covered. For example, a refueling bridge
 
ARM reading may rise due to planned evolutions such as head lift, or even a fuel assembly
 
being raised in the manipulator mast. Also, a monitor could in fact be properly responding to a
 
known event involving transfer or relocation of a source, stored in or near the fuel pool or
 
responding to a planned evolution such as removal of the reactor head. Generally, elevated
 
radiation monitor indications will need to be combined with another indicator (or personnel
 
report) of water loss
.
For refueling events where the water level drops below the RPV flange classification would be
 
via CU2. This event escalates to an Alert per AA2 if irradiated fuel outside the reactor vessel is
 
uncovered. For events involving irradiated fuel in the reactor vessel, escalation would be via the
 
Fission Product Barrier Matrix for events in operating Modes 1-4.
 
EAL #2 This EAL addresses rises 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.
 
This EAL excludes radiation level rises that result from planned activities such as use of
 
radiographic sources and movement of radioactive waste materials. A specific list of ARMs is
 
not required as it would restrict the applicability of the Threshold. The intent is to identify loss of
 
control of radioactive material in any monitored area.
 
to 0CAN121102
 
Page 15 of 110
 
ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT AA1  Initiating Condition - ALERT Any release of gaseous or liquid radioactivity to the environment > 200 times the ODCM limits for  15 minutes Operating Mode Applicability:
All  Example Emergency Action Level(s): 
(1 or 2 or 3)
Note: The SM / TSC Director / EOF 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 Channel 7 on any of the following radiation monitors > the reading shown for  15 minutes:
MONITORS - UNIT 1 LIMIT RX-9820 Containment Purge 5.90E0 &#xb5;Ci/cc RX-9825 Radwaste Area 5.36E0 &#xb5;Ci/cc RX-9830 Fuel Handling Area 4.54E0 &#xb5;Ci/cc RX-9835 Emergency Penetration Room 9.56E+1 &#xb5;Ci/cc MONITORS - UNIT 2 LIMIT 2RX-9820 Containment Purge 4.46E0 &#xb5;Ci/cc 2RX-9825 Radwaste Area 3.32E0 &#xb5;Ci/cc 2RX-9830 Fuel Handling Area 4.46E0 &#xb5;Ci/cc 2RX-9835 Emergency Penetration Room 8.84E+1 &#xb5;Ci/cc 2RX-9840 Post Accident Sampling Building 4.42E+1 &#xb5;Ci/cc 2RX-9845 Aux. Building Extension 1.26E+1 &#xb5;Ci/cc 2RX-9850 Low Level Radwaste Storage Bldg. 1.77E+1 &#xb5;Ci/cc OR    to 0CAN121102
 
Page 16 of 110
 
ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT AA1  2. EITHER VALID reading on any of the following radiation monitors > 200 times the alarm setpoint established by a current release  permit for  15 minutes OR VALID reading greater than the value listed for  15 minutes.
MONITORS - UNIT 1 LIMIT RX-9820 Containment Purge (Channel 7 or 9) N/A RE-4830 Waste Gas Radiation Monitor 9.5E7 cpm RE-4642 Liquid Radwaste Monitor 9.5E7 cpm MONITORS - UNIT 2 LIMIT 2RX-9820 Containment Purge (Channel 7 or 9) N/A 2RE-2429 Waste Gas Monitoring System 9.5E5 cpm 2RE-2330 BMS Liquid Discharge Monitor 9.5E5 cpm 2RE-4423 Regenerative Waste Discharge Monitor 9.5E5 cpm 2RE-4425 SG Blowdown to Flume Radiation Monitor 9.5E5 cpm OR  3. Confirmed grab sample analyses for gaseous or liquid releases indicates concentrations or release rates > 200 times the applicable values of the ODCM for  15 minutes.
Basis:  The SM / TSC Director / EOF 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.
 
This IC addresses an actual or substantial potential reduction in the level of safety of the plant
 
as indicated by a radiological release that exceeds regulatory commitments for an extended
 
period of time. ANO incorporates features intended to control the release of radioactive
 
effluents to the environment. Further, there are administrative controls established to prevent unintentional releases, or control and monitor intentional releases. The occurrence of extended, uncontrolled radioactive releases to the environment is indicative of a degradation in these
 
features and/or controls.
 
The ODCM multiples are specified in AU1 and AA1 only to distinguish between non-emergency
 
conditions, and from each other. While these multiples obviously correspond to an offsite dose
 
or dose rate, the emphasis in classifying these events is the degradation in the level of safety of
 
the plant, NOT the magnitude of the associated dose or dose rate.
 
to 0CAN121102
 
Page 17 of 110
 
ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT AA1  Releases should not be prorated or averaged. For example, a release exceeding 600 times ODCM limits for 5 minutes does not meet the threshold for this IC.
 
This Initiating Condition includes any release for which a release permit was not prepared, or a
 
release that exceeds the conditions (e.g., mi nimum dilution flow, maximum discharge flow, alarm setpoints, etc.) on the applicable permit.
 
EAL #1 This EAL addresses radioactivity releases, that for whatever reason, cause effluent radiation
 
monitor readings to exceed the threshold identified in the Initiating Condition.
 
This EAL is intended for sites that have established effluent monitoring on non-routine release
 
pathways for which a discharge permit would not normally be prepared.
 
EAL #2 This EAL addresses radioactivity releases, that for whatever reason, cause effluent radiation
 
monitor readings to exceed the threshold identified in this Initiating Condition established by the
 
radioactivity discharge permit. This value may be associated with a planned batch release, or a
 
continuous release path. The limit values provided are for those cases in which the maximum
 
monitor range is less than the release permit value multiplied by 200.
 
EAL #3 This EAL addresses uncontrolled releases that are detected by sample analyses, particularly on
 
unmonitored pathways, e.g., spills of radioactive liquids into storm drains, heat exchanger
 
leakage in river water systems, lake, etc.
 
EAL #1 and #2 directly correlate with the IC since annual average meteorology is required to be
 
used in showing compliance with the ODCM and is used in calculating the alarm setpoints.
 
Reference Documents
:  1. 1604.051, "Eberline Radiation Monitor System"
: 2. Offsite Dose Calculation Manual to 0CAN121102
 
Page 18 of 110
 
ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT AA2  Initiating Condition - ALERT 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 Mode Applicability:
All  Example Emergency Action Level(s): 
(1 or 2) 
: 1. A water level drop in the refueling canal or spent fuel pool that will result in irradiated fuel becoming uncovered.
OR  2. VALID alarm on any of the following radiation monitors due to damage to irradiated fuel or loss of water level.
Unit 1 RX-9820 Containment Purge (Channel 7 or 9) RX-9825 Radwaste Area (Channel 7 or 9) RX-9830 Fuel Handling Area (Channel 7 or 9) RE-8060 Containment High Range Radiation Monitors RE-8061 Containment High Range Radiation Monitors RE-8009 Spent Fuel Area RE-8017 Fuel Handling Unit 2 2RX-9820 Containment Purge (Channel 7 or 9) 2RX-9825 Radwaste Area (Channel 7 or 9) 2RX-9830 Fuel Handling Area (Channel 7 or 9) 2RE-8905 Containment Equipment Hatch Area 2RE-8909 Containment Personnel Access Area 2RE-8925-1 Containment High Range Radiation Monitors 2RE-8925-2 Containment High Range Radiation Monitors 2RE-8914 Spent Fuel Area 2RE-8915 Spent Fuel Area 2RE-8916 Spent Fuel Area  2RE-8912 Containment Incore Inst.
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Page 19 of 110
 
ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT AA2  Basis:  This IC addresses rises in radiation dose rates within plant buildings, and may be a precursor to
 
a radioactivity release to the environment.
These events represent a loss of control over radioactive material and represent an actual or substantial potential degradation in the level of
 
safety of the plant.
 
These events escalate from AU2 in that fuel activity has been released, or is anticipated due to
 
fuel heatup. This IC applies to spent fuel requiring water coverage and is not intended to
 
address spent fuel which is licensed for dry storage.
 
EAL #1 Indications may include instrumentation such as water level and local area radiation monitors, and personnel (e.g., refueling crew) reports. Depending on available level indication, the
 
declaration may be based on indications of water makeup rate or drop in applicable borated
 
water storage tank level. Video cameras (Security or outage-related) may allow remote
 
observation of level.
 
EAL #2 This EAL addresses radiation monitor indications of fuel uncovery and/or fuel damage.
 
Elevated ventilation monitor readings may be indication of a radioactivity release from the fuel, confirming that damage has occurred. Elevated background at the ventilation monitor due to
 
water level drop may mask elevated ventilation exhaust airborne activity and needs to be
 
considered.
 
While a radiation monitor could detect a rise in dose rate due to a drop in the water level, it
 
might not be a reliable indication of whether or not the fuel is covered.
 
For example, a refueling bridge ARM reading may rise due to planned evolutions such as head
 
lift, or even a fuel assembly being raised in the manipulator mast. Also, a monitor could in fact
 
be properly responding to a known event involving transfer or relocation of a source, stored in or
 
near the fuel pool or responding to a planned evolution such as removal of the reactor head. 
 
Generally, elevated radiation monitor indications will need to be combined with another indicator (or personnel report) of water loss.
 
Escalation of this emergency classification level, if appropriate, would be based on AS1 or AG1.
 
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Page 20 of 110
 
ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT AA3  Initiating Condition - ALERT Rise in radiation levels within the facility that impedes operation of systems required to maintain
 
plant safety functions
 
Operating Mode Applicability:
All  Example Emergency Action Level(s):
Dose rate > 15 mR/hr in any of the following areas requiring continuous occupancy to maintain plant safety functions:
 
Unit 1 Control Room  Unit 2 Control Room  Central Alarm Station Basis:  This IC addresses elevated radiation levels that impact continued operation in areas requiring
 
continuous occupancy to maintain safe operation or to perform a safe shutdown.
 
The cause and/or magnitude of the rise in radiation levels is not a concern of this IC. The SM/TSC Director/EOF Director must consider the source or cause of the elevated radiation
 
levels and determine if any other IC may be involved.
 
This IC is not meant to apply to rises in the containment dome radiation monitors as these are
 
events which are addressed in the fission product barrier matrix EALs.
 
Areas requiring continuous occupancy include the Control Rooms and the Central Alarm
 
Station.
 
to 0CAN121102
 
Page 21 of 110
 
ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT AS1  Initiating Condition -- SITE AREA EMERGENCY Offsite dose resulting from an actual or IMMINENT release of gaseous radioactivity > 100 mR
 
TEDE or 500 mR child thyroid CDE for the actual or projected duration of the release
 
Operating Mode Applicability:
All  Example Emergency Action Level(s): 
(1 or 2 or 3)
Note: The SM / TSC Director / EOF 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, the classification should be based on EAL #2 instead of EAL #1. Do not delay declaration awaiting dose assessment results.
: 1. VALID reading on Channel 9 on any of the following radiation monitors > the reading shown for  15 minutes:
MONITORS - UNIT 1 LIMIT RX-9820 Containment Purge 5.90E+1 &#xb5;Ci/cc RX-9825 Radwaste Area 5.36E+1 &#xb5;Ci/cc RX-9830 Fuel Handling Area 4.54E+1 &#xb5;Ci/cc RX-9835 Emergency Penetration Room 9.56E+2 &#xb5;Ci/cc MONITORS - UNIT 2 LIMIT 2RX-9820 Containment Purge 4.46E+1 &#xb5;Ci/cc 2RX-9825 Radwaste Area 3.32E+1 &#xb5;Ci/cc 2RX-9830 Fuel Handling Area 4.46E+1 &#xb5;Ci/cc 2RX-9835 Emergency Penetration Room 8.84E+2 &#xb5;Ci/cc 2RX-9840 Post Accident Sampling Building 4.42E+2 &#xb5;Ci/cc 2RX-9845 Aux. Building Extension 1.26E+2 &#xb5;Ci/cc 2RX-9850 Low Level Radwaste Storage Bldg. 1.77E+2 &#xb5;Ci/cc OR  2. Dose assessment using actual meteorology indicates doses > 100 mR TEDE or 500 mR child thyroid CDE at or beyond the site boundary.
to 0CAN121102
 
Page 22 of 110
 
ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT AS1  OR  3. Field survey results indicate closed window dose rates > 100 mR/hr expected to continue for  60 minutes; or analyses of field survey samples indicate child thyroid CDE > 500 mR for one hour of inhalation, at or beyond the site boundary.
Basis:  This IC addresses radioactivity releases that result in doses at or beyond the site boundary that
 
exceed 10% of the EPA Protective Action Guides (PAGs). Releases of this magnitude are
 
associated with the failure of plant systems needed for the protection of the public.
 
EAL #1 The monitor list in EAL #1 includes monitors on all potential release pathways (plant stack, primary-secondary leak, fuel handling accident).
 
EAL #2 Since dose assessment in EAL #2 is based on actual meteorology, whereas the monitor
 
readings in EAL #1 are not, the results from these assessments may indicate that the
 
classification is not warranted, or may indicate that a higher classification is warranted. For this
 
reason, emergency implementing procedures shoul d call for the timely performance of dose assessments using actual meteorology and release information. If the results of these dose
 
assessments are available when the classification is made (e.g., initiated at a lower
 
classification level), the dose assessment results override the monitor reading EALs.
 
EAL #3 Field team surveys in EAL #3 should be perfo rmed at or beyond the SITE BOUNDARY and at the most accurate indicator of the condition. Field data are independent of release elevation
 
and meteorology. The assumed release duration is one hour. Expected post accident source
 
terms would be dominated by noble gases providing the dose rate value. Sampling of
 
radioiodine by adsorption on a charcoal cartridge should determine the iodine value.
 
Reference Documents:
: 1. 1604.051, "Eberline Radiation Monitor System"
: 2. Offsite Dose Calculation Manual to 0CAN121102
 
Page 23 of 110
 
ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT AG1  Initiating Condition -- GENERAL EMERGENCY Offsite dose resulting from an actual or IMMINENT release of gaseous radioactivity > 1000 mR
 
TEDE or 5000 mR child thyroid CDE for the actual or projected duration of the release using
 
actual meteorology
 
Operating Mode Applicability:
All  Example Emergency Action Level(s): 
(1 or 2 or 3)
Note: The SM / TSC Director / EOF 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, the classification should be based on EAL #2 instead of EAL #1. Do not delay declaration awaiting dose assessment results.
: 1. VALID reading on Channel 9 on any of the following radiation monitors > the reading shown for  15 minutes:
MONITORS - UNIT 1 LIMIT RX-9820 Containment Purge 5.90E+2 (Ci/cc) RX-9825 Radwaste Area 5.36E+2 (Ci/cc) RX-9830 Fuel Handling Area 4.54E+2 (Ci/cc) RX-9835 Emergency Penetration Room 9.56E+3 (Ci/cc) MONITORS - UNIT 2 LIMIT 2RX-9820 Containment Purge 4.46E+2 (Ci/cc) 2RX-9825 Radwaste Area 3.32E+2 (Ci/cc) 2RX-9830 Fuel Handling Area 4.46E+2 (Ci/cc) 2RX-9835 Emergency Penetration Room 8.84E+3 (Ci/cc) 2RX-9840 Post Accident Sampling Building 4.42E+3 (Ci/cc) 2RX-9845 Aux. Building Extension 1.26E+3 (Ci/cc) 2RX-9850 Low Level Radwaste Storage Building 1.77E+3 (Ci/cc)  OR  2. Dose assessment using actual meteorology indicates doses > 1000 mR TEDE or 5000 mR child thyroid CDE at or beyond the site boundary.
OR    to 0CAN121102
 
Page 24 of 110
 
ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT AG1  3. Field survey results indicate closed window dose rates > 1000 mR/hr expected to continue for  60 minutes; or analyses of field survey samples indicate child thyroid CDE > 5000 mR for one hour of inhalation, at or beyond the site boundary.
Basis:  This IC addresses radioactivity releases that result in doses at or beyond the site boundary that
 
exceed the EPA Protective Action Guides (PAGs).
Public protective actions will be necessary.
Releases of this magnitude are associated with the failure of plant systems needed for the
 
protection of the public and likely involve fuel damage.
 
EAL #1 The monitor list in EAL #1 includes monitors on all potential release pathways (plant stack, primary-secondary leak, fuel handling accident).
 
EAL #2 Since dose assessment in EAL #2 is based on actual meteorology, whereas the monitor
 
readings in EAL #1 are not, the results from these assessments may indicate that the
 
classification is not warranted. For this reason, emergency implementing procedures should
 
call for the timely performance of dose assessments using actual meteorology and release
 
information. If the results of these dose assessments are available when the classification is
 
made (e.g., initiated at a lower classification level), the dose assessment results override the
 
monitor reading EALs.
 
EAL #3 Field team surveys in EAL #3 should be perfo rmed at or beyond the SITE BOUNDARY and at the most accurate indicator of the condition. Field data are independent of release elevation
 
and meteorology. The assumed release duration is one hour. Expected post accident source
 
terms would be dominated by noble gases providing the dose rate value. Sampling of
 
radioiodine by adsorption on a charcoal cartridge should determine the iodine value.
 
Reference Documents
:  1. 1604.051, "Eberline Radiation Monitor System"
: 2. Offsite Dose Calculation Manual
 
to 0CAN121102
 
Page 25 of 110
 
Cold Shutdown / Refueling System Malfunction    to 0CAN121102
 
Page 26 of 110
 
COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION CU1  Initiating Condition -- NOTIFICATION OF UNUSUAL EVENT RCS leakage
 
Operating Mode Applicability:
Cold Shutdown (Mode 5)
Example Emergency Action Level(s):
Note: The SM 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 level within Pressurizer or RCS level target band for  15 minutes.
Basis:  This IC is considered to be a potential degradation of the level of safety of the plant. The
 
inability to maintain or restore level is indicative of loss of RCS inventory.
 
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.
 
Prolonged loss of RCS Inventory may result in escalation to the Alert emergency classification
 
level via either CA1 or CA3.
 
. to 0CAN121102
 
Page 27 of 110
 
COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION CU2  Initiating Condition -- NOTIFICATION OF UNUSUAL EVENT UNPLANNED loss of RCS / reactor vessel inventory
 
Operating Mode Applicability:
Refueling (Mode 6)
Example Emergency Action Level(s): 
(1 or 2)  Note: The SM 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 / reactor vessel level drop as indicated by either of the following:
: a. RCS / reactor vessel water level drop below the reactor vessel flange for  15 minutes when the RCS / reactor vessel level band is established above the reactor vessel
 
flange  OR  b. RCS / reactor vessel water level drop below the RCS / reactor vessel level band for  15 minutes or longer when the RCS / reactor vessel level band is established below the reactor vessel flange.
OR  2. RCS / reactor vessel  level cannot be monitored with a loss of RCS / reactor vessel inventory as indicated by an unexplained level rise in (as applicable) the Reactor Building
 
Sump, Reactor Drain Tank, Aux. Building Equipment Drain Tank, Aux. Building Sump, or
 
Quench Tank.
Basis:  This IC is a precursor of more serious conditions and considered to be a potential degradation
 
of the level of safety of the plant.
 
Refueling evolutions that lower RCS water level below the reactor vessel flange are carefully
 
planned and procedurally controlled. An UNPLANNED event that results in water level dropping
 
below the reactor vessel  flange, or below the planned RCS water level for the given evolution (if
 
the planned RCS water level is already below the reactor vessel  flange), warrants declaration
 
of an NUE 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 emergency classification
 
level via either CA1 or CA3. to 0CAN121102
 
Page 28 of 110
 
COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION CU2  EAL #1 This EAL involves a drop in RCS level below the top of the reactor vessel flange that continues
 
for 15 minutes due to an UNPLANNED event. This EAL is not applicable to drops in flooded
 
reactor cavity level, which is addressed by AU2 EAL1, until such time as the level drops to the
 
level of the vessel flange.
 
If reactor vessel level continues to drop and reaches the Bottom ID of the RCS Loop then
 
escalation to CA1 would be appropriate.
 
EAL #2 This EAL addresses conditions in the refueling mode when normal means of core temperature
 
indication and RCS level indication may not be available. Redundant means of reactor vessel
 
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 reactor
 
vessel inventory loss was occurring by observing sump and tank level changes. Sump and tank
 
level rises 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 the Alert emergency classification level would be via either CA1 or CA3.
 
to 0CAN121102
 
Page 29 of 110
 
COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION CU3  Initiating Condition -- NOTIFICATION OF UNUSUAL EVENT UNPLANNED loss of decay heat removal capability with irradiated fuel in the reactor vessel
 
Operating Mode Applicability:
Cold Shutdown (Mode 5)  Refueling (Mode 6)
 
Example Emergency Action Level(s): 
(1 or 2)  Note: The SM 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 200 &deg;F.
 
OR  2. Loss of all RCS temperature and RCS/reactor vessel level indication for  15 minutes.
Basis:  This IC is 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.
 
During refueling the level in the reactor vessel will normally be maintained above the reactor
 
vessel flange. Refueling evolutions that lower water level below the reactor vessel flange are
 
carefully planned and procedurally controlled. Loss of forced decay heat removal at reduced
 
inventory may result in more rapid rises in RCS/reactor vessel temperatures depending on the
 
time since shutdown.
 
Normal means of core temperature indication and RCS level indication may not be available in
 
the refueling mode. Redundant means of reactor vessel 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 of refueling
 
modes, EAL 2 would result in declaration of an NUE if both temperature and 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 CA3 based on exceeding its
 
temperature criteria.
 
to 0CAN121102
 
Page 30 of 110
 
COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION CU5  Initiating Condition -- NOTIFICATION OF UNUSUAL EVENT AC power capability to Vital 4.16 KV busses reduced to a single power source  15 minutes such that any additional single failure would result in station blackout
 
Operating Mode Applicability:
Cold Shutdown (Mode 5)  Refueling (Mode 6)
 
Example Emergency Action Level(s):
Note: The SM 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 Vital 4.16 KV busses reduced to a single power source  15 minutes.
AND  b. Any additional single power source failure will result in station blackout.
Basis:  The condition indicated by this IC is the degradat ion of the offsite and onsite AC power systems such that any additional single failure would result in a station blackout. This condition could
 
occur due to a loss of offsite power with a concurrent failure of all but one emergency generator
 
to supply power to its emergency busses. The subsequent loss of this single power source
 
would escalate the event to an Alert in accordance with CA5.
 
Fifteen minutes was selected as a threshold to exclude transient or momentary losses of power.
 
The EAL allows credit for operation of the Alternate AC Diesel Generator.
 
Reference Documents
:  1. 1202.007, "
Degraded Powe r"  2. 1202.008, "
Blackout"  3. 2202.007, "
Loss of Off-Site Power"
: 4. 2202.008, "
Station Blackout"
: 5. 2104.037, "
Alternate AC Diesel Generator Operations" to 0CAN121102
 
Page 31 of 110
 
COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION CU6  Initiating Condition -- NOTIFICATION OF UNUSUAL EVENT Loss of required DC power  15 minutes Operating Mode Applicability:
Cold Shutdown (Mode 5)  Refueling (Mode 6)
 
Example Emergency Action Level(s):
Note: The SM 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. < 105 volts on required Vital DC bus  15 minutes.
Basis:  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.
 
It is intended that the loss of the operating (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 CA3.
 
Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.
 
to 0CAN121102
 
Page 32 of 110
 
COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION CU7  Initiating Condition -- NOTIFICATION OF UNUSUAL EVENT Inadvertent criticality
 
Operating Mode Applicability:
Cold Shutdown (Mode 5)  Refueling (Mode 6)
 
Example Emergency Action Level(s):
: 1. UNPLANNED sustained positive startup rate observed on nuclear instrumentation.
 
Basis:  This IC addresses criticality events that occur in Cold Shutdown or Refueling modes such as
 
fuel mis-loading events and inadvertent dilution events. This IC indicates a potential
 
degradation of the level of safety of the plant, warranting an NUE classification.
 
This condition can be identified using the startup rate meter. The term "sustained" is used in
 
order to allow exclusion of expected short te rm positive startup rates from planned fuel bundle or control rod movements during core alteration. These short term positive startup rates are the
 
result of the rise in neutron population due to subcritical multiplication.
 
Escalation would be by SM judgment.
 
to 0CAN121102
 
Page 33 of 110
 
COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION CU8  Initiating Condition -- NOTIFICATION OF UNUSUAL EVENT Loss of all onsite or offsite communications capabilities
 
Operating Mode Applicability:
Cold Shutdown (Mode 5)  Refueling (Mode 6)
Defueled
 
Example Emergency Action Level(s): 
(1 or 2) 
: 1. Loss of all Table C2 onsite communication methods affecting the ability to perform routine operations.
OR  2. Loss of all Table C3 offsite communication methods affecting the ability to perform offsite notifications.
Table C2 Onsite Communications Methods  Table C3 Offsite Communications Methods Station radio system  All telephone lines (commercial and microwave) Plant paging system  ENS In-plant telephones  Gaitronics Basis:  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 issues with offsite authorities. The loss of off-
 
site communications ability is expected to be significantly more comprehensive than the
 
condition addressed by 10 CFR 50.72.
 
The availability of one method of ordinary offsite communications is sufficient to inform federal, state, and local authorities of plant issues. This EAL is intended to be used only when
 
extraordinary means (e.g., relaying of information from radio transmissions, individuals being
 
sent to offsite locations, etc.) are being utilized to make communications possible.
 
to 0CAN121102
 
Page 34 of 110
 
COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION CA1  Initiating Condition - ALERT Loss of RCS / reactor vessel inventory
 
Operating Mode Applicability:
Cold Shutdown (Mode 5)  Refueling (Mode 6)
 
Example Emergency Action Level(s): 
(1 or 2)  Note: The SM / TSC Director / EOF 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 / reactor vessel inventory as indicated by:
 
Unit 1: RVLMS Levels 1 through 8 indicate DRY Unit 2: RVLMS Levels 1 through 5 indicate DRY OR  Unit 1: Reactor vessel level < 368 ft., 0 in. (bottom of the hot leg)
Unit 2: Reactor vessel level < 369 ft., 1.5 in. (bottom of the hot leg)
OR  2. RCS / reactor vessel level cannot be monitored for  15 minutes with a loss of RCS /
reactor vessel inventory as indicated by an unexplained level rise in (as applicable) the
 
Reactor Building Sump, Reactor Drain Tank, Aux. Building Equipment Drain Tank, Aux.
 
Building Sump, or Quench Tank.
Basis:  These 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 reactor vessel level lowering and potential core uncovery. This
 
condition will result in a minimum emergency classification level of an Alert.
 
EAL #1 The bottom of the RCS hot leg penetration into the reactor vessel is approximately RLVMS
 
Level 8 (Unit 1) or RVLMS Level 5 (Unit 2). Ho wever, RVLMS may not be available in mode 6.
Redundant means level indication is provided in this mode and included in EAL #1. The bottom
 
of the RCS hot leg penetration into the reactor vessel is 368 ft., 0 in. (Unit 1) or 369 ft., 1.5 in.
(Unit 2). Below this level, reactor vessel level indication will be lost and loss of suction to decay
 
heat removal systems will occur. The inability to restore and maintain level after reaching this
 
setpoint would be indicative of a failure of the RCS barrier. to 0CAN121102
 
Page 35 of 110
 
COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION CA1  EAL #2 In the cold shutdown mode, normal RCS level and reactor vessel level instrumentation systems
 
will usually be available. In the refueling mode, normal means of reactor vessel level indication
 
may not be available. Redundant means of reactor vessel level indication will usually 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 reactor vessel inventory loss was
 
occurring by observing sump and tank level changes. Sump and tank level rises 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.
 
If reactor vessel level continues to lower then escalation to Site Area Emergency will be via CS1.
 
to 0CAN121102
 
Page 36 of 110
 
COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION CA3  Initiating Condition - ALERT Inability to maintain plant in Cold Shutdown
 
Operating Mode Applicability:
Cold Shutdown (Mode 5)  Refueling (Mode 6)
 
Example Emergency Action Level(s): 
(1 or 2) 
: 1. An UNPLANNED event results in RCS temperature > 200 &deg;F > the specified duration in Table C1.
Table C1 RCS Reheat Duration Thresholds RCS Containment Closure Duration Intact (but not RCS Lowered Inventory) N/A 60 minutes* Established 20 minutes*
Not intact or RCS Lowered Inventory Not Established 0 minutes
* If an RCS heat removal system is in operation within this time frame and RCS temperature is being reduced, the EAL is not applicable.
OR  Note: EAL #2 does not apply in solid plant conditions.
: 2. An UNPLANNED event results in RCS pressure rise > 10 psi due to a loss of RCS cooling.
 
Basis:  EAL #1 The RCS Reheat Duration Threshold table addresses complete loss of functions required for
 
core cooling for greater than 60 minutes during refueling and cold shutdown modes when RCS
 
integrity is established. 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 60 minute time frame should allow sufficient time to restore
 
cooling without there being a substantial degradation in plant safety.
 
The RCS Reheat Duration Threshold table also addresses the complete loss of functions
 
required for core cooling for greater than 20 minutes during refueling and cold shutdown modes
 
when CONTAINMENT CLOSURE is established but RCS integrity is not established or RCS
 
inventory is reduced (e.g., mid-loop operation ). As discussed above, RCS integrity should be    to 0CAN121102
 
Page 37 of 110
 
COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION CA3  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.
 
Finally, the EAL addresses complete loss of functions required for core cooling during refueling
 
and cold shutdown modes when neither CONTAINMENT CLOSURE nor RCS integrity are
 
established.
 
The (*) indicates that this EAL is not applicable if actions are successful in restoring an RCS
 
heat removal system to operation and RCS temperature is being reduced within the specified
 
time frame.
 
EAL #2 The 10 psi pressure rise addresses situations where, due to high decay heat loads, the time
 
provided to restore temperature control, should be less than 60 minutes. The RCS pressure
 
setpoint chosen should be 10 psi or the lowest pressure that the site can read on installed
 
Control Board instrumentation that is equal to or greater than 10 psi.
 
Escalation to Site Area Emergency would be via CS1 should boiling result in significant reactor
 
vessel 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 the Technical Specification cold
 
shutdown temperature limit when the heat removal function is available.
 
The SM / TSC Director / EOF Director must remain alert to events or conditions that lead to the
 
conclusion that exceeding the EAL is IMMINENT. If, in the judgment of the SM / TSC Director /
 
EOF Director, an IMMINENT situation is at hand, the classification should be made as if the
 
threshold has been exceeded.
 
to 0CAN121102
 
Page 38 of 110
 
COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION CA5  Initiating Condition - ALERT Loss of all offsite and all onsite AC power to Vital 4.16KV busses  15 minutes Operating Mode Applicability:
Cold Shutdown (Mode 5)  Refueling (Mode 6)
Defueled
 
Example Emergency Action Level(s):
Note: The SM / TSC Director / EOF 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 offsite and all on-site AC power to Vital 4.16KV busses  15 minutes.
Basis:  Loss of all AC power compromises all plant sa fety systems requiring electric power including DHR/shutdown cooling, emergency core cooling, containment cooling, spent fuel pool cooling
 
and the ultimate heat sink.
 
The event can be classified as an Alert when in cold shutdown, refueling, or defueled mode
 
because of the significantly reduced decay heat and lower temperature and pressure, which
 
allow raising 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 Radiation Levels/
 
Radiological Effluent (TAB A) ICs.
 
Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.
 
to 0CAN121102
 
Page 39 of 110
 
COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION CS1  Initiating Condition - SITE AREA EMERGENCY Loss of RCS / reactor vessel inventory affecting core decay heat removal capability
 
Operating Mode Applicability:
Cold Shutdown (Mode 5)  Refueling (Mode 6)
 
Example Emergency Action Level(s): 
(1 or 2)  Note: The SM / TSC Director / EOF 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:
Unit 1: RVLMS Levels 1 through 9 indicate DRY Unit 2: RVLMS Levels 1 through 6 indicate DRY OR  2. With CONTAINMENT CLOSURE established, core exit thermocouples indicate superheat.
 
OR  3. RCS / reactor vessel level cannot be monitored for  30 minutes with a loss of RCS /
reactor vessel inventory as indicated by any of the following:  Containment High Range Radiation Monitor reading > 10 R/hr  Erratic source range monitor indication  Unexplained level rise in Reactor Building Sump, Reactor Drain Tank, Quench Tank, Aux. Building Equipment Drain Tank, or Aux. Building Sump.
Basis:  Under the conditions specified by this IC, continued lowering in RCS / reactor vessel level is
 
indicative of a loss of inventory control. Inventory loss may be due to an RCS breach, pressure
 
boundary leakage, or continued boiling in the RPV. Thus, declaration of a Site Area Emergency
 
is warranted.
 
Escalation to a General Emergency is via CG1 or AG1.
 
to 0CAN121102
 
Page 40 of 110
 
COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION CS1  EAL #3 In the cold shutdown mode, normal RCS level and reactor vessel level instrumentation systems
 
will usually be available. In the refueling mode, normal means of reactor vessel level indication
 
may not be available. Redundant means of reactor vessel level indication will usually 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 reactor vessel inventory loss was
 
occurring by observing sump and tank level changes. Sump and tank level rises 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 30-minute duration allows sufficient time for actions to be performed to recover inventory
 
control equipment.
 
As water level in the reactor vessel lowers, the dose rate above the core will rise. The dose rate
 
due to this core shine should result in site specific monitor indication and possible alarm.
 
to 0CAN121102
 
Page 41 of 110
 
COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION CG1  Initiating Condition - GENERAL EMERGENCY Loss of RCS / reactor vessel inventory affecting fuel clad integrity with containment challenged
 
Operating Mode Applicability:
Cold Shutdown (Mode 5)  Refueling (Mode 6)
 
Example Emergency Action Level(s):
Note: The SM / TSC Director / EOF 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. Core exit thermocouples indicate superheat for  30 minutes.
AND  b. Any of the following containment challenge indications:  CONTAINMENT CLOSURE not established  Explosive mixture inside containment  UNPLANNED rise in containment pressure OR  2. a. RCS / reactor vessel level cannot be monitored with core uncovery indicated by any of the following for  30 minutes:  Containment High Range Radiation Monitor reading > 10R/hr  Erratic source range monitor indication  UNPLANNED level rise in Reactor Building Sump, Reactor Drain Tank, Quench Tank, Aux. Building Equipment Drain Tank, or Aux. Building Sump AND  b. Any of the following containment challenge indications:  CONTAINMENT CLOSURE not established  Explosive mixture inside containment  UNPLANNED rise in containment pressure to 0CAN121102
 
Page 42 of 110
 
COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION CG1  Basis:  This IC represents the inability to restore and maintain reactor vessel level to above the top of
 
active fuel with containment challenged. Fuel damage is probable if reactor vessel level cannot
 
be restored, as available decay heat will cause boiling, further reducing the reactor vessel level. 
 
With the 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.
 
A number of variables can have a significant impact on heat removal capability challenging the
 
fuel clad barrier. Examples include mid-loop, reduced level / flange level, head in place, cavity
 
flooded, RCS venting strategy, decay heat removal system design, vortexing pre-disposition, and steam generator U-tube draining.
 
Analysis indicates that core damage may occur within an hour following continued core
 
uncovery therefore, 30 minutes was conservatively chosen.
 
If CONTAINMENT CLOSURE is re-established prior to exceeding the 30 minute core uncovery
 
time limit then escalation to GE would not occur.
 
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 Containment. However, 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.
 
Sump and tank level rises 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.
 
In the cold shutdown mode, normal RCS level and reactor vessel level instrumentation systems
 
will usually be available. In the refueling mode, normal means of reactor vessel level indication
 
may not be available. Redundant means of reactor vessel level indication will usually 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 reactor vessel inventory loss was
 
occurring by observing sump and tank level changes. Sump and tank level rises 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.
 
As water level in the reactor vessel lowers, the dose rate above the core will rise. The dose rate
 
due to this core shine should result in site specific monitor indication and possible alarm.
 
Reference Documents
:  1. ULD-1-SYS-24, "Unit 1 Inadequate Core Cooling"
: 2. ULD-2-SYS-24, "Unit 2 Inadequate Core Cooling" to 0CAN121102
 
Page 43 of 110
 
INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI) MALFUNCTION to 0CAN121102
 
Page 44 of 110
 
ISFSI MALFUNCTION E-HU1    Initiating Condition - NOTIFICATION OF UNUSUAL EVENT Damage to a loaded cask CONFINEMENT BOUNDARY
 
Operating Mode Applicability:
All  Example Emergency Action Level(s):
: 1. Damage to a loaded cask CONFINEMENT BOUNDARY.
 
Basis:  An NUE in this IC is categorized on the basis of the occurrence of an event of sufficient
 
magnitude that a loaded cask CONFINEMENT BOUNDARY is damaged or violated. This includes classification based on a loaded fuel storage cask CONFINEMENT BOUNDARY loss
 
leading to the degradation of the fuel during storage or posing an operational safety problem
 
with respect to its removal from storage.
 
This EAL addresses a dropped cask, a tipped over cask, EXPLOSION, PROJECTILE damage, FIRE damage or natural phenomena affecting a cask (e.g., seismic event, tornado, etc.).
 
to 0CAN121102
 
Page 45 of 110
 
FISSION PRODUCT BARRIER DEGRADATION    to 0CAN121102
 
Page 46 of 110
 
FISSION PRODUCT BARRIERS General Bases The logic used for these initiating conditions reflects the following considerations:
 
The Fuel Clad Barrier and the RCS Barrier are weighted more heavily than the Containment Barrier (See Sections 3.4 and 3.8). NUE ICs associated with RCS and
 
Fuel Clad Barriers are addressed under System Malfunction (S) ICs.
At the Site Area Emergency level, there must be some ability to dynamically assess how far present conditions are from the threshold for a General Emergency. For example, if
 
Fuel Clad and RCS Barrier "Loss" EALs existed, that, in addition to off-site dose
 
assessments, would require continual assessments of radioactive inventory and
 
containment integrity. Alternatively, if both Fuel Clad and RCS Barrier "Potential Loss"
 
EALs existed, the SM / TSC Director / EOF Director would have more assurance that
 
there was no immediate need to escalate to a General Emergency.
The ability to escalate to higher emergency classes as an event deteriorates must be maintained. For example, RCS leakage steadily increasing would represent an
 
increasing risk to public health and safety.
The Containment Barrier should not be declared lost or potentially lost based on exceeding Technical Specification action statement criteria, unless there is an event in
 
progress requiring mitigation by the Containment barrier. When no event is in progress (Loss or Potential Loss of either Fuel Clad and/or RCS) the Containment Barrier status
 
is addressed by Technical Specifications.
 
to 0CAN121102
 
Page 47 of 110
 
FISSION PRODUCT BARRIERS FUEL CLAD Fuel Clad Barrier Emergency Action Levels: FCB1 OR FCB2 OR FCB3 OR FCB4 OR FCB5 OR FCB6 The Fuel Clad barrier consists of the zircalloy or stainless steel fuel bundle tubes that contain
 
the fuel pellets.
: 1. Primary Coolant Activity Level (FCB1)
Loss:  1. Coolant activity > 300 &#xb5;Ci/gm dose equiva lent I-131 activity by Chemistry sample OR  2. Radiation levels > 1000 MR/hr
 
Unit 1: at SA-229 Unit 2: at 2TCD-19 Potential Loss:
None  Basis:  Loss The site specific value corresponds to 300 Ci/gm I-131 equivalent. Assessment by the EAL Task Force indicates that this amount of coolant activity is well above that expected for iodine
 
spikes and corresponds to less than 5% fuel clad damage. This amount of radioactivity
 
indicates significant clad damage and thus the Fuel Clad Barrier is considered lost.
 
A reading of greater than 1000 mR/hr within at one foot from the RCS sample lines (SA-229 for
 
Unit 1, 2TCD-19 for Unit 2) has been determined to correspond to fuel clad failure of
 
approximately 2-5%, and thus the fuel clad barrier is considered lost. This reading is well above
 
that expected for iodine spikes and thus indicates significant clad damage and thus the fuel clad
 
barrier is considered lost.
 
Potential Loss
 
There is no Potential Loss EAL associated with this item.
 
Reference Documents
: 1. ANO Calculation 03-E-0002-01, "Radiation Monitor EAL Setpoints for Fission Product Barrier Degradation" to 0CAN121102
 
Page 48 of 110
 
FISSION PRODUCT BARRIERS FUEL CLAD
: 2. Core Exit Thermocouple Readings (FCB2)
Loss: > 1200 &deg;F CET temperature.
Potential Loss:
Unit 1: ICC exists as evidenced by CETs indicating superheated conditions Unit 2: Average CETs indicate superheat for current RCS pressure Basis:  Loss The Loss EAL of > 1200 &deg;F is consistent with NEI 99-01 and corresponds to significant superheating of the coolant.
 
Potential Loss
 
The Potential Loss EAL corresponds to a loss of subcooling margin.
 
Note that the loss or potential loss EAL for this category will occur after a loss of adequate sub-
 
cooling margin, which represents a loss of the RCS barrier in EAL RCB1, and therefore
 
represents the loss of two barriers, resulting in a Site Area Emergency per FS1. Any loss or
 
potential loss of the containment barrier at that point would escalate to a General Emergency.
 
Reference Documents
: 1. Unit 1 EOP 1202.005, "Inadequate Core Cooling"
: 2. Unit 1 EOP 1202.013, "EOP Figures"
: 3. Unit 2 OP 2202.009, "Functional Recovery"
: 4. ANO Procedure OP 1302.022, "Core Damage Assessment"
: 5. CE-NPSD-241, "Development of the Comprehensive Procedure Guideline for Core Damage Assessment," Task 467
: 6. BWOG EOP Technical Bases Document, Vol. 3, Chapter III.F to 0CAN121102
 
Page 49 of 110
 
FISSION PRODUCT BARRIERS FUEL CLAD
: 3. Reactor Vessel Water Level (FCB3)
Loss: None  Potential Loss:
Unit 1: RVLMS Levels 1 through 9 indicate DRY Unit 2: RVLMS Levels 1 through 7 indicate DRY Basis:  Loss There is no Loss EAL associated with this item.
 
Potential Loss
 
The Reactor Vessel Level Monitoring Systems at AN O do not provide positive indication of core uncovery. The above core level indication provided is used to monitor the approach to and
 
recovery from ICC conditions, but the CETs are used to identify core uncovery, and are the only
 
positive indication of core uncovery.
 
Per reference document #1, the reactor vessel level indicators installed in Unit 1 extend from
 
the top of the reactor vessel to the fuel alignment plate, and information in reference document
 
#2 indicates that the lowest sensor is greater than 2 feet above the top of active fuel. If any of
 
the 4 RCPs are running, flow induced turbulence produced by the pumps renders the reactor
 
vessel level indicator readings invalid.
 
Per reference document #3, only the reactor vessel level indicators above the core are
 
considered part of the ICC monitoring system. Per reference document #4, the lowest sensor
 
above the core, RVLMS LVL 6 on the ICC monitoring panel 2C388, is 47 inches above the top
 
of the core. If any of the 4 RCPs are running, flow induced turbulence produced by the pumps
 
renders the reactor vessel level indicator readings invalid.
 
For either unit then, should CET indication be unavailable and reactor vessel level indication be
 
unavailable due to RCP operation or any other cause, a degraded ability to monitor the barrier
 
would exist.
 
Reference Documents
:  1. ULD-1-SYS-24, "Unit 1 Inadequate Core Cooling System"
: 2. Calculation 84-EQ-0080-02, "Loop Error Analysis for Reactor Vessel Level Monitoring System"  3. ULD-2-SYS-24, "Unit 2 Inadequate Core Cooling Monitoring System"
: 4. Calculation 90-E-0116-01, "Unit 2 EOP Setpoint Document,"
Setpoint R.3    to 0CAN121102
 
Page 50 of 110
 
FISSION PRODUCT BARRIERS FUEL CLAD
: 4. Containment Radiation Monitoring (FCB4)
Loss: Containment high range radiation monitor reading > 1000 R/hr Potential Loss:
None  Basis:  Loss The 1000 R/hr reading on the containment high range radiation monitors (RE-8060 or RE-8061
 
for Unit 1, 2RE-8925-1 or 2RE-8925-2 for Unit 2) is a value which indicates the release of
 
reactor coolant, with elevated activity indicative of fuel damage, into the containment.
 
Reactor coolant concentrations of this magnitude are several times larger than the maximum
 
concentrations (including iodine spiking) allowed within technical specifications and are
 
therefore indicative of fuel damage.
 
This radiation monitor value is higher than that specified for RCS barrier Loss EAL RCB3. 
 
Thus, this EAL indicates a loss of both the Fuel Clad barrier and RCS barrier that appropriately
 
escalates the emergency classification to a Site Area Emergency per FS1.
 
Potential Loss
 
There is no Potential Loss EAL associated with this item.
 
Reference Documents
:  1. NUREG 1228, "Source Term Estimation During Incident Response to Severe Nuclear Power Plant Accidents"
: 2. ANO Calculation 03-E-0002-01, "Radiation Monitor EAL Setpoints for Fission Product Barrier Degradation"
: 5. Core Damage Assessment (FCB5)
Loss: At least 5% fuel clad damage as determined from core damage assessment Potential Loss:
None  Basis:  Loss This level is consistent with other fuel clad barrier loss EALs indicative of significant fuel clad
 
damage, but uses core damage assessment eval uations by Technical Support personnel. The fuel clad barrier is considered lost.
 
to 0CAN121102
 
Page 51 of 110
 
FISSION PRODUCT BARRIERS FUEL CLAD If this determination is made from the high range containment radiation monitor readings, or if accompanied by other indications of a loss or potential loss of the RCS barrier, this EAL
 
condition represents a Site Area Emergency per FS1.
 
Potential Loss
 
There is no potential loss EAL associated with this item.
 
Reference Documents
:  1. ANO Procedure OP-1302.022, "Core Damage Assessment"
: 6. Emergency Director Judgment (FCB6)
Any condition in the opinion of the SM / TSC Director / EOF Director that indicates Loss or
 
Potential Loss of the Fuel Clad barrier.
 
Basis:  This EAL addresses any other factors that are to be used by the SM / TSC Director / EOF
 
Director in determining whether the Fuel Clad barrier is lost or potentially lost. In addition, the
 
inability to monitor the barrier should also be incorporated in this EAL as a factor in SM / TSC
 
Director / EOF Director judgment that the barrier may be considered lost or potentially lost.
 
to 0CAN121102
 
Page 52 of 110
 
FISSION PRODUCT BARRIERS RCS  RCS Barrier EALs:
RCB1 OR RCB2 OR RCB3 OR RCB4 The RCS Barrier includes the RCS primary side and its connections up to and including the
 
pressurizer safety and relief valves, and other connections up to and including the primary
 
isolation valves.
: 1. RCS Leak Rate (RCB1)
Loss: RCS leak rate > available makeup capacity as indicated by:
Unit 1: Loss of adequate subcooling margin Unit 2: RCS subcooling (MTS) can NOT be maintained at least 30 &deg;F Potential Loss:
Unit 1: UNISOLABLE RCS leak > 50 gpm with Letdown isolated Unit 2: UNISOLABLE RCS leak > 44 gpm with Letdown isolated Basis:  Loss  This EAL addresses conditions where leakage from the RCS is greater than available inventory control capacity such that a loss of subcooling has occurred. The loss of subcooling is the
 
fundamental indication that the inventory control systems are inadequate in maintaining RCS
 
pressure and inventory against the mass loss through the leak.
 
Potential Loss This EAL is based on the apparent inability to maintain normal liquid inventory within the
 
Reactor Coolant System (RCS) by normal oper ation of the Makeup and Purification System (Unit 1) or the Chemical and Volume Control System (Unit 2).
For Unit 1 this is based on indications that leakage is greater than normal makeup capacity. 
 
The operator could not batch in water and boric acid to the makeup system fast enough to
 
maintain the makeup tank level during a 50 gpm RCS leak. It is not necessary to perform a
 
detailed assessment of the RCS leakrate to implement this EAL. Any event or condition which, in the judgment of the SM / TSC Director / EOF Director, could result in RCS leakage in excess
 
of Unit 1 normal makeup capacity would meet the intent of this EAL; for example:  Need to open the BWST suction for the operating makeup pump due to lowering makeup tank level  Full or partial HPI is needed to maintain the RCS pressure or pressurizer level  Two out of three seal stages failed on any RCP  RCS pressure lowering due to failure of a primary relief valve to reseat    to 0CAN121102
 
Page 53 of 110
 
FISSION PRODUCT BARRIERS RCS  For Unit 2, this is considered as the capacity of one charging pump discharging to the charging header (44 gpm). Any event or condition which, in the judgment of the SM / TSC Director / EOF
 
Director, could result in RCS leakage in excess of Unit 2 normal makeup capacity would meet
 
the intent of this EAL; for example:  A second charging pump being required is indicative of a substantial RCS leak  Three out of four seal stages failed on any RCP  RCS pressure lowering due to failure of a primary relief valve to reseat
 
Isolating letdown is a standard abnormal operating procedure action and may prevent
 
unnecessary classifications when a non-RCS leakage path such as a Makeup and Purification
 
System or CVCS leak exists. The intent of this condition is met if attempts to isolate Letdown
 
are NOT successful. Additional charging pumps being required is indicative of a substantial
 
RCS leak.
 
Reference Documents
:  1. Unit 1 EOP 1202.013, Figure 1, "Saturation and Adequate SCM"
: 2. Unit 1 EOP Setpoint Document, Calculation  90-E-0116-07, Setpoint B.19
: 3. Unit 2 EOP 2202.009, "Functional Recovery"
: 4. Unit 2 EOP Setpoint Document, Calculation 90-E-0116-01
: 5. Unit 2 SAR Table 9.3-14, Charging Pumps Design Data
: 2. SG Tube Rupture (RCB2)
Loss: SGTR  that results in an ECCS (SI) actuation Potential Loss:
None  Basis:  Loss This EAL addresses the full spectrum of Steam Generator (SG) tube rupture events in conjunction with Containment barrier Loss EALs. It addresses RUPTURED SG(s) for which the
 
leakage is large enough to cause actuation (either automatic or manual) of ECCS (SI). This is
 
consistent to the RCS leak rate barrier Potential Loss EAL.
 
By itself, this EAL will result in the declaration of an Alert. However, if the SG is also FAULTED (i.e., two barriers failed), the declaration escalates to a Site Area Emergency per Containment
 
barrier Loss EAL CNB3. to 0CAN121102
 
Page 54 of 110
 
FISSION PRODUCT BARRIERS RCS  Potential Loss There is no Potential Loss EAL associated with this item.
: 3. Containment Radiation Monitoring (RCB3)
Loss: Containment high range radiation monitor reading > 100 R/hr.
Potential Loss:
None  Basis  Loss  The 100 R/hr reading on the containment high range radiation monitors (RE-8060 or RE-8061
 
for Unit 1, 2RE-8925-1 or 2RE-8925-2 for Unit 2) is a value which indicates the release of
 
reactor coolant to the containment.
This reading is less than that specified for Fuel Clad barrier EAL FCB4. Thus, this EAL is
 
indicative of a RCS leak only. If the radiation monitor reading rose to that specified by Fuel Clad
 
barrier EAL, fuel damage would also be indicated.
During the initial fifteen minutes after a thermal event inside containment, the high range
 
radiation monitor readings are considered invalid due to possibility of a transient thermally-
 
induced current.
 
Potential Loss There is no Potential Loss EAL associated with this item.
 
Reference Documents
:  1. ANO Calculation 03-E-0002-01 , "Radiation Monitor EAL Setpoints for Fission Product Barrier Degradation"
: 4. Emergency Director Judgment (RCB4)
Any condition in the opinion of the SM / TSC Director / EOF Director that indicates Loss or
 
Potential Loss of the RCS Barrier.
 
Basis:  This EAL addresses any other factors that are to be used by the SM / TSC Director / EOF
 
Director in determining whether the RCS barrier is lost or potentially lost. In addition, the
 
inability to monitor the barrier should also be incorporated in this EAL as a factor in SM / TSC
 
Director / EOF Director judgment that the barrier may be considered lost or potentially lost. to 0CAN121102
 
Page 55 of 110
 
FISSION PRODUCT BARRIERS CONTAINMENT Containment Barrier EALs:
CNB1 OR CNB2 OR CNB3 OR CNB4 OR CNB5 OR CNB6 OR CNB7  The Containment Barrier includes the containment building and connections up to and including
 
the outermost containment isolation valves. This barrier also includes the main steam, feedwater, and blowdown line extensions outside the containment building up to and including
 
the outermost secondary side isolation valve.
: 1. Containment Pressure (CNB1)
Loss:  1. Rapid unexplained drop in containment pressure following an initial rise in containment pressure OR  2. Containment pressure or sump level response not consistent with LOCA conditions
 
Potential Loss:
: 1. Unit 1: Containment pressure > 73.7 PSIA (59 PSIG) and rising Unit 2: Containment pressure > 73.7 PSIA (59 PSIG) and rising OR  2. Explosive mixture exists inside containment.
 
OR  3. a. Containment Pressure > containment spray actuation setpoint
 
UNIT 1: 44.7 PSIA (30 PSIG)
UNIT 2: 23.3 PSIA (8.6 PSIG)
AND  b. LESS THAN one full train of spray operating Basis:  Loss  Rapid unexplained loss of pressure (i.e., not attributable to containment spray or condensation effects) following an initial pressure rise from a primary or secondary high energy line break
 
indicates a loss of containment integrity. Containment pressure and sump levels should rise as
 
a result of mass and energy release into containment from a LOCA. Thus, sump level or
 
pressure not rising indicates containment bypass and a loss of containment integrity. to 0CAN121102
 
Page 56 of 110
 
FISSION PRODUCT BARRIERS CONTAINMENT This indicator relies on operator recognition of an unexpected response for the condition and therefore, does not have a specific value associated with it. The unexpected response is
 
important because it is the indicator for a containment bypass condition.
 
Potential Loss 1.
 
The site specific pressure is based on the containment design pressure.
 
Potential Loss 2.
 
Existence of an explosive mixture means a hy drogen and oxygen concentration of at least the lower deflagration limit curve exists. The hy drogen concentration of 4% has been recognized by the NRC staff as a well-established lower flammability limit in air or steam-air atmospheres that
 
is adequately conservative for protecting against an H2 explosion. Hydrogen control systems at
 
ANO are designed and operated as to maintain the containment hydrogen concentration below
 
this level, so that indications of hydrogen concentrations above this are considered a potential
 
challenge to the containment integrity.
 
Potential Loss 3.
 
This EAL represents a potential loss of containment in that the containment heat
 
removal/depressurization system (e.g., containment sprays, ice condenser fans, etc., but not
 
including containment venting strategies) are ei ther lost or performing in a degraded manner, as indicated by containment pressure greater than the setpoint at which the equipment was
 
supposed to have actuated.
 
Reference Documents
:  1. Unit 1 OP-1105.003, "Engineering Safeguards Actuation System"
: 2. Unit 1 SAR Sections 1.4.43, 5.2.1.2.1, 14.2.2.5.5.1 (reactor building design pressure)
: 3. Unit 1 SAR Section 6.6 (Post-Loss of Coolant Accident Hydrogen Control)
: 4. Unit 1 TS Table 3.3.5-1
: 5. Unit 2 SAR Section 6.2.5 (Combustible Gas Control In Containment)
: 6. Unit 2 SAR Section 3.8.1.3.1.D (Containment Design Pressure)
: 7. Unit 2 TS Table 3.3-4
: 8. Regulatory Guide 1.7, "Control of Combustible Gas Concentrations in Containment Following a Loss-of-Coolant Accident, Rev. 2 1978" to 0CAN121102
 
Page 57 of 110
 
FISSION PRODUCT BARRIERS CONTAINMENT
: 2. Core Exit Thermocouple Readings (CNB2)
Loss: None  Potential Loss:
: 1. a. CETs indicate > 1200 &deg;F AND  b. Restoration procedures not effective within 15 minutes.
OR  2. a. CETs indicate > 700 &deg;F AND  b. RVLMS indicates:
Unit 1: Levels 1 through 9 DRY Unit 2: Levels 1 through 7 DRY AND  c. Restoration procedures not effective within 15 minutes.
Basis:  Loss There is no Loss EAL associated with this item.
 
Potential Loss
 
The conditions in these EALs represent an IMMINENT core melt sequence which, if not
 
corrected, could lead to vessel failure and a higher potential for containment failure. In
 
conjunction with the Core Cooling and RCS Leakage criteria in the Fuel and RCS barrier
 
columns, this threshold would result in the declaration of a General Emergency, i.e., loss of two
 
barriers and the potential loss of a third. If the function restoration procedures are ineffective, there is no "success" path.
 
The function restoration procedures are those emergency operating procedures that address
 
the recovery of the core cooling critical safety functions. The procedure is considered effective
 
if the temperature is dropping or if the vessel water level is rising.
 
Whether or not the procedures will be effective should be apparent within 15 minutes. The SM /
 
TSC Director / EOF Director should make the declaration as soon as it is determined that the
 
procedures have been, or will be ineffective. to 0CAN121102
 
Page 58 of 110
 
FISSION PRODUCT BARRIERS CONTAINMENT
: 3. SG Secondary Side Release With Primary-to-Secondary Leakage (CNB3)
Loss:  1. Primary-to-secondary leakrate > 10 gpm
 
AND  2. UNISOLABLE steam release from affected steam generator to the environment
 
Potential Loss:
None  Basis:  This loss EAL recognizes that SG tube leakage can represent a bypass of the containment
 
barrier as well as a loss of the RCS barrier.
 
This EAL results in a NUE for smaller breaks that; (1) do not exceed the Normal Makeup
 
Capacity for Unit 1 or the capacity of one charging pump in the normal charging lineup for Unit 2 
 
EAL in RCS leak rate barrier Potential Loss , or (2) do not result in ECCS actuation in RCS SG
 
tube rupture barrier Loss. For larger breaks, RCS barrier threshold criteria would result in an
 
Alert. For SG tube ruptures which may involve multiple steam generators or UNISOLABLE
 
secondary line breaks, this condition would exist in conjunction with RCS barrier conditions and
 
would result in a Site Area Emergency. Escalation to General Emergency would be based on "Potential Loss" of the Fuel Clad Barrier.
 
Loss 1.
This EAL addresses the condition in which a RUPTURED steam generator is also FAULTED. 
 
This condition represents a bypass of the RCS and containment barriers and is a subset of the
 
second threshold. In conjunction with RCS leak rate barrier loss EAL RCB2, this would always
 
result in the declaration of a Site Area Emergency.
 
Loss 2.
This EAL addresses SG tube leaks that exceed 10 gpm in conjunction with an UNISOLABLE
 
release path to the environment from the affected steam generator. The threshold for
 
establishing the UNISOLABLE secondary side release is intended to be a prolonged release of
 
radioactivity from the RUPTURED steam generator directly to the environment. This could be expected to occur when the main condenser is unavailable to accept the contaminated steam (i.e., SG tube rupture with concurrent loss of off-site power and the RUPTURED steam
 
generator is required for plant cooldown or a stuck open relief valve). The time it takes to
 
isolate a SG with tube leakage > 10 gpm in accordance with plant specific EOPs is not
 
considered a prolonged release. In this case the SG with tube leakage > 10 gpm with a
 
concurrent loss of offsite power is normally steamed to the environment in a controlled manner to achieve and maintain a RCS Hot Leg temperature below that which corresponds to the Main
 
Steam Safety Valve relief settings. However, if the SG cannot be isolated or if both SGs have tube leakage > 10 gpm, a prolonged release will likely be necessary to support plant cooldown. to 0CAN121102
 
Page 59 of 110
 
FISSION PRODUCT BARRIERS CONTAINMENT If the main condenser is available, there may be releases via air ejectors, gland seal exhausters, and other similar controlled, and o ften monitored, pathways. These pathways do not meet the intent of an UNISOLABLE release path to the environment. These minor releases
 
are assessed using Abnormal Radiation Levels / Radiological Effluent ICs (TAB A).
 
Potential Loss There is no Potential Loss EAL associated with this item.
: 4. Containment Isolation Failure or Bypass (CNB4)
Loss:  1. UNISOLABLE breach of containment AND  2. Direct downstream pathway to the environment exists after containment isolation signal Potential Loss:
None  Basis:  Loss  This EAL addresses incomplete containment isolation that allows a direct release to the
 
environment. A breach of containment has also occurred if an inboard and outboard pair of
 
isolation valves fails to close on an automatic actuation signal or from a manual action in the
 
Control Room and opens a release path to the environment.
The breach is not isolable from the Control Room if an attempt for isolation from the Control
 
Room has been made and was unsuccessful. An attempt for isolation should be made prior to
 
the accident classification. If isolable upon identification, then this Initiating Condition is not
 
applicable.
The use of the modifier "direct" in defining the release path discriminates against release paths
 
through interfacing liquid systems. The existence of an in-line charcoal filter does not make a
 
release path indirect since the filter is not effective at removing fission product noble gases. 
 
Typical filters have an efficiency of 95-99% removal of iodine. Given the magnitude of the core
 
inventory of iodine, significant releases could still occur.
In addition, since the fission product release would be driven by boiling in the reactor vessel, the
 
high humidity in the release stream can be expected to render the filters ineffective in a short
 
period.
 
Potential Loss
 
There is no Potential Loss EAL associated with this item. to 0CAN121102
 
Page 60 of 110
 
FISSION PRODUCT BARRIERS CONTAINMENT
: 5. Containment Radiation Monitoring (CNB5)
Loss: None  Potential Loss:
Containment high range radiation monitor reading > 4000 R/hr
 
Basis:  Loss There is no Loss EAL associated with this item.
 
Potential Loss
 
The 4000 R/hr reading on the containment high range radiation monitors (RE-8060 or RE-8061
 
for Unit 1, 2RE-8925-1 or 2RE-8925-2 for Unit 2) is a value which indicates significant fuel
 
damage well in excess of the EALs associated with both loss of Fuel Clad and loss of RCS
 
barriers. A major release of radioactivity requiring off-site protective actions from core damage
 
is not possible unless a major failure of fuel cladding allows radioactive material to be released
 
from the core into the reactor coolant.
 
Regardless of whether containment is challenged, this amount of activity in containment, if released, could have such severe consequences that it is prudent to treat this as a potential
 
loss of containment, such that a General Emergency declaration is warranted.
 
Because the monitor reading exceeds the readings for Fuel Clad Barrier loss in FCB4 and RCS Barrier loss in RCB3 , the SM/TSC Director/EOF Director should declare a General Emergency when this value on the Containment High Range Rad Monitor is exceeded as a loss of two barriers (fuel clad and RCS) and potential loss of the third (containment).
 
Reference Documents
:  1. ANO Calculation 03-E-0002-01, "Radiation Monitor EAL Setpoints for Fission Product Barrier Degradation"
: 2. NUREG 1228, "Source Term Estimation During Incident Response to Severe Nuclear Power Plant Accidents" to 0CAN121102
 
Page 61 of 110
 
FISSION PRODUCT BARRIERS CONTAINMENT
: 6. Other Indications (CNB6)
Elevated readings on the following radiation monitors that indicate loss or potential loss of the Containment barrier:
MONITORS - UNIT 1 RX-9820 Containment Purge RX-9825 Radwaste Area RX-9830 Fuel Handling Area RX-9835 Emergency Penetration Room MONITORS - UNIT 2 2RX-9820 Containment Purge 2RX-9825 Radwaste Area 2RX-9830 Fuel Handling Area 2RX-9835 Emergency Penetration Room 2RX-9840 Post Accident Sampling Building 2RX-9845 Aux. Building Extension Basis:  This EAL covers other indications that may unambiguously indicate the loss or potential loss of
 
the containment barrier.
: 7. Emergency Director Judgment (CNB7)
Any condition in the opinion of the SM / TSC Director / EOF Director that indicates Loss or Potential Loss of the Containment Barrier.
 
Basis:  This EAL addresses any other factors that are to be used by the SM / TSC Director / EOF Director in determining whether the Containment barrier is lost or potentially lost. In addition, the inability to monitor the barrier should also be incorporated in this EAL as a factor in SM /
 
TSC Director / EOF Director judgment that the barrier may be considered lost or potentially lost.
The Containment barrier should not be declared lost or potentially lost based on exceeding
 
Technical Specification action statement criteria, unless there is an event in progress requiring
 
mitigation by the Containment barrier. When no event is in progress (Loss or Potential Loss of
 
either Fuel Clad and/or RCS) the Containment barrier status is addressed by Technical
 
Specifications. to 0CAN121102
 
Page 62 of 110
 
HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY    to 0CAN121102
 
Page 63 of 110
 
HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU1  Initiating Condition - NOTIFICATION OF UNUSUAL EVENT Confirmed SECURITY CONDITION or threat which indicates a potential degradation in the level
 
of safety of the plant
 
Operating Mode Applicability:
All  Example Emergency Action Level(s): 
(1 or 2 or 3)
: 1. A SECURITY CONDITION that does not involve a HOSTILE ACTION as reported by ANO Security Shift Supervision.
OR  2. A credible site specific security threat notification.
 
OR  3. A validated notification from NRC providing information of an aircraft threat.
 
Basis:  NOTE: Timely and accurate communication between Security Shift Supervision and the Control Room is crucial for the implementation of effective Security EALs.
 
Security events which do not represent 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 HA1, HS1 and HG1.
 
A higher initial classification could be made based upon the nature and timing of the security
 
threat and potential consequences. Consideration shall be given to upgrading the emergency
 
response status and emergency classification in accordance with the Safeguards Contingency
 
Plan and Emergency Plan.
 
EAL #1 The Security Shift Supervisor is the designated individual 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 EAL is based on the Safeguards Contingency Plan. The Safeguards Contingency Plan is
 
based on guidance provided in NEI 03-12. to 0CAN121102
 
Page 64 of 110
 
HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU1  EAL #2 This EAL is included to ensure that appropriate notifications for the security threat are made in a
 
timely manner. This includes information of a credible threat. Only the plant to which the
 
specific threat is made need declare the NUE.
 
The determination of "credible" is made through use of information found in the 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.
 
Only the plant to which the specific threat is made need declare the NUE.
 
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 via HA1 would be appropriate if the threat involves an airliner within
 
30 minutes of the plant.
 
to 0CAN121102
 
Page 65 of 110
 
HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU2  Initiating Condition - NOTIFICATION OF UNUSUAL EVENT Other conditions exist which in the judgment of the SM warrant declaration of an NUE
 
Operating Mode Applicability:
All  Example Emergency Action Level(s):
: 1. Other conditions exist which in the judgment of the SM indicate that events are in progress or have occurred which indicate a potential degradation of the level of safety of the plant or
 
indicate a security threat to facility protection has been initiated. No releases of radioactive
 
material requiring offsite response or monitoring are expected unless further degradation of safety systems occurs.
Basis:  This EAL addresses unanticipated conditions not addressed explicitly elsewhere but that
 
warrant declaration of an emergency because conditions exist which are believed by the SM to
 
fall under the NUE emergency classification level.
 
to 0CAN121102
 
Page 66 of 110
 
HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU4  Initiating Condition - NOTIFICATION OF UNUSUAL EVENT FIRE within the PROTECTED AREA not extinguished within 15 minutes of detection or EXPLOSION within the PROTECTED AREA Operating Mode Applicability:
All  Example Emergency Action Level(s):
(1 or 2)  Note: The SM 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 in any Table H1 structure or area not extinguished 1) within 15 minutes of Control Room notification or 2) within 15 minutes of verification of a Control Room FIRE alarm.
Table H1 Unit 1 Unit 2 CA-1 & HP Office Area Condensate Demineralizer Room
 
Corridor 98
 
Fire Area C
 
Lower North Electrical Penetration Room (LNEPR) Lower South Electrical Equipment Room (LSEER)
/ Air Compressor Room Lower South Electrical Penetration Room (LSEPR) Lower South Piping Penetration Room (LSPPR)
 
Main Steam Isolation Violation (MSIV) Room
 
North Engineered Safeguards (ES) SWGR Room (A4) South ES SWGR Room
 
Turbine Building  A1, A2, H1, H2 SWGR area  354' Bowling Alley north end west of Breathing Air compressor room  368' West Heater Deck from LSEER (orange door) along east wall of ES SWGR Rooms to
 
Corridor 98 door.
Upper North Electrical Penetration Room (UNEPR) / Hot Tool Room / Decon Room Upper South Electrical Penetration Room (USEPR) Upper South Piping Penetration Room (USPPR) 2A3 Room 2A4, 2D02, & East Battery Room
 
2B53 Room
 
2B63 Room
 
2B9/2B10 Room
 
2Y11/13 Equipment Room
 
Auxiliary Building 317' General Access
 
Auxiliary Building 335'
 
Auxiliary Building 354'
 
'B' Engineered Safeguards Features (ESF)
Room Corridor Behind Door 340
 
Turbine Building  2A1, 2A2, 2H1, 2H2 Area  354' West wall of Demineralizer area  368' West Heater Deck north of north Switchgear (SWGR) Room (2A3) and
 
East of LNEPR Intake Structure  354' or 366'
 
LNEPR LSEPR Motor-Generator (MG) Set Room
 
Steam Pipe Area
 
Hot Machine Shop
 
UNEPR, UNPPR, LNPPR, USPPR    to 0CAN121102
 
Page 67 of 110
 
HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU4    OR  2. EXPLOSION within the PROTECTED AREA.
 
Basis:  This IC 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 15-minute duration is to size the FIRE and to discriminate against small FIRES
 
that are readily extinguished (e.g., smoldering waste paper basket).
 
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 SM also needs to consider any security aspects of the EXPLOSION, if applicable.
 
Escalation of this emergency classification level, if appropriate, would be based on HA4.
 
to 0CAN121102
 
Page 68 of 110
 
HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU5  Initiating Condition - NOTIFICATION OF UNUSUAL EVENT Release of toxic, corrosive, asphyxiant, or flammable gases deemed detrimental to NORMAL PLANT OPERATIONS.
 
Operating Mode Applicability:
All  Example Emergency Action Level(s):
(1 or 2) 
: 1. Toxic, corrosive, asphyxiant or flammable gas es 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 offsite event.
Basis:  This IC is based on the release of toxic, corrosive, asphyxiant or flammable gases of sufficient
 
quantity to affect NORMAL PLANT OPERATIONS.
 
The fact that SCBAs may be worn does not eliminate the need to declare the event.
 
This IC 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 HA5.
 
to 0CAN121102
 
Page 69 of 110
 
HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU6  Initiating Condition - NOTIFICATION OF UNUSUAL EVENT Natural or destructive phenomena affecting the PROTECTED AREA
 
Operating Mode Applicability:
All  Example Emergency Action Level: 
(1 or 2 or 3 or 4 or 5 or 6)
: 1. Seismic event identified by any 2 of the following:  Seismic event confirmed by annunciation of the 0.01g acceleration alarm  Earthquake felt in plant  National Earthquake Center OR  2. Tornado striking within PROTECTED AREA boundary or high winds > 67 mph.
 
OR  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 structures or areas in Table H1.
to 0CAN121102
 
Page 70 of 110
 
HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU6    Table H1 Unit 1 Unit 2 CA-1 & HP Office Area Condensate Demineralizer Room
 
Corridor 98
 
Fire Area C
 
Lower North Electrical Penetration Room (LNEPR) Lower South Electrical Equipment Room (LSEER)
/ Air Compressor Room Lower South Electrical Penetration Room (LSEPR) Lower South Piping Penetration Room (LSPPR)
 
Main Steam Isolation Violation (MSIV) Room
 
North Engineered Safeguards (ES) SWGR Room (A4) South ES SWGR Room
 
Turbine Building  A1, A2, H1, H2 SWGR area  354' Bowling Alley north end west of Breathing Air compressor room  368' West Heater Deck from LSEER (orange door) along east wall of ES SWGR Rooms to
 
Corridor 98 door.
Upper North Electrical Penetration Room (UNEPR) / Hot Tool Room / Decon Room Upper South Electrical Penetration Room (USEPR) Upper South Piping Penetration Room (USPPR) 2A3 Room 2A4, 2D02, & East Battery Room
 
2B53 Room
 
2B63 Room
 
2B9/2B10 Room
 
2Y11/13 Equipment Room
 
Auxiliary Building 317' General Access
 
Auxiliary Building 335'
 
Auxiliary Building 354'
 
'B' Engineered Safeguards Features (ESF)
Room Corridor Behind Door 340
 
Turbine Building  2A1, 2A2, 2H1, 2H2 Area  354' West wall of Demineralizer area  368' West Heater Deck north of north Switchgear (SWGR) Room (2A3) and
 
East of LNEPR Intake Structure  354' or 366'
 
LNEPR LSEPR Motor-Generator (MG) Set Room
 
Steam Pipe Area
 
Hot Machine Shop
 
UNEPR, UNPPR, LNPPR, USPPR OR  4. Turbine failure resulting in casing penetration or damage to turbine or generator seals.
 
OR  5. Lake Dardanelle level < 335 feet.
 
OR  6. Lake Dardanelle level > 345 feet. to 0CAN121102
 
Page 71 of 110
 
HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU6  Basis:  These EALs are categorized on the basis of the occurrence of an event of sufficient magnitude
 
to be of concern to plant operators.
 
EAL #1 Damage may be caused to some portions of the site, but should not affect ability of safety
 
functions to operate.
 
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 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 Center can confirm if an earthquake has occurred in the area of the
 
plant.
 
EAL #2 This EAL is based on a tornado striking (touching down) or high winds within the PROTECTED
 
AREA.
 
The high wind value in EAL #2 is conservatively based on the SAR design basis for Unit 1 of
 
67 mph. Unit 2 Design basis is 80 mph.
 
Escalation of this emergency classification level, if appropriate, would be based on VISIBLE
 
DAMAGE, or by other in plant conditions, via HA6.
 
EAL #3 This EAL addresses the effect of internal flooding caused by events such as component
 
failures, equipment misalignment, or outage activity mishaps.
 
Escalation of this emergency classification level, if appropriate, would be via HA6, or by other
 
plant conditions.
 
EAL #4 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 HU4 and HU5. to 0CAN121102
 
Page 72 of 110
 
HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU6  This EAL is consistent with the definition of an NUE while maintaining the anticipatory nature desired and recognizing the risk to non-safety related equipment.
 
Escalation of this emergency classification level, if appropriate, would be to HA6 based on
 
damage done by PROJECTILES generated by the failure or in conjunction with a steam
 
generator tube rupture. These latter events would be classified by the radiological (A) ICs or
 
Fission Product Barrier (F) ICs.
 
EALs #5 and #6
 
EALs #5 and #6 are based on the levels of Lake Dardanelle at which the site will take specific
 
action to reduce the impact of the lake level on plant safety by initiating plant shutdown.
 
Reference Documents
:  1. OP-1203.025, "Natural Emergencies"
: 2. OP-2203.008, "Natural Emergencies"
: 3. Unit 1 FSAR
: 4. Unit 2 FSAR
 
to 0CAN121102
 
Page 73 of 110
 
HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HA1  Initiating Condition - ALERT HOSTILE ACTION within the OWNER CONTROLLED AREA or airborne attack threat
 
Operating Mode Applicability:
All  Example Emergency Action Level(s): 
(1 or 2) 
: 1. A HOSTILE ACTION is occurring or has occurred within the OWNER CONTROLLED AREA as reported by ANO Security Shift Supervision.
OR  2. A validated notification from NRC of an airliner attack threat within 30 minutes of the site.
 
Basis:  NOTE: Timely and accurate communication between Security Shift Supervision and the Control Room is crucial for the implem entation 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, hunt ers, or physical disputes between employees within the OWNER CONTROLLED AREA. 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 Independent Spent Fuel Storage
 
Installations that may be outside the PROTECTED AREA but still in the OWNER
 
CONTROLLED AREA. to 0CAN121102
 
Page 74 of 110
 
HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HA1  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. Only the plant to which the specific threat is
 
made need declare the Alert.
 
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.
 
to 0CAN121102
 
Page 75 of 110
 
HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HA2  Initiating Condition - ALERT Other conditions exist which in the judgment of the SM / TSC Director / EOF Director warrant
 
declaration of an Alert
 
Operating Mode Applicability:
All  Example Emergency Action Level(s):
: 1. Other conditions exist which in the judgment of the SM / TSC Director / EOF Director indicate that events are in progress or have occurred which involve an actual or potential
 
substantial degradation of the level of safety of the plant or a security event that involves
 
probable life threatening risk to site personnel or damage to site equipment because of
 
HOSTILE ACTION. Any releases are expected to be limited to small fractions of the EPA
 
Protective Action Guideline exposure levels.
Basis:  This EAL addresses unanticipated conditions not addressed explicitly elsewhere, but that
 
warrant declaration of an emergency because conditions exist which are believed by the SM /
 
TSC Director / EOF Director to fall under the Alert emergency classification level.
 
to 0CAN121102
 
Page 76 of 110
 
HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HA3  Initiating Condition - ALERT Control room evacuation has been initiated
 
Operating Mode Applicability:
All  Example Emergency Action Level(s):
: 1. Alternate Shutdown procedure requires Control Room evacuation:
 
Unit 1: 1203.002, "Alternate Shutdown" Unit 2: 2203.014, "Alternate Shutdown" Basis:  With the Control Room evacuated, additional support, monitoring and direction through the
 
Technical Support Center and/or other emergency response facilities may be necessary.
 
Inability to establish plant control from outside the Control Room will escalate this event to a
 
Site Area Emergency.
 
to 0CAN121102
 
Page 77 of 110
 
HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HA4  Initiating Condition - ALERT FIRE or EXPLOSION affecting the operability of plant safety systems required to establish or
 
maintain safe shutdown
 
Operating Mode Applicability:
All  Example Emergency Action Level(s):
: 1. FIRE or EXPLOSION resulting in VISIBLE DAMAGE to any Table H1 structure or area containing safety systems or components or Control Room indication of degraded performance of those safety systems.
Table H1 Unit 1 Unit 2 CA-1 & HP Office Area Condensate Demineralizer Room
 
Corridor 98
 
Fire Area C
 
Lower North Electrical Penetration Room (LNEPR) Lower South Electrical Equipment Room (LSEER)
/ Air Compressor Room Lower South Electrical Penetration Room (LSEPR) Lower South Piping Penetration Room (LSPPR)
 
Main Steam Isolation Violation (MSIV) Room
 
North Engineered Safeguards (ES) SWGR Room (A4) South ES SWGR Room
 
Turbine Building  A1, A2, H1, H2 SWGR area  354' Bowling Alley north end west of Breathing Air compressor room  368' West Heater Deck from LSEER (orange door) along east wall of ES SWGR Rooms to
 
Corridor 98 door.
Upper North Electrical Penetration Room (UNEPR) / Hot Tool Room / Decon Room Upper South Electrical Penetration Room (USEPR) Upper South Piping Penetration Room (USPPR) 2A3 Room 2A4, 2D02, & East Battery Room
 
2B53 Room
 
2B63 Room
 
2B9/2B10 Room
 
2Y11/13 Equipment Room
 
Auxiliary Building 317' General Access
 
Auxiliary Building 335'
 
Auxiliary Building 354'
 
'B' Engineered Safeguards Features (ESF)
Room Corridor Behind Door 340
 
Turbine Building  2A1, 2A2, 2H1, 2H2 Area  354' West wall of Demineralizer area  368' West Heater Deck north of north Switchgear (SWGR) Room (2A3) and
 
East of LNEPR Intake Structure  354' or 366'
 
LNEPR LSEPR Motor-Generator (MG) Set Room
 
Steam Pipe Area
 
Hot Machine Shop
 
UNEPR, UNPPR, LNPPR, USPPR to 0CAN121102
 
Page 78 of 110
 
HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HA4  Basis:  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 or areas containi ng 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 SM/TSC Director/EOF Director with the resources needed to perform detailed damage assessments.
 
The SM / TSC Director / EOF Director also needs to consider any security aspects of the EXPLOSION.
 
Escalation of this emergency classification level, if appropriate, will be based on System
 
Malfunction (S), Fission Product Barrier Degradation (F) or Abnormal Radiation Levels /
 
Radiological Effluent (A) ICs.
 
to 0CAN121102
 
Page 79 of 110
 
HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HA5  Initiating Condition - ALERT 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 Mode Applicability:
All  Example Emergency Action Level(s):
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.
Basis:  Gases in a VITAL AREA can affect the ability to safely operate or safely shutdown the reactor.
 
The fact that SCBAs may be worn does not eliminate the need to declare the event.
Declaration should not be delayed for confirmation from atmospheric testing if the atmosphere
 
poses an immediate threat to life and health or an immediate threat of severe exposure to
 
gases. This could be based upon documented analysis, indication of personal ill effects from
 
exposure, or operating experience with the hazards.
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 shut down beyond that already allowed by Technical Specifications at the time of the event.
An asphyxiant is a gas capable of reducing the level of oxygen in the body to dangerous levels.
 
Most commonly, asphyxiants work by merely displacing air in an enclosed environment. This
 
reduces the concentration of oxygen below the normal level of around 19%, which can lead to
 
breathing difficulties, unconsciousness or even death.
An uncontrolled release of flammable gasses within a facility structure has the potential to affect
 
safe operation of the plant by limiting either operator or equipment operations due to the
 
potential for ignition and resulting equipment damage/personnel injury. Flammable gasses, such as hydrogen and acetylene, are routinely used to maintain plant systems (hydrogen) or to
 
repair equipment/components (acetylene - used in welding). This EAL assumes concentrations
 
of flammable gasses which can ignite/support combustion.
Escalation of this emergency classification level, if appropriate, will be based on System
 
Malfunction (S), Fission Product Barrier Degradation (F) or Abnormal Radiation Levels /
 
Radioactive Effluent (A) ICs.
 
. to 0CAN121102
 
Page 80 of 110
 
HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HA6  Initiating Condition - ALERT Natural or destructive phenomena affecting VITAL AREAS
 
Operating Mode Applicability:
All  Example Emergency Action Level(s):
  (1 or 2 or 3 or 4 or 5 or 6)
: 1. a. Seismic event > Operating Basis Earthquake (OBE) as indicated by annunciation of the 0.1g acceleration alarm.
AND  b. Earthquake confirmed by any of the following:
Earthquake felt in plant  National Earthquake Center  Control Room indication of degraded perform ance of systems required for the safe shutdown of the plant OR  2. Tornado striking or high winds > 67 mph resulting in VISIBLE DAMAGE to any of the following structures/equipment contai ning safety systems or components or Control Room indication of degraded performance of those safety systems:
Reactor Building Turbine Building Intake Structure Q Condensate Storage Tank (QCST)
Ultimate Heat Sink Control Room Startup Transformers Auxiliary Building Diesel Fuel Vault Borated Water Storage Tank (BWST)
 
Refueling Water Tank (RWT)
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:
Intake Structure Turbine Building Ultimate Heat Sink Control Room BWST / RWT Startup Transformers Auxiliary Building Diesel Fuel Vault
 
QCST  OR    to 0CAN121102
 
Page 81 of 110
 
HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HA6  4. Turbine failure-generated PROJECTILES resulting in VISIBLE DAMAGE to or penetration of any of the structures/equipment in Table H2 containing safety systems or components or Control Room indication of degraded perfo rmance of those safety systems:
Table H2 Reactor Building Turbine Building Intake Structure QCST Ultimate Heat Sink Control Room BWST/RWT Startup Transformers Auxiliary Building Diesel Fuel Vault OR  5. Lake Dardanelle level < 335 feet and Emergency Cooling Pond inoperable.
 
OR  6. Vehicle crash resulting in VISIBLE DAMAGE to any of the structures/equipment in Table H2 containing safety systems or com ponents or Control Room indication of degraded performance of those safety systems:
Table H2 Reactor Building Turbine Building Intake Structure QCST Ultimate Heat Sink Control Room BWST/RWT Startup Transformers Auxiliary Building Diesel Fuel Vault Basis:  These EALs escalate from HU6 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 t hose structures evidenced by Control Room indications of degraded system response or performance. The occurrence of VISIBLE
 
DAMAGE and/or degraded system response is int ended 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.
 
Escalation of this emergency classification level, if appropriate, would be based on System
 
Malfunction (S) ICs.
 
to 0CAN121102
 
Page 82 of 110
 
HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HA6  EAL #1 Seismic events of this magnitude 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.
 
The National Earthquake Center can confirm if an earthquake has occurred in the area of the
 
plant.
 
EAL #2 This EAL is based on a tornado striking (touching down) or high winds that have caused
 
VISIBLE DAMAGE to structures containing functions or systems required for safe shutdown of
 
the plant. The high wind value in EAL #2 is conservatively based on the SAR design basis for
 
Unit 1 of 67 mph. Unit 2 Design basis is 80 mph.
 
EAL #3 This EAL addresses the effect of internal flooding caused by events such as component
 
failures, equipment misalignment, or outage activity mishaps. It is based on the degraded
 
performance of systems, or has created industrial safety hazards (e.g., electrical shock) that preclude necessary access to operate or monitor safety equipment. The inability to access, operate or monitor safety equipment represents an actual or substantial potential degradation of
 
the level of safety of the plant.
 
Flooding as used in this EAL describes a condition where water is entering the room faster than
 
installed equipment is capable of removal, resulting in a rise of water level within the room. 
 
Classification of this EAL should not be delayed while corrective actions are being taken to
 
isolate the water source.
 
EAL #4 This EAL 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 #5 This EAL addresses vehicle crashes within the PROTECTED AREA that result in VISIBLE
 
DAMAGE to VITAL AREAS or indication of damage to safety structures, systems, or
 
components containing functions and systems required for safe shutdown of the plant.
 
EAL #6 EAL #6 addresses site specific phenomena which has the potential for the loss of primary and
 
secondary heat sink.
 
to 0CAN121102
 
Page 83 of 110
 
HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HA6  Reference Documents
:  1. OP-1203.025, "Natural Emergencies"
: 2. OP-2203.008, "Natural Emergencies"
: 3. Unit 1 FSAR
: 4. Unit 2 FSAR
 
to 0CAN121102
 
Page 84 of 110
 
HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HS1  Initiating Condition - SITE AREA EMERGENCY HOSTILE ACTION within the PROTECTED AREA
 
Operating Mode Applicability:
All  Example Emergency Action Level(s):
: 1. A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by ANO Security Shift Supervision.
Basis:  This condition represents an escalated threat to plant safety above that contained in the Alert in
 
that a HOSTILE FORCE has progressed from the OWNER CONTROLLED AREA to the
 
PROTECTED AREA.
 
This EAL addresses the contingency for a very rapid progression of events, such as that
 
experienced on September 11, 2001. It is not prem ised 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 Organization readiness and
 
preparation for the implementation of protective measures.
 
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, hunt ers, or physical disputes between employees within the PROTECTED AREA. Those events are adequately addressed by other EALs.
 
Escalation of this emergency classification level, if appropriate, would be based on actual plant
 
status after impact or progression of attack.
 
to 0CAN121102
 
Page 85 of 110
 
HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HS2  Initiating Condition - SITE AREA EMERGENCY Other conditions exist which in the judgment of the SM / TSC Director / EOF Director warrant
 
declaration of a Site Area Emergency
 
Operating Mode Applicability:
All  Example Emergency Action Level(s):
: 1. Other conditions exist which in the judgment of the SM / TSC Director / EOF 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.
Basis:  This EAL addresses unanticipated conditions not addressed explicitly elsewhere but that
 
warrant declaration of an emergency because conditions exist which are believed by the SM /
 
TSC Director / EOF Director to fall under the emergency classification level description for Site
 
Area Emergency.
 
to 0CAN121102
 
Page 86 of 110
 
HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HS3  Initiating Condition - SITE AREA EMERGENCY Control Room evacuation has been initiated and plant control cannot be established
 
Operating Mode Applicability:
All  Example Emergency Action Level(s):
: 1. a. Control room evacuation has been initiated
 
AND  b. Control of the plant cannot be established in accordance with the following procedures within 15 minutes:
Unit 1: 1203.002, "Alternate Shutdown" Unit 2: 2203.014, "Alternate Shutdown" Basis:  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 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
 
such as reactivity control (ability to shutdown the reactor and maintain it shutdown), RCS
 
inventory (ability to cool the core), and decay heat removal (ability to maintain a heat sink).
 
The determination of whether or not control is established at the remote shutdown panel is
 
based on SM / TSC Director / EOF Director judgment. The SM / TSC Director / EOF Director is
 
expected to make a reasonable, informed judgment within 15 minutes that the plant staff has
 
control of the plant from the remote shutdown panel.
 
Escalation of this emergency classification level, if appropriate, would be by Fission Product
 
Barrier Degradation (F) or Abnormal Radiation Levels/Radiological Effluent (A) EALs.
 
to 0CAN121102
 
Page 87 of 110
 
HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HG1  Initiating Condition - GENERAL EMERGENCY HOSTILE ACTION resulting in loss of physical control of the facility
 
Operating Mode Applicability:
All  Example Emergency Action Level(s):
  (1 or 2)
: 1. A HOSTILE ACTION has occurred such that plant personnel are unable to operate equipment required to maintain safety functions.
OR  2. A HOSTILE ACTION has caused failure of Spent Fuel Cooling Systems and IMMINENT fuel damage is likely for a freshly off-loaded reactor core in pool.
Basis:  EAL #1 This EAL encompasses conditions under which a HOSTILE ACTION has resulted in a loss of
 
physical control of VITAL AREAS (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. These safety functions are reactivity control (ability to shut
 
down the reactor and keep it shutdown) RCS inventory (ability to cool the core), and secondary
 
heat removal (ability to maintain a heat sink).
 
Loss of physical control of the Control Room or remote shutdown/alternate shutdown capability
 
alone may not prevent the ability to maintain safety functions per se. Design of the remote
 
shutdown/alternate 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 threshold is not met.
 
EAL #2 This EAL addresses failure of spent fuel 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. At ANO, the term "freshly off-loaded reactor core" refers to fuel that has been
 
discharged from the core and stored in the spent fuel pool for a period of LESS THAN one year.
 
to 0CAN121102
 
Page 88 of 110
 
HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HG2  Initiating Condition - GENERAL EMERGENCY Other conditions exist which in the judgment of the SM / TSC Director / EOF Director warrant
 
declaration of a General Emergency
 
Operating Mode Applicability:
All  Example Emergency Action Level(s):
: 1. Other conditions exist which in the judgment of the SM / TSC Director / EOF Director indicate that events are in progress or have occurred which involve actual or IMMINENT
 
substantial core degradation or melting with potential for loss of containment integrity or
 
HOSTILE ACTION that results in an actual loss of physical control of the facility. Releases
 
can be reasonably expected to exceed EPA Protec tive Action Guideline exposure levels offsite for more than the immediate site area.
Basis:  This EAL addresses unanticipated conditions not addressed explicitly elsewhere but that
 
warrant declaration of an emergency because conditions exist which are believed by the SM /
 
TSC Director / EOF Director to fall under the emergency classification level description for
 
General Emergency.
 
to 0CAN121102
 
Page 89 of 110
 
SYSTEM MALFUNCTION    to 0CAN121102
 
Page 90 of 110
 
SYSTEM MALFUNCTION SU1  Initiating Condition - NOTIFICATION OF UNUSUAL EVENT Loss of all offsite AC power to Vital 4.16 KV busses  15 minutes Operating Mode Applicability: Power Operations (Mode 1)  Startup (Mode 2)
Hot Standby (Mode 3)
Hot Shutdown (Mode 4)
 
Example Emergency Action Level(s):
Note: The SM 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 offsite AC power to Vital 4.16 KV busses  15 minutes.
Basis:  Prolonged loss of offsite 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 to emergency busses.
 
Fifteen minutes was selected as a threshold to exclude transient or momentary losses of off-site
 
power.
 
Reference Documents
:  1. 1202.007, "Degraded Power"
: 2. 1202.008, "Blackout"
: 3. 2202.007, "Loss of Off-Site Power"
: 4. 2202.008, "Station Blackout"
 
to 0CAN121102
 
Page 91 of 110
 
SYSTEM MALFUNCTION SU6  Initiating Condition - NOTIFICATION OF UNUSUAL EVENT UNPLANNED loss of safety system annunciati on or indication in the Control Room  15 minutes Operating Mode Applicability: Power Operations (Mode 1)  Startup (Mode 2)
Hot Standby (Mode 3)
Hot Shutdown (Mode 4)
 
Example Emergency Action Level(s):
Note: The SM 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 > approximately 75% of the following > 15 minutes:
: a. Control Room annunciators associated with safety systems.
OR  b. Control Room safety system indication.
Basis:  This IC and its associated EAL are intended to recognize the difficulty associated with
 
monitoring changing plant conditions without the use of a major portion of the annunciation or
 
indication equipment.
 
Recognition of the availability of computer based indication equipment is considered e.g.,
SPDS, plant computer, etc.
 
"Planned" loss of annunciators or indicators includes scheduled maintenance and testing
 
activities.
 
Quantification is arbitrary, however, it is esti mated that if approximately 75% of the safety system annunciators or indicators are lost, there is an increased risk that a degraded plant
 
condition could go undetected. It is not intended that plant personnel perform a detailed count
 
of the instrumentation lost but use the value as a judgment threshold for determining the
 
severity of the plant conditions.
 
It is further recognized that most plant designs provide redundant safety system indication
 
powered from separate uninterruptible power supplies. While failure of a large portion of
 
annunciators is more likely than a failure of a large portion of indications, the concern is
 
included in this EAL due to difficulty associated with assessment of plant conditions. The loss
 
of specific, or several, safety system indicators should remain a function of that specific system or component operability status. This will be addressed by the specific Technical Specification. to 0CAN121102
 
Page 92 of 110
 
SYSTEM MALFUNCTION SU6  The initiation of a Technical Specification imposed plant shutdown related to the instrument loss will be reported via 10 CFR 50.72. If the shutdown is not in compliance with the Technical
 
Specification action, the NUE is based on SU11 "Inability to reach required operating mode
 
within Technical Specification limits."
 
Indicators associated with safety systems are those indicators for reactivity control, core
 
cooling, maintaining reactor coolant system integrity or maintaining containment integrity.
 
Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.
 
This NUE will be escalated to an Alert based on a concurrent loss of compensatory indications
 
or if a SIGNIFICANT TRANSIENT is in progress during the loss of annunciation or indication (SA6).
 
Reference Documents
:  1. 1203.043, "Loss Control Room Annunciators"
: 2. 2203.042, "Loss of Control Room Annunciators"
 
to 0CAN121102
 
Page 93 of 110
 
SYSTEM MALFUNCTION SU7  Initiating Condition - NOTIFICATION OF UNUSUAL EVENT RCS leakage
 
Operating Mode Applicability: Power Operations (Mode 1)  Startup (Mode 2)
Hot Standby (Mode 3)
Hot Shutdown (Mode 4)
 
Example Emergency Action Level(s):
  (1 or 2)
: 1. Unidentified or pressure boundary leakage > 10 gpm.
 
OR  2. Identified leakage > 25 gpm.
 
Basis:  With respect to this IC, RCS leakage is defined as a loss of RCS inventory due to a leak in the
 
RCS or a supporting system that is not or cannot be isolated within 10 minutes. For example, isolation of the RCS Letdown (purification) system is a standard abnormal operating procedure
 
action and may prevent unnecessary classifications when a non-RCS leakage path leak exists. 
 
However, the intent of this condition is met if attempts to isolate the RCS leak are NOT
 
successful.
 
This IC is included as an NUE 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. The
 
10 gpm value for the unidentified or pressure boundary leakage was selected as it is observable
 
with normal Control Room indications. Lesser values must generally be determined through
 
time-consuming surveillance tests (e.g., mass balances).
 
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 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. Steam generator tube
 
leakage is identified leakage. In either case, escalation of this IC to the Alert level is via Fission
 
Product Barrier Degradation (F) ICs.
 
to 0CAN121102
 
Page 94 of 110
 
SYSTEM MALFUNCTION SU8  Initiating Condition - NOTIFICATION OF UNUSUAL EVENT Loss of all onsite or offsite communications capabilities
 
Operating Mode Applicability: Power Operations (Mode 1)  Startup (Mode 2)
Hot Standby (Mode 3)
Hot Shutdown (Mode 4)
 
Example Emergency Action Level(s):
  (1 or 2)
: 1. Loss of all Table M1 onsite communications methods affecting the ability to perform routine operations.
OR  2. Loss of all Table M2 offsite communications methods affecting the ability to perform offsite notifications.
Table M1 Onsite Communications Methods  Table M2 Offsite Communications Methods Station radio system  All telephone lines (commercial and microwave) Plant paging system  ENS In-plant telephones  Gaitronics Basis:  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 issues with offsite authorities.
 
The availability of one method of ordinary offsite communications is sufficient to inform federal, state, and local authorities of plant problems. This EAL is intended to be used only when
 
extraordinary means (e.g., relaying of information from non-routine radio transmissions, individuals being sent to off-site locations, etc.) are being used to make communications
 
possible.
 
Reference Documents
:  1. 1903.062, "Communications System Operating Procedure"
 
to 0CAN121102
 
Page 95 of 110
 
SYSTEM MALFUNCTION SU9  Initiating Condition - NOTIFICATION OF UNUSUAL EVENT Fuel clad degradation
 
Operating Mode Applicability: Power Operations (Mode 1)  Startup (Mode 2)
Hot Standby (Mode 3)
Hot Shutdown (Mode 4)
 
Example Emergency Action Level(s):
  (1 or 2)
: 1. Failed Fuel Iodine radiation monitor reading indicates fuel clad degradation > Technical Specification allowable limits:
Unit 1: RI-1237S reads > 1.3 x 10 5 counts per minute Unit 2: 2RITS-4806B reads > .65 x 10 5 counts per minute OR  2. RCS sample activity value indicating fuel clad degradation > Technical Specification allowable limits:
uCi/gm Dose Equivalent I-131 for more than 48 hours OR  Unit 1:  60 uCi/gm Dose Equivalent I-131 Unit 2: > 60 uCi/gm Dose Equivalent I-131 OR  Unit 1: > 2200 &#xb5;Ci/gm Dose Equivalent Xe-133 for more than 48 hours Unit 2: > 3100 &#xb5;Ci/gm Dose Equivalent Xe-133 for more than 48 hours    to 0CAN121102
 
Page 96 of 110
 
SYSTEM MALFUNCTION SU9  Basis:  This IC is included because it is a precursor of more serious conditions and, as result, is
 
considered to be a potential degradation of the level of safety of the plant.
 
EAL #1 This threshold addresses the Letdown Radiation Monitor readings that provide indication of a
 
degradation of fuel clad integrity.
 
EAL #2 This EAL addresses coolant samples exceeding coolant technical specifications for transient
 
iodine spiking limits and coolant samples exceeding coolant Technical Specifications for
 
nominal operating limits for the time period specified in the Technical Specifications.
 
Escalation of this IC to the Alert level is via the Fission Product Barriers (F).
 
Reference Documents
:  1. ANO1 Technical Specifications
: 2. ANO2 Technical Specifications
 
to 0CAN121102
 
Page 97 of 110
 
SYSTEM MALFUNCTION SU10  Initiating Condition - NOTIFICATION OF UNUSUAL EVENT Inadvertent criticality
 
Operating Mode Applicability: Hot Standby (Mode 3)  Hot Shutdown (Mode 4)
 
Example Emergency Action Level(s):
: 1. UNPLANNED sustained positive startup rate observed on nuclear instrumentation.
 
Basis:  This IC addresses inadvertent criticality events. This IC indicates a potential degradation of the
 
level of safety of the plant, warranting an NUE classification. This IC excludes inadvertent
 
criticalities that occur during planned reactivity changes associated with reactor startups (e.g.,
criticality earlier than estimated).
 
This condition can be identified using the startup rate meter. The term "sustained" is used in
 
order to allow exclusion of expected short term positive startup rates from planned control rod movements for (such as shutdown bank withdrawal). These short term positive startup rates
 
are the result of the rise in neutron population due to subcritical multiplication.
 
Escalation would be by the Fission Product Barrier Table (F), as appropriate to the operating
 
mode at the time of the event.
 
Reference Documents
:  1. 1203.012G, "Annunciator K08 Corrective Action"
: 2. 2203.012D, "Annunciator 2K04 Corrective Action" to 0CAN121102
 
Page 98 of 110
 
SYSTEM MALFUNCTION SU11  Initiating Condition - NOTIFICATION OF UNUSUAL EVENT Inability to reach required operating mode within Technical Specification limits
 
Operating Mode Applicability: Power Operations (Mode 1)  Startup (Mode 2)
Hot Standby (Mode 3)
Hot Shutdown (Mode 4)
 
Example Emergency Action Level(s):
: 1. Plant is not brought to required operating mode within Technical Specifications LCO Action Statement time.
Basis:  Limiting Conditions of Operation (LCOs) require the plant to be brought to a required operating
 
mode when the Technical Specification required configuration cannot be restored. Depending
 
on the circumstances, this may or may not be an emergency or precursor to a more severe
 
condition. In any case, the initiation of plant shutdown required by the site Technical
 
Specifications requires a four hour report under 10 CFR 50.72 (b) Non-emergency events. The
 
plant is within its safety envelope when being shut down within the allowable action statement
 
time in the Technical Specifications. An immediate NUE is required when the plant is not
 
brought to the required operating mode within the allowable action statement time in the
 
Technical Specifications. Declaration of an NUE is based on the time at which the LCO-
 
specified action statement time period elapses under the site Technical Specifications and is not
 
related to how long a condition may have existed.
 
Reference Documents
:  1. ANO2 Technical Specifications
: 2. ANO1 Technical Specifications
 
to 0CAN121102
 
Page 99 of 110
 
SYSTEM MALFUNCTION SA1  Initiating Condition - ALERT AC power capability to Vital 4.16 KV busses reduced to a single power source  15 minutes such that any additional single failure would result in station blackout
 
Operating Mode Applicability: Power Operations (Mode 1)  Startup (Mode 2)
Hot Standby (Mode 3)
Hot Shutdown (Mode 4)
 
Example Emergency Action Level(s):
Note: The SM / TSC Director / EOF 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 Vital 4.16 KV busses reduced to a single power source  15 minutes.
AND  b. Any additional single power source failure will result in station blackout.
Basis:  The condition indicated by this IC is the degradat ion of the offsite and onsite AC power systems such that any additional single failure would result in a station blackout. This condition could
 
occur due to a loss of offsite power with a concurrent failure of all but one emergency generator
 
to supply power to its emergency busses. Another related condition could be the loss of all
 
offsite power and loss of onsite emergency generators with only one train of emergency busses
 
being backfed from the unit main generator, or the loss of onsite emergency generators with only
 
one train of emergency busses being backfed from offsite power. The subsequent loss of this single power source would escalate the event to a Site Area Emergency in accordance with SS1.
Fifteen minutes was selected as a threshold to exclude transient or momentary losses of power.
 
The EAL allows credit for operation of the Alternate AC Diesel Generator.
 
Reference Documents
:  1. 1202.007, "Degraded Power"
: 2. 1202.008, "Blackout"
: 3. 2202.007, "Loss of Off-Site Power"
: 4. 2202.008, "Station Blackout"
: 5. 2104.037, "Alternate AC Diesel Generator Operations" to 0CAN121102
 
Page 100 of 110
 
SYSTEM MALFUNCTION SA3  Initiating Condition - ALERT Automatic trip fails to shutdown the reactor and the manual actions taken from the reactor
 
control console are successful in shutting down the reactor
 
Operating Mode Applicability: Power Operations (Mode 1)  Startup (Mode 2)
 
Example Emergency Action Level(s):
: 1. a. An automatic trip failed to shutdown the reactor as indicated by reactor power  5%. AND  b. Manual actions taken at the reactor control console successfully shutdown the reactor as indicated by reactor power < 5%.
Basis:  Manual trip actions taken at the reactor control console are any set of actions by the Reactor
 
Operator(s) which causes or should cause control rods to be rapidly inserted into the core and
 
shuts down the reactor. Any action taken to trip the reactor from any location other than panel
 
C03 (Unit 1) or 2C03/2C14 (Unit 2) constitutes a failure of the manual trip function. Failure of manual trip would escalate the event to a Site Area Emergency (SS3).
This condition indicates failure of the automatic protection system to trip the reactor. This
 
condition is more than a potential degradation of a safety system in that a front line automatic
 
protection system did not function in response to a plant transient. Thus the plant safety has
 
been compromised because design limits of the fuel may have been exceeded. An Alert is
 
indicated because conditions may exist that lead to potential loss of fuel clad or RCS and
 
because of the failure of the Reactor Protection System to automatically shutdown the plant.
 
If manual actions taken at the reactor control console fail to shutdown the reactor, the event
 
would escalate to a Site Area Emergency.
 
to 0CAN121102
 
Page 101 of 110
 
SYSTEM MALFUNCTION SA6  Initiating Condition - ALERT UNPLANNED loss of safety system annunciation or i ndication in the Control Room with either (1) a SIGNIFICANT TRANSIENT in progress, or (2) compensatory indicators unavailable
 
Operating Mode Applicability: Power Operations (Mode 1)  Startup (Mode 2)
Hot Standby (Mode 3)
Hot Shutdown (Mode 4)
 
Example Emergency Action Level(s):
Note: The SM/TSC Director/EOF 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. UNPLANNED loss of > approximately 75% of the following  15 minutes:
Control Room annunciators associated with safety systems OR  Control Room safety system indication AND  b. Either of the following:
A SIGNIFICANT TRANSIENT is in progress OR  Compensatory indications are unavailable.
Basis:  This IC is intended to recognize the difficulty associated with monitoring changing plant
 
conditions without the use of a major portion of the annunciation or indication equipment during
 
a SIGNIFICANT TRANSIENT.
 
Recognition of the availability of computer based indication equipment is considered (e.g.,
SPDS, plant computer, etc.).
 
"Planned" loss of annunciators or indicators includes scheduled maintenance and testing
 
activities. to 0CAN121102
 
Page 102 of 110
 
SYSTEM MALFUNCTION SA6  Quantification is arbitrary, however, it is esti mated that if approximately 75% of the safety system annunciators or indicators are lost, there is an increased risk that a degraded plant
 
condition could go undetected. It is not intended that plant personnel perform a detailed count
 
of the instrumentation lost but use the value as a judgment threshold for determining the
 
severity of the plant conditions. It is also not intended that the Shift Manager be tasked with
 
making a judgment decision as to whether additional personnel are required to provide
 
increased monitoring of system operation.
 
It is further recognized that most plant designs provide redundant safety system indication
 
powered from separate uninterruptible power supplies. While failure of a large portion of
 
annunciators is more likely than a failure of a large portion of indications, the concern is
 
included in this EAL due to difficulty associated with assessment of plant conditions. The loss
 
of specific, or several, safety system indicators should remain a function of that specific system or component operability status. This will be addressed by the specific Technical Specification. 
 
The initiation of a Technical Specification imposed plant shutdown related to the instrument loss
 
will be reported via 10 CFR 50.72. If the shutdown is not in compliance with the Technical
 
Specification action, the NUE is based on SU11 "Inability to reach required operating mode
 
within Technical Specification limits."
 
Indicators associated with safety systems are those indicators for reactivity control, core
 
cooling, maintaining reactor coolant system integrity or maintaining containment integrity.
 
"Compensatory indications" in this context includes computer based information such as SPDS, QSPDS, COLSS, etc. If both a major portion of the annunciation system and all computer
 
monitoring are unavailable, the Alert is required.
 
Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.
 
This Alert will be escalated to a Site Area Emergency if the operating crew cannot monitor the
 
transient in progress due to a concurrent loss of compensatory indications with a SIGNIFICANT
 
TRANSIENT in progress during the loss of annunciation or indication.
 
Reference Documents
:  1. 1015.037, "Post Transient Review"
: 2. 1203.043, "Loss of Control Room Annunciators"
: 3. 2203.042, "Loss of Control Room Annunciators"
 
to 0CAN121102
 
Page 103 of 110
 
SYSTEM MALFUNCTION SS1  Initiating Condition - SITE AREA EMERGENCY Loss of all offsite and all onsite AC power to Vital 4.16 KV busses  15 minutes Operating Mode Applicability: Power Operations (Mode 1)  Startup (Mode 2)
Hot Standby (Mode 3)
Hot Shutdown (Mode 4)
 
Example Emergency Action Level(s):
Note: The SM / TSC Director / EOF 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 offsite and all onsite AC power to Vital 4.16 KV busses  15 minutes.
Basis:  Loss of all AC power to emergency busses comp romises all plant safety systems requiring electric power including Shutdown Cooling, ECCS, Containment Heat Removal and the Ultimate
 
Heat Sink. Prolonged loss of all AC power to emergency busses will lead to loss of Fuel Clad, RCS, and Containment, thus this event can escalate to a General Emergency.
 
Fifteen minutes was selected as a threshold to exclude transient or momentary losses of offsite
 
power.
 
Escalation to General Emergency is via Fission Product Barrier Degradation (F) or IC SG1, "Prolonged loss of all offsite and all onsite AC power to Vital 4.16 KV busses."
 
Reference Documents
:  1. 1202.007, "Degraded Power"
: 2. 1202.008, "Blackout"
: 3. 2202.007, "Loss of Off-Site Power"
: 4. 2202.008, "Station Blackout"
: 5. 2104.037, "Alternate AC Diesel Generator Operations"
 
to 0CAN121102
 
Page 104 of 110
 
SYSTEM MALFUNCTION SS3  Initiating Condition - SITE AREA EMERGENCY Automatic trip fails to shutdown the reactor and manual actions taken from the reactor control
 
console are not successful in shutting down the reactor
 
Operating Mode Applicability: Power Operations (Mode 1)  Startup (Mode 2)
 
Example Emergency Action Level(s):
: 1. a. An automatic trip failed to shutdown the reactor.
 
AND  b. Manual actions taken at the reactor control console do not shutdown the reactor as indicated by reactor power  5%. Basis:  Under these conditions, the reactor is producing more heat than the maximum decay heat load
 
for which the safety systems are designed and efforts to bring the reactor subcritical are
 
unsuccessful. A Site Area Emergency is warranted because conditions exist that lead to
 
IMMINENT loss or potential loss of both fuel clad and RCS.
 
Manual trip actions taken at the reactor control console are any set of actions by the Reactor
 
Operator(s) which causes or should cause control rods to be rapidly inserted into the core and
 
shuts down the reactor.
 
Manual trip actions are not considered successful if action away from panel C03 (Unit 1) or panels 2C03/2C14 (Unit 2) is required to trip the reactor. This EAL is still applicable even if
 
actions taken away from panel C03 (Unit 1) or panels 2C03/2C14 (Unit 2) are successful in shutting the reactor down because the design limits of the fuel may have been exceeded or
 
because of the gross failure of the Reactor Protection System to shutdown the plant.
 
Escalation of this event to a General Emergency would be due to a prolonged condition leading
 
to an extreme challenge to either core-cooling or heat removal.
 
to 0CAN121102
 
Page 105 of 110
 
SYSTEM MALFUNCTION SS4  Initiating Condition - SITE AREA EMERGENCY Loss of all vital DC power  15 minutes Operating Mode Applicability: Power Operations (Mode 1)  Startup (Mode 2)
Hot Standby (Mode 3)
Hot Shutdown (Mode 4)
 
Example Emergency Action Level(s):
Note: The SM / TSC Director / EOF 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. < 105 volts on all Vital DC busses  15 minutes.
Basis:  Loss of all DC power compromises ability to monitor and control plant safety functions. 
 
Prolonged loss of all DC power will cause core uncovering and loss of containment integrity
 
when there is significant decay heat and sensible heat in the reactor system.
 
Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.
 
Escalation to a General Emergency would occu r by Abnormal Radiation Levels/Radiological Effluent (A), Fission Product Barrier Degradation (F).
 
to 0CAN121102
 
Page 106 of 110
 
SYSTEM MALFUNCTION SS6  Initiating Condition - SITE AREA EMERGENCY Inability to monitor a SIGNIFICANT TRANSIENT in progress
 
Operating Mode Applicability: Power Operations (Mode 1)  Startup (Mode 2)
Hot Standby (Mode 3)
Hot Shutdown (Mode 4)
 
Example Emergency Action Level(s):
Note: The SM / TSC Director / EOF 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 > approximately 75% of the following  15 minutes:
Control Room annunciators associated with safety systems OR  Control Room safety system indication AND  b. A SIGNIFICANT TRANSIENT is in progress.
AND  c. Compensatory indications are unavailable.
Basis:  This IC is intended to recognize the threat to plant safety associated with the complete loss of
 
capability of the control room staff to monitor plant response to a SIGNIFICANT TRANSIENT.
 
"Planned" and "UNPLANNED" actions are not differentiated since the loss of instrumentation of
 
this magnitude is of such significance during a transient that the cause of the loss is not an
 
ameliorating factor.
 
Quantification is arbitrary, however, it is esti mated that if approximately 75% of the safety system annunciators or indicators are lost, there is an increased risk that a degraded plant
 
condition could go undetected. It is not intended that plant personnel perform a detailed count
 
of the instrumentation lost but use the value as a judgment threshold for determining the
 
severity of the plant conditions. It is also not intended that the Shift Manager be tasked with
 
making a judgment decision as to whether additional personnel are required to provide
 
increased monitoring of system operation. to 0CAN121102
 
Page 107 of 110
 
SYSTEM MALFUNCTION SS6  It is further recognized that most plant designs provide redundant safety system indication powered from separate uninterruptible power supplies. While failure of a large portion of
 
annunciators is more likely than a failure of a large portion of indications, the concern is
 
included in this EAL due to difficulty associated with assessment of plant conditions. The loss
 
of specific, or several, safety system indicators should remain a function of that specific system or component operability status. This will be addressed by the specific Technical Specification. 
 
The initiation of a Technical Specification imposed plant shutdown related to the instrument loss
 
will be reported via 10 CFR 50.72. If the shutdown is not in compliance with the Technical
 
Specification action, the NUE is based on SU11 "Inability to reach required operating mode
 
within Technical Specification limits."
 
A Site Area Emergency is considered to exist if the Control Room staff cannot monitor safety functions needed for protection of the public while a significant transient is in progress.
 
Site specific indications needed to monitor safety functions necessary for protection of the
 
public must include Control Room indications, computer generated indications and dedicated
 
annunciation capability.
 
Indicators associated with safety systems are those indicators for reactivity control, core
 
cooling, maintaining reactor coolant system integrity or maintaining containment integrity.
 
"Compensatory indications" in this context includes computer based information such as SPDS, QSPDS, COLSS, etc. This should include all computer systems available for this use depending on specific plant design and subsequent retrofits.
 
Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.
 
Reference Documents
:  1. 015.037, "Post Transient Review"
: 2. 1203.043, "Loss of Control Room Annunciators"
: 3. 2203.042, "Loss of Control Room Annunciators" to 0CAN121102
 
Page 108 of 110
 
SYSTEM MALFUNCTION SG1  Initiating Condition - GENERAL EMERGENCY Prolonged loss of all offsite and all onsite AC power to safety busses
 
Operating Mode Applicability: Power Operations (Mode 1)  Startup (Mode 2)
Hot Standby (Mode 3)
Hot Shutdown (Mode 4)
 
Example Emergency Action Level(s):
: 1. a. Loss of all offsite and all onsite AC power to safety busses.
 
AND  b. Either of the following:
Restoration of at least one safety bus in < 4 hours is not likely.
OR  Continuing degradation of core cooling based on Fission Product Barrier monitoring as indicated by CETs  700 &deg;F. Basis:  Loss of all AC power to emergency busses comp romises all plant safety systems requiring electric power including Shutdown Cooling, ECCS, Containment Heat Removal and the Ultimate
 
Heat Sink. Prolonged loss of all AC power to emergency busses will lead to loss of fuel clad, RCS, and containment, thus warranting declaration of a General Emergency.
 
This IC is specified to assure that in the unlikely event of a prolonged station blackout, timely
 
recognition of the seriousness of the event occurs and that declaration of a General Emergency
 
occurs as early as is appropriate, based on a reasonable assessment of the event trajectory.
 
The likelihood of restoring at least one emergency bus should be based on a realistic appraisal
 
of the situation since a delay in an upgrade decision based on only a chance of mitigating the
 
event could result in a loss of valuable time in preparing and implementing public protective
 
actions.
 
In addition, under these conditions, fission product barrier monitoring capability may be
 
degraded.
 
to 0CAN121102
 
Page 109 of 110
 
SYSTEM MALFUNCTION SG1  Although it may be difficult to predict when power can be restored, it is necessary to give the SM / TSC Director / EOF Director a reasonable idea of how quickly (s)he may need to declare a
 
General Emergency based on two major considerations:
: 1. Are there any present indications that core cooling is already degraded to the point that loss or potential loss of Fission Product Barriers is IMMINENT?
: 2. If there are no present indications of such core cooling degradation, how likely is it that power can be restored in time to assure that a loss of two barriers with a potential loss of
 
the third barrier can be prevented?
 
Thus, indication of continuing core cooling degradation must be based on Fission Product
 
Barrier monitoring with particular emphasis on SM / TSC Director / EOF Director judgment as it
 
relates to IMMINENT loss or potential loss of fission product barriers and degraded ability to
 
monitor fission product barriers.
 
Reference Documents
:  1. Unit 1 Calculation 85-E-0072-02, "Time from Loss of All AC Power to Loss of Subcooling"
: 2. Unit 2 Calculation 85-E-0072-01, "Time from Loss of All AC Power to Loss of Subcooling"
 
to 0CAN121102
 
Page 110 of 110
 
SYSTEM MALFUNCTION SG3  Initiating Condition - GENERAL EMERGENCY Automatic trip and all manual actions fail to shutdown the reactor and indication of an extreme
 
challenge to the ability to cool the core exists
 
Operating Mode Applicability: Power Operations (Mode 1)  Startup (Mode 2)
 
Example Emergency Action Level(s):
: 1. a. An automatic trip failed to shutdown the reactor
 
AND  b. All manual actions do not shutdown the reactor as indicated by reactor power  5%. AND  c. Either of the following exist or have occurred due to continued power generation:
CET temperatures at or approaching 1200 &deg;F OR  Feedwater flow rate less than:
Unit 1: 430 gpm Unit 2: 485 gpm  Basis:  Under these conditions, the reactor is producing more heat than the maximum decay heat load
 
for which the safety systems are designed and efforts to bring the reactor subcritical are
 
unsuccessful.
 
In the event either of these challenges exists at a time that the reactor has not been brought
 
below the power associated with the safety system design a core melt sequence exists. In this situation, core degradation can occur rapidly. For this reason, the General Emergency
 
declaration is intended to be anticipatory of the fission product barrier table declaration to permit
 
maximum off-site intervention time.
 
Attachment 4 to 0CAN121102 Proposed EAL Matrix Chart and Review Table (for information)    to 0CAN121102
 
Page 1 of 30
 
GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT ABNORMAL RADIOLOGICAL EFFLUENTS AG1  Offsite dose resulting from an actual
 
or IMMINENT release of gaseous
 
radioactivity > 1000 mR TEDE or
 
5000 mR child thyroid CDE for the
 
actual or projected duration of the
 
release using actual meteorology Emergency Action Level(s):
NOTE: The SM / TSC Director / EOF Director should not wait until the applicable time has elapsed, but should declare the ev ent as soon as it is determined that the condition will likely exceed the applicable time. If dose assessment results are available, the
 
classification should be based on EAL #2
 
instead of EAL #1. Do not delay declaration awaiting dose assessment results.
: 1. VALID reading on Channel 9 on any of the following radiation
 
monitors > the reading shown for 15 minutes: MONITORS - Unit 1 LIMIT RX-9820 Containment Purge 5.90E+2 &#xb5;Ci/cc RX-9825 Radwaste Area 5.36E+2 &#xb5;Ci/cc RX-9830 Fuel Handling Area 4.54E+2 &#xb5;Ci/cc RX-9835 Emerg. Penetration Room 9.56E+3 &#xb5;Ci/cc MONITORS - Unit 2 LIMIT 2RX-9820 Containment Purge 4.46E+2 &#xb5;Ci/cc 2RX-9825 Radwaste Area 3.32E+2 &#xb5;Ci/cc 2RX-9830 Fuel Handling Area 4.46E+2 &#xb5;Ci/cc 2RX-9835 Emerg. Penetration Room 8.84E+3 &#xb5;Ci/cc 2RX-9840 PASS Building 4.42E+3 &#xb5;Ci/cc 2RX-9845 Aux. Building Extension 1.26E+3 &#xb5;Ci/cc 2RX-9850 LLRW Storage Building 1.77E+3 &#xb5;Ci/cc AS1  Offsite dose resulting from an
 
actual or IMMINENT release of
 
gaseous radioactivity > 100 mR
 
TEDE or 500 mR child thyroid CDE
 
for the actual or projected duration
 
of the release Emergency Action Level(s):
NOTE: The SM / TSC Director / EOF 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, the
 
classification should be based on EAL #2
 
instead of EAL #1. Do not delay declaration awaiting dose assessment results.
: 1. VALID reading on Channel 9 on any of the following radiation
 
monitors > the reading shown for 15 minutes: MONITORS - Unit 1 LIMIT RX-9820 Containment Purge 5.90E+1 &#xb5;Ci/cc RX-9825 Radwaste Area 5.36E+1 &#xb5;Ci/cc RX-9830 Fuel Handling Area 4.54E+1 &#xb5;Ci/cc RX-9835 Emerg. Penetration Room9.56E+2 &#xb5;Ci/cc MONITORS - Unit 2 LIMIT 2RX-9820 Containment Purge 4.46E+1 &#xb5;Ci/cc 2RX-9825 Radwaste Area 3.32E+1 &#xb5;Ci/cc 2RX-9830 Fuel Handling Area 4.46E+1 &#xb5;Ci/cc 2RX-9835 Emerg. Penetration Room8.84E+2 &#xb5;Ci/cc 2RX-9840 PASS Building 4.42E+2 &#xb5;Ci/cc 2RX-9845 Aux. Building Extension 1.26E+2 &#xb5;Ci/cc 2RX-9850 LLRW Storage Building 1.77E+2 &#xb5;Ci/cc AA1  Any release of gaseous or liquid
 
radioactivity to the environment
 
> 200 times the ODCM limits for 15 minutes Emergency Action Level(s):
NOTE: The SM / TSC Director / EOF 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 Channel 7 on any of the following radiation
 
monitors > the reading shown for 15 minutes: MONITORS - Unit 1 LIMIT RX-9820 Containment Purge 5.90E0 &#xb5;Ci/cc RX-9825 Radwaste Area 5.36E0 &#xb5;Ci/cc RX-9830 Fuel Handling Area 4.54E0 &#xb5;Ci/cc RX-9835 Emerg. Penetration Room9.56E+1 &#xb5;Ci/cc MONITORS - Unit 2 LIMIT 2RX-9820 Containment Purge 4.46E0 &#xb5;Ci/cc 2RX-9825 Radwaste Area 3.32E0 &#xb5;Ci/cc 2RX-9830 Fuel Handling Area 4.46E0 &#xb5;Ci/cc 2RX-9835 Emerg. Penetration Room8.84E+1 &#xb5;Ci/cc 2RX-9840 PASS Building 4.42E+1 &#xb5;Ci/cc 2RX-9845 Aux. Building Extension 1.26E+1 &#xb5;Ci/cc 2RX-9850 LLRW Storage Building 1.77E+1 &#xb5;Ci/cc OR AU1  Any release of gaseous or liquid
 
radioactivity to the environment
> 2 times the ODCM limits for 60 minutes Emergency Action Level(s):
NOTE: The SM 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 Channel 7 on any of the following radiation
 
monitors > the reading shown for 60 minutes: MONITORS - Unit 1 LIMIT RX-9820 Containment Purge 5.90E-2 &#xb5;Ci/cc RX-9825 Radwaste Area 5.36E-2 &#xb5;Ci/cc RX-9830 Fuel Handling Area 4.54E-2 &#xb5;Ci/cc RX-9835 Emerg. Penetration Room9.56E-1 &#xb5;Ci/cc MONITORS - Unit 2 LIMIT 2RX-9820 Containment Purge 4.46E-2 &#xb5;Ci/cc 2RX-9825 Radwaste Area 3.32E-2 &#xb5;Ci/cc 2RX-9830 Fuel Handling Area 4.46E-2 &#xb5;Ci/cc 2RX-9835 Emerg. Penetration Room8.84E-1 &#xb5;Ci/cc 2RX-9840 PASS Building 4.42E-1 &#xb5;Ci/cc 2RX-9845 Aux. Building Extension 1.26E-1 &#xb5;Ci/cc 2RX-9850 LLRW Storage Building 1.77E-1 &#xb5;Ci/cc OR 1  2  3  4  5  6  D 1  2  3  4  5  6  D 1  2  3  4  5  6  D 1  2  3  4  5  6  D    to 0CAN121102
 
Page 2 of 30
 
GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT ABNORMAL RADIOLOGICAL EFFLUENTS AG1 (continued)
OR 2. Dose assessment using actual meteorology indicates doses
 
> 1000 mR TEDE or 5000 mR
 
child thyroid CDE at or beyond
 
the site boundary.
OR 3. Field survey results indicate closed window dose rates
 
> 1000 mR/hr expected to
 
continue for  60 minutes; or analyses of field survey samples
 
indicate child thyroid CDE
 
> 5000 mR for one hour of
 
inhalation, at or beyond the site
 
boundary.
AS1 (continued)
OR 2. Dose assessment using actual meteorology indicates doses
 
> 100 mR TEDE or 500 mR child
 
thyroid CDE at or beyond the
 
site boundary.
OR 3. Field survey results indicate closed window dose rates
 
> 100 mR/hr expected to
 
continue for  60 minutes; or analyses of field survey samples
 
indicate child thyroid CDE
 
> 500 mR for one hour of
 
inhalation, at or beyond the site
 
boundary. AA1 (continued)
: 2. EITHER VALID reading on any of the following radiation
 
monitors > 200 times the alarm
 
setpoint established by a current
 
release permit for  15 minutes OR VALID reading greater than the value listed for 15 minutes: MONITORS - Unit 1 LIMIT RX-9820 Cont. Purge (Ch. 7 or 9) N/A RX-4830 Waste Gas Monitor 9.5E7 cpm RX-4642 Liquid Radwaste Monitor 9.5E7 cpm RX-9835 Emerg. Penetration RoomN/A MONITORS - Unit 2 LIMIT 2RX-9820 Cont. Purge (Ch. 7 or 9) N/A 2RX-2429 Waste Gas Monitor 9.5E5 cpm 2RX-2330 BMS Discharge Monitor 9.5E5 cpm 2RX-4423 LRW Discharge Monitor 9.5E5 cpm 2RX-4425 SG BD to Flume Monitor 9.5E5 cpm OR 3. Confirmed grab sample analyses for gaseous or liquid
 
releases indicates
 
concentrations or release rates >
 
200 times the applicable values
 
of the ODCM for  15 minutes.
AU1 (continued)
: 2. VALID reading on any of the following radiation monitors
 
> 2 times the alarm setpoint
 
established by a current release
 
permit for  60 minutes:
MONITORS - Unit 1 RX-9820 Cont. Purge (Ch. 7 or 9) RX-4830 Waste Gas Monitor RX-4642 Liquid Radwaste Monitor RX-9835 Emerg. Penetration Room MONITORS - Unit 2 2RX-9820 Cont. Purge (Ch. 7 or 9) 2RX-2429 Waste Gas Monitor 2RX-2330 BMS Discharge Monitor 2RX-4423 LRW Discharge Monitor 2RX-4425 SG BD to Flume Monitor OR 3. Confirmed grab sample analyses for gaseous or liquid
 
releases indicates
 
concentrations or release rates >
 
2 times the applicable values of
 
the ODCM for  60 minutes.
to 0CAN121102
 
Page 3 of 30
 
GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT ABNORMAL RADIATION LEVELS 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 Emergency Action Level(s):
: 1. A water level drop in the refueling canal or spent fuel pool
 
that will result in irradiated fuel
 
becoming uncovered.
OR 2. VALID alarm on any of the following radiation monitors due
 
to damage to irradiated fuel or
 
loss of water level: MONITORS - Unit 1 RX-9820 Containment Purge (Channel 7 or 9) RX-9825 Radwaste Area (Channel 7 or 9) RX-9830 Fuel Handling Area (Channel 7 or 9) RE-8060 Containment High Range Monitor RE-8061 Containment High Range Monitor RE-8009 Spent Fuel Area RE-8017 Fuel Handling MONITORS - Unit 2 2RX-9820 Containment Purge (Channel 7 or 9) 2RX-9825 Radwaste Area (Channel 7 or 9) 2RX-9830 Fuel Handling Area (Channel 7 or 9) 2RE-8905 Containment Equipment Hatch Area 2RE-8909 Containment Personnel Hatch Area 2RE-8925-1/2 Containment High Range Monitors 2RE-8914/15/16Spent Fuel Area Monitors 2RE-8912 Containment Incore Instruments AU2  Unexpected rise in plant radiation
 
levels Emergency Action Level(s):
: 1. a. UNPLANNED lowering of water level in the refueling
 
canal or spent fuel pool as
 
indicated by:  Personnel observation, refueling crew report, indication on area security
 
camera, borated water
 
source (BWST or RWT)
 
level drop due to makeup
 
demands. AND b. VALID Area Radiation Monitor reading rise on any of
 
the following: MONITORS - Unit 1 RE-8009 Spent Fuel Area RE-8017 Fuel Handling Area MONITORS - Unit 2 2RE-8914 Spent Fuel Area 2RE-8915 Spent Fuel Area 2RE-8916 Spent Fuel Area 2RE-8912 Containment Incore Instrumentation OR  1  2  3  4  5  6  D 1  2  3  4  5  6  D    to 0CAN121102
 
Page 4 of 30
 
GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT ABNORMAL RADIATION LEVELS AA3  Rise in radiation levels within the facility that impedes operation of
 
systems required to maintain plant safety functions. Emergency Action Level(s):
: 1. Dose rate > 15 mR/hr in any of the following areas requiring
 
continuous occupancy to maintain plant safety functions:  Unit 1 Control Room  Unit 2 Control Room  Central Alarm Station AU2 (continued)
: 2. UNPLANNED VALID Area Radiation Monitor readings or survey results indicate a rise by
 
a factor of 1000 over normal*
 
levels. NOTE: For area radiation monitors with ranges incapable of measuring 1000 times normal*
 
levels, classification shall be based on VALID full scale indication unless surveys confirm that area radiation levels are below 1000 times normal* within 15 minutes of the Area Radiation Monitor indications going to full scale indication.
* Normal can be cons idered as the highest reading in the past twenty-four hours
 
excluding the current peak value.
1  2  3  4  5  6  D    to 0CAN121102
 
Page 5 of 30
 
GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTION - Loss of RCS / Reactor Vessel Inventory CG1  Loss of RCS / reactor vessel
 
inventory affecting fuel clad integrity
 
with containment challenged Emergency Action Level(s):
NOTE: The SM / TSC Director / EOF Director should not wait until the applicable time has elapsed, but should declare the ev ent as soon as it is determined that the condition will likely exceed the applicable time.
: 1. a. Core exit thermocouples indicate superheat for
 
> 30 minutes.
AND b. Any of the following containment challenge
 
indications:  CONTAINMENT CLOSURE not established  Explosive mixture inside containment  UNPLANNED rise in containment pressure OR 2. a. RCS / reactor vessel level cannot be monitored with
 
core uncovery indicated by
 
any of the following for 30 minutes:
CS1  Loss of RCS / reactor vessel
 
inventory affecting core decay heat
 
removal capability Emergency Action Level(s):
NOTE: The SM / TSC Director / EOF 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:
Loss of RCS / reactor vessel level as indicated by:
Unit 1: RVLMS Levels 1 through 9 indicate DRY Unit 2: RVLMS Levels 1 through 6 indicate DRY OR 2. With CONTAINMENT CLOSURE established, core exit
 
thermocouples indicate
 
superheat.
OR 
 
CA1  Loss of RCS / reactor vessel
 
inventory Emergency Action Level(s):
NOTE: The SM / TSC Director / EOF 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 / reactor vessel inventory as indicated by:
Unit 1: RVLMS Levels 1 through 8 indicate DRY Unit 2: RVLMS Levels 1 through 5 indicate DRY OR Unit 1: Reactor vessel level
< 368 ft., 0 in. (bottom
 
of the hot leg)
Unit 2: Reactor vessel level
< 369 ft., 1.5 in. (bottom
 
of the hot leg)
OR 
 
CU1  RCS leakage Emergency Action Level(s):
NOTE: The SM 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
 
level within Pressurizer or RCS
 
level target band for 15 minutes.
 
CU2  UNPLANNED loss of RCS / reactor
 
vessel Inventory Emergency Action Level(s):
NOTE: The SM 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 / reactor vessel level drop as indicated by
 
either of the following:
5  6 5  6 5  6 5 6    to 0CAN121102
 
Page 6 of 30
 
GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTION - Loss of RCS / Reactor Vessel Inventory CG1 (continued)
Containment High Range Radiation Monitor reading
 
>10 R/hr  Erratic source range monitor indication  Unexplained level rise in Reactor Building Sump, Reactor Drain Tank, Quench Tank, Aux. Building
 
Equipment Drain Tank, or
 
Aux. Building Sump AND b. Any of the following containment challenge
 
indications:  CONTAINMENT CLOSURE not established  Explosive mixture inside containment  UNPLANNED rise in containment pressure CS1 (continued)
: 3. RCS / reactor vessel level cannot be monitored for 30 minutes with a loss of RCS
/ reactor vessel inventory as
 
indicated by any of the following:  Containment High Range Radiation Monitor reading
 
> 10 R/hr  Erratic source range monitor indication  Unexplained level rise in Reactor Building Sump, Reactor Drain Tank, Quench
 
Tank, Aux. Building Equipment
 
Drain Tank, or Aux. Building
 
Sump  CA1 (continued)
: 2. RCS / reactor vessel level cannot be monitored for  15 minutes with a loss of RCS / reactor
 
vessel inventory as indicated by
 
an unexplained level rise in the
 
Reactor Building Sump, Reactor
 
Drain Tank, Aux. Building
 
Equipment Drain Tank, Aux.
 
Building Sump, or Quench Tank.
CU2 (continued)
: a. RCS / reactor vessel water level drop below the reactor
 
vessel flange for 15 minutes when the RCS / reactor
 
vessel level band is
 
established above the reactor
 
vessel flange.
OR b. RCS / reactor vessel water level drop below the RCS /
 
reactor vessel level band for 15 minutes when the RCS /
reactor vessel level band is
 
established below the reactor
 
vessel flange. 2. RCS / reactor vessel level cannot be monitored with a loss
 
of RCS / reactor vessel inventory
 
as indicated by an unexplained
 
level rise in (as applicable) the
 
Reactor Building Sump, Reactor
 
Drain Tank, Aux. Building
 
Equipment Drain Tank, Aux.
 
Building Sump, or Quench Tank.
to 0CAN121102
 
Page 7 of 30
 
GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTION - Loss of Decay Heat Removal CA3  Inability to maintain plant in Cold
 
Shutdown Emergency Action Level(s):
: 1. An UNPLANNED event results in RCS temperature > 200 &deg;F >
 
the specified duration in Table C1. Table C1 RCS Reheat Du ration Thresholds RCS Containment Closure Duration Intact (but not RCS reduced inventory) N/A  60 minutes*Established  20 minutes*Not intact or RCS reduced inventory Not Established  0 minutes
* If an RCS heat removal system is in operation within this time frame and RCS temperature is being reduced, the EAL is not applicable.
OR NOTE: EAL #2 does not apply in solid plant conditions. 2. An UNPLANNED event results in RCS pressure rise > 10 psi
 
due to a loss of RCS cooling.
CU3  UNPLANNED loss of decay heat
 
removal capability with irradiated
 
fuel in the reactor vessel Emergency Action Level(s):
NOTE: The SM 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
 
200 &deg;F. OR 2. Loss of all RCS temperature and RCS / reactor vessel level indication for  15 minutes
. 5  6 5  6    to 0CAN121102
 
Page 8 of 30
 
GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTION - Loss of AC Power CA5  Loss of all offsite and all onsite AC
 
power to Vital 4.16 KV busses 15 minutes Emergency Action Level(s):
NOTE: The SM / TSC Director / EOF 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 offsite and all onsite AC power to Vital 4.16KV
 
busses  15 minutes.
CU5  AC power capability to Vital 4.16 KV
 
busses reduced to a single power
 
source  15 minutes such that any additional single failure would result
 
in station blackout Emergency Action Level(s):
NOTE: The SM 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 Vital 4.16 KV busses reduced to a
 
single power source 15 minutes.
AND b. Any additional single power source failure will result in
 
station blackout.
5  6 5  6  D    to 0CAN121102
 
Page 9 of 30
 
GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTION - Loss of DC Power CU6  Loss of required DC power 15 minutes Emergency Action Level(s):
NOTE: The SM 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. < 105 volts on required Vital DC bus  15 minutes.
 
GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTION - Inadvertant Criticality CU7  Inadvertent criticality Emergency Action Level(s):
: 1. UNPLANNED su stained positive startup rate observed on nuclear
 
instrumentation.
5  6 5  6    to 0CAN121102
 
Page 10 of 30
 
GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTION - Loss of Communications CU8  Loss of all onsite or offsite
 
communications capabilities Emergency Action Level(s):
: 1. Loss of all Table C2 onsite communication methods
 
affecting the ability to perform
 
routine operations.
Table C2 Onsite Communications Equipment Station radio system Plant paging system In-plant telephones Gaitronics OR 2. Loss of all Table C3 offsite communication methods
 
affecting the ability to perform
 
offsite notifications.
Table C3 Offsite Communications Equipment All telephone lines (commercial and microwave)
ENS    5  6  D    to 0CAN121102
 
Page 11 of 30
 
GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT ISFSI MALFUNCTION - Cask Damage E-HU1  Damage to a loaded cask
 
CONFINEMENT BOUNDARY Emergency Action Level(s):
: 1. Damage to a loaded cask CONFINEMENT BOUNDARY.
 
GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT FISSION PRODUCT BARRIER MALFUNCTION - Barriers FG1  Loss of ANY two barriers AND loss
 
or potential loss of third barrier FS1  Loss or potential loss of ANY two
 
barriers FA1  ANY loss or ANY potential loss of
 
EITHER fuel clad or RCS FU1  ANY loss or ANY potential loss of
 
containment Note: Determine which combination of the three barriers are lo st or have a potential loss and use the above key to classify the event. Also, multiple events could occur which result in the conclusion that exceeding the loss or potential loss EALs is IMMINENT. In this IMMINENT loss situation use judgment and classify as if the EALs are exc eeded. 1  2  3  4  5  6  D 1  2  3  4 1  2  3  4 1  2  3  4 1  2  3  4    to 0CAN121102
 
Page 12 of 30
 
Fuel Clad Barrier EALs RCS Barrier EALs Containment Barrier EALs LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS
: 1. Primary Coolant Activity Level (FCB1)
: 1. RCS Leak Rate (RCB1)
: 1. Containment Pressure (CNB1)
: 1. Coolant activity
> 300 &#xb5;Ci/gm dose equivalent I-131 activity by Chemistry sample OR 2. Radiation levels
> 1000 MR/hr Unit 1: at SA-229 Unit 2: at 2TCD-19 None RCS leak rate > available makeup capacity as indicated by:
Unit 1: Loss of adequate subcooling margin Unit 2: RCS subcooling (MTS) can NOT be
 
maintained at least
 
30 &deg;F Unit 1: UNISOLABLE RCS leak > 50 gpm with Letdown isolated Unit 2: UNISOLABLE RCS leak > 44 gpm with Letdown isolated 1. Rapid unexplained drop in containment pressure following an initial rise in
 
containment pressure OR 2. Containment pressure or sump level response not consistent with LOCA
 
conditions
: 1. Unit 1: Containment pressure 73.7 PSIA
 
(59 PSIG) and
 
rising Unit 2: Containment pressure 73.7 PSIA and
 
rising OR 2. Explosive mixture exists inside Containment OR 3. a. Containment Pressure >
containment spray
 
actuation setpoint Unit 1: 44.7 PSIA (30 PSIG)
Unit 2: 23.3 PSIA AND b. LESS THAN one full train of spray
 
operating to 0CAN121102
 
Page 13 of 30
 
Fuel Clad Barrier EALs RCS Barrier EALs Containment Barrier EALs LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS
: 2. Core Exit Thermocouple Readings (FCB2)
: 2. SG Tube Rupture (RCB2)
: 2. Core Exit Thermocouple Readings (CNB2)
> 1200 &deg;F CET temperature Unit 1: ICC exists as evidenced by CETs
 
indicating
 
superheated
 
conditions Unit 2: Average CETs indicate superheat for current RCS
 
pressure SGTR that results in an ECCS (SI) actuation None None 1. a. CETs indicate >
1200 &deg;F AND b. Restoration procedures not effective within
 
15 minutes OR 2. a. CETs indicate >
700 &deg;F AND b. RVLMS indicates Unit 1: Levels 1 through 9 DRY Unit 2: Levels 1 through 7 DRY AND c. Restoration procedures not effective within
 
15 minutes 3. Reactor Vessel Water Level (FCB3)
: 3. Containment Radiation Monitoring (RCB3)3. SG Secondary Side Release With Primary-to-Secondary Leakage (CNB3)
None Unit 1: RVLMS Levels 1 through 9 indicate DRY Unit 2: RVLMS Levels 1 through 7 indicate DRY Containment high range radiation monitor reading >
 
100 R/hr None 1. Primary-to-secondary leakrate > 10 gpm AND 2. UNISOLABLE steam release from affected
 
steam generator to the
 
environment None    to 0CAN121102
 
Page 14 of 30
 
Fuel Clad Barrier EALs RCS Barrier EALs Containment Barrier EALs LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS
: 4. Containment Radiat ion Monitoring (FCB4)4. Emergency Director Judgment (RCB4)
: 4. Containment Isolation Failure or Bypass (CNB4) Containment high range radiation monitor reading >
 
1000 R/hr None Any condition in the opinion of the SM / TSC Director / EOF Director that indicates Loss or Potential Loss of the RCS
 
barrier 1. UNISOLABLE breach of containment AND 2. Direct downstream pathway to the
 
environment exists after
 
containment isolation
 
signal None 5. Core Damage Assessment (FCB5)
: 5. Containment Radiation Monitoring (CNB5)
At least 5% fuel clad damage as determined from core
 
damage assessment None  None Containment high range radiation monitor reading >
 
4000 R/hr 6. Emergency Director Judgment (FCB6)
: 6. Other Indications (CNB6)
Any condition in the opinion of the SM/TSC Director/EOF Director that indicates Loss or Potential Loss of the fuel clad barrier  Elevated readings on the followi ng radiation monitors that indicate loss or potential loss of the Containment barrier:
 
MONITORS - Unit 1 RX-9820 Containment Purge RX-9825 Radwaste Area RX-9830 Fuel Handling Area RX-9835 Emergency Penetration Room MONITORS - Unit 2 2RX-9820 Containment Purge 2RX-9825 Radwaste Area 2RX-9830 Fuel Handling Area 2RX-9835 Emergency Penetration Room 2RX-9840 Post Accident Sampling Building 2RX-9845 Auxiliary Building Extension to 0CAN121102
 
Page 15 of 30
 
Fuel Clad Barrier EALs RCS Barrier EALs Containment Barrier EALs LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS
: 7. Emergency Director Judgment (CNB7)
Any condition in the opinion of the SM / TSC Director / EOF Director that indicates Loss or Potential Loss of the
 
containment barrier
 
GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY - Security HG1  HOSTILE ACTION resulting in loss of
 
physical control of the facility Emergency Action Level(s):
: 1. A HOSTILE ACTION has occurred such that plant personnel are
 
unable to operate equipment
 
required to maintain safety
 
functions.
OR 2. A HOSTILE ACTION has caused failure of Spent Fuel Cooling
 
Systems and IMMINENT fuel
 
damage is likely for a freshly off-
 
loaded reactor core in pool.
HS1  HOSTILE ACTION within the
 
PROTECTED AREA Emergency Action Level(s):
: 1. A HOSTILE ACTION is occurring or has occurred within the
 
PROTECTED AREA as reported by
 
ANO Security Shift Supervision.
HA1  HOSTILE ACTION within the OWNER
 
CONTROLLED AREA or airborne attack threat Emergency Action Level(s):
: 1. A HOSTILE ACTION is occurring or has occurred within the OWNER
 
CONTROLLED AREA as reported
 
by ANO Security Shift Supervision.
OR 2. A validated notification from NRC of an airliner attack threat within
 
30 minutes of the site.
HU1  Confirmed SECURITY CONDITION or
 
threat which indicates a potential
 
degradation in the level of safety of the
 
plant Emergency Action Level(s):
: 1. A SECURITY CONDITION that does not involve a HOSTILE ACTION as reported by ANO
 
Security Shift Supervision.
OR 2. A credible site specific security threat notification.
OR 3. A validated notification from NRC providing information of an aircraft
 
threat. 1  2  3  4  5  6  D 1  2  3  4  5  6  D 1  2  3  4  5  6  D 1  2  3  4  5  6  D    to 0CAN121102
 
Page 16 of 30
 
GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY - Discretionary HG2  Other conditions exist which in the judgment of the SM / TSC Director /
EOF Director warrant declaration of
 
General Emergency Emergency Action Level(s):
: 1. Other conditions exist which in the judgment of the SM / TSC
 
Director / EOF Director indicate
 
that events are in progress or
 
have occurred which involve
 
actual or IMMINENT substantial
 
core degradation or melting with
 
potential for loss of containment
 
integrity or HOSTILE ACTION
 
that results in an actual loss of
 
physical control of the facility. 
 
Releases can be reasonably
 
expected to exceed EPA
 
Protective Action Guideline
 
exposure levels offsite for more
 
than the immediate site area.
HS2  Other conditions exist which in the
 
judgment of the SM / TSC Director /
EOF Director warrant declaration of
 
a Site Area Emergency Emergency Action Level(s):
: 1. Other conditions exist which in the judgment of the SM / TSC
 
Director / EOF 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.
HA2  Other conditions exist which in the judgment of the SM / TSC Director /
EOF Director warrant declaration
 
of an Alert Emergency Action Level(s):
: 1. Other conditions exist which in the judgment of the SM / TSC
 
Director / EOF Director indicate
 
that events are in progress or
 
have occurred which involve an
 
actual or potential substantial
 
degradation of the level of safety
 
of the plant or a security event
 
that involves probable life
 
threatening risk to site personnel
 
or damage to site equipment
 
because of HOSTILE ACTION. 
 
Any releases are expected to be
 
limited to small fractions of the EPA Protective Action Guideline
 
exposure levels.
HU2  Other conditions exist which in the judgment of the SM warrant
 
declaration of an NUE Emergency Action Level(s):
: 1. Other conditions exist which in the judgment of the SM indicate
 
that events are in progress or
 
have occurred which indicate a
 
potential degradation of the level
 
of safety of the plant or indicate a
 
security threat to facility
 
protection has been initiated. No
 
releases of radioactive material
 
requiring offsite response or
 
monitoring are expected unless
 
further degradation of safety systems occurs.
1  2  3  4  5  6  D 1  2  3  4  5  6  D 1  2  3  4  5  6  D 1  2  3  4  5  6  D    to 0CAN121102
 
Page 17 of 30
 
GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY - Control Room Evacuation HS3  Control Room evacuation has been
 
initiated and plant control cannot be
 
established Emergency Action Level(s):
: 1. a. Control Room evacuation has been initiated.
AND b. Control of the plant cannot be established in  accordance
 
with the following procedures
 
within 15 minutes:
Unit 1: 1203.002, "Alternate Shutdown" Unit 2: 2203.014, "Alternate Shutdown" HA3  Control Room evacuation has been
 
initiated Emergency Action Level(s):
: 1. Alternate Shutdown procedure requires Control Room
 
evacuation:
Unit 1: 1203.002, "Alternate Shutdown" Unit 2: 2203.014, "Alternate Shutdown" 1  2  3  4  5  6  D 1  2  3  4  5  6  D    to 0CAN121102
 
Page 18 of 30
 
GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY Fire  Table H1 Unit 1 Unit 2 CA-1 & HP Office Area Condensate Demineralizer Room
 
Corridor 98
 
Fire Area C Lower North Electrical Penetration Room (LNEPR) Lower South Electrical Equipment Room (LSEER) / Air Compressor Room Lower South Electrical Penetration Room (LSEPR) Lower South Piping Penetration Room (LSPPR) Main Steam Isolation Violation (MSIV) Room
 
North Engineered Safeguards (ES) SWGR Room (A4)
South ES SWGR Room
 
Turbine Building  A1, A2, H1, H2 SWGR area  354' Bowling Alley north end west of Breathing Air compressor room  368' West Heater Deck from LSEER (orange door) along east wall of ES
 
SWGR Rooms to Corridor 98 door Upper North Electrical Penetration Room (UNEPR) / Hot Tool Room / Decon Room Upper South Electrical Penetration Room (USEPR) Upper South Piping Penetration Room (USPPR) 2A3 Room 2A4, 2D02, & East Battery Room
 
2B53 Room
 
2B63 Room
 
2B9/2B10 Room 2Y11/13 Equipment Room Auxiliary Building 317' General Access Auxiliary Building 335' Auxiliary Building 354'
 
'B' Engineered Safeguards Features (ESF) Room Corridor Behind Door 340
 
Turbine Building  2A1, 2A2, 2H1, 2H2 Area  354' West wall of Demineralizer area  368' West Heater Deck north of north Switchgear (SWGR)
 
Room (2A3) and East of
 
LNEPR Intake Structure  354' or 366'
 
LNEPR LSEPR Motor-Generator (MG) Set Room
 
Steam Pipe Area
 
Hot Machine Shop
 
UNEPR, UNPPR, LNPPR, USPPR HA4  FIRE or EXPLOSION affecting the
 
operability of plant safety systems required to establish or maintain
 
safe shutdown Emergency Action Level(s):
: 1. FIRE or EXPLOSION resulting in VISIBLE DAMAGE to any Table H1 structure or area containing safety systems or
 
components or Control Room indication of degraded
 
performance of those safety systems:  HU4  FIRE within the PROTECTED AREA
 
not extinguished within 15 minutes of detection or EXPLOSION within
 
the PROTECTED AREA Emergency Action Level(s):
NOTE: The SM 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. FIRE in any Table H1 structure or area not extinguished
: 1) within 15 minutes of Control
 
Room notification or 2) within
 
15 minutes of verification of a
 
Control Room FIRE alarm.
OR 2. EXPLOSION within the PROTECTED AREA.
1  2  3  4  5  6  D 1  2  3  4  5  6  D    to 0CAN121102
 
Page 19 of 30
 
GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY Toxic Gas  Table H1 Unit 1 Unit 2 CA-1 & HP Office Area Condensate Demineralizer Room
 
Corridor 98
 
Fire Area C Lower North Electrical Penetration Room (LNEPR) Lower South Electrical Equipment Room (LSEER) / Air Compressor Room Lower South Electrical Penetration Room (LSEPR) Lower South Piping Penetration Room (LSPPR) Main Steam Isolation Violation (MSIV) Room
 
North Engineered Safeguards (ES) SWGR Room (A4)
South ES SWGR Room
 
Turbine Building  A1, A2, H1, H2 SWGR area  354' Bowling Alley north end west of Breathing Air compressor room  368' West Heater Deck from LSEER (orange door) along east wall of ES
 
SWGR Rooms to Corridor 98 door Upper North Electrical Penetration Room (UNEPR) / Hot Tool Room / Decon Room Upper South Electrical Penetration Room (USEPR) Upper South Piping Penetration Room (USPPR) 2A3 Room 2A4, 2D02, & East Battery Room
 
2B53 Room
 
2B63 Room
 
2B9/2B10 Room 2Y11/13 Equipment Room Auxiliary Building 317' General Access Auxiliary Building 335' Auxiliary Building 354'
 
'B' Engineered Safeguards Features (ESF) Room Corridor Behind Door 340
 
Turbine Building  2A1, 2A2, 2H1, 2H2 Area  354' West wall of Demineralizer area  368' West Heater Deck north of north Switchgear (SWGR)
 
Room (2A3) and East of
 
LNEPR Intake Structure  354' or 366'
 
LNEPR LSEPR Motor-Generator (MG) Set Room
 
Steam Pipe Area
 
Hot Machine Shop
 
UNEPR, UNPPR, LNPPR, USPPR HA5  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 Emergency Action Level(s):
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.
HU5  Release of toxic, corrosive, asphyxiant, or flammable gases
 
deemed detrimental to NORMAL
 
PLANT OPERATIONS Emergency Action Level(s):
: 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
 
offsite event.
1  2  3  4  5  6  D 1  2  3  4  5  6  D    to 0CAN121102
 
Page 20 of 30
 
GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY Natural or Destructive Phenomena Table H1 Unit 1 Unit 2 CA-1 & HP Office Area Condensate Demineralizer Room
 
Corridor 98
 
Fire Area C Lower North Electrical Penetration Room (LNEPR) Lower South Electrical Equipment Room (LSEER) / Air Compressor Room Lower South Electrical Penetration Room (LSEPR) Lower South Piping Penetration Room (LSPPR) Main Steam Isolation Violation (MSIV) Room
 
North Engineered Safeguards (ES) SWGR Room (A4)
South ES SWGR Room
 
Turbine Building  A1, A2, H1, H2 SWGR area  354' Bowling Alley north end west of Breathing Air compressor room  368' West Heater Deck from LSEER (orange door) along east wall of ES
 
SWGR Rooms to Corridor 98 door Upper North Electrical Penetration Room (UNEPR) / Hot Tool Room / Decon Room Upper South Electrical Penetration Room (USEPR) Upper South Piping Penetration Room (USPPR) 2A3 Room 2A4, 2D02, & East Battery Room
 
2B53 Room
 
2B63 Room
 
2B9/2B10 Room 2Y11/13 Equipment Room Auxiliary Building 317' General Access Auxiliary Building 335' Auxiliary Building 354'
 
'B' Engineered Safeguards Features (ESF) Room Corridor Behind Door 340
 
Turbine Building  2A1, 2A2, 2H1, 2H2 Area  354' West wall of Demineralizer area  368' West Heater Deck north of north Switchgear (SWGR)
 
Room (2A3) and East of
 
LNEPR Intake Structure  354' or 366'
 
LNEPR LSEPR Motor-Generator (MG) Set Room
 
Steam Pipe Area
 
Hot Machine Shop
 
UNEPR, UNPPR, LNPPR, USPPR HA6  Natural or destructive phenomena affecting VITAL AREAS Emergency Action Level(s):
: 1. a. Seismic event > Operating Basis Earthquake (OBE) as
 
indicated by annunciation of the 0.1g acceleration alarm.
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 > 67 mph resulting in VISIBLE
 
DAMAGE to any of the following
 
structures/equipment containing safety systems or components or Control Room indication of
 
degraded performance of those safety systems:
HU6  Natural or destructive phenomena affecting the PROTECTED AREA Emergency Action Level(s):
: 1. Seismic event identified by any 2 of the following:  Seismic event confirmed by annunciation of the 0.01g
 
acceleration alarm  Earthquake felt in plant  National Earthquake Center OR 2. Tornado striking within PROTECTED AREA boundary or high winds > 67 mph.
OR 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 structures or areas in Table H1. OR 4. Turbine failure resulting in casing penetration or damage to turbine or generator seals.
OR 1  2  3  4  5  6  D 1  2  3  4  5  6  D    to 0CAN121102
 
Page 21 of 30
 
GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY Natural or Destructive Phenomena
 
Table H2 Reactor Building Turbine Building Intake Structure QCST Ultimate Heat Sink Control Room BWST/RWT Startup Transformers Auxiliary Building Diesel Fuel Vault
 
HA6 (continued)
Reactor Building  Intake Structure  Ultimate Heat Sink  BWST/RWT  Auxiliary Building  Turbine Building  QCST  Control Room  Startup Transformers  Diesel Fuel Vault 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:  Intake Structure  Ultimate Heat Sink  BWST/RWT  Auxiliary Building  Turbine Building  QCST  Control Room  Startup Transformers  Diesel Fuel Vault HU6  (continued)
: 5. Lake Dardanelle level < 335 feet.
OR 6. Lake Dardanelle level > 345 feet.
to 0CAN121102
 
Page 22 of 30
 
GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY Natural or Destructive Phenomena
 
Table H2 Reactor Building Turbine Building Intake Structure QCST Ultimate Heat Sink Control Room BWST/RWT Startup Transformers Auxiliary Building Diesel Fuel Vault
 
HA6 (continued)
OR 4. Turbine failure-generated PROJECTILES resulting in
 
VISIBLE DAMAGE to or penetration of any of the structures/equipment in Table H2 containing safety systems or components or Control Room indication of
 
degraded performance of those safety systems.
OR 5. Lake Dardanelle level < 335 feet and Emergency Cooling Pond
 
inoperable.
OR 6. Vehicle crash resulting in VISIBLE DAMAGE to any of the structures/equipment in Table H2 containing safety systems or
 
components or Control Room indication of degraded
 
performance of those safety systems.      to 0CAN121102
 
Page 23 of 30
 
GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT SYSTEM MALFUNCTION - Loss of AC Power SG1  Prolonged loss of all offsite and all
 
onsite AC power to Vital 4.16 KV
 
busses Emergency Action Level(s):
: 1. a. Loss of all offsite and all onsite AC power to Vital
 
4.16 KV busses.
AND b. Either of the following:  Restoration of at least one Vital 4.16 KV bus in
 
< 4 hours is not likely.
OR  Continuing degradation of core cooling based on
 
Fission Product Barrier
 
monitoring as indicated by CETs  700 &deg;F. SS1  Loss of all offsite and all onsite AC
 
power to Vital 4.16 KV busses 15 minutes Emergency Action Level(s):
NOTE: The SM / TSC Director / EOF 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 offsite and all onsite AC power to Vital 4.16 KV
 
busses  15 minutes.
SA1  AC power capability to Vital 4.16 KV
 
busses reduced to a single power
 
source  15 minutes such that any additional single failure would result
 
in station blackout Emergency Action Level(s):
NOTE: The SM / TSC Director / EOF 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 Vital 4.16 KV busses reduced to a
 
single power source 15 minutes.
AND b. Any additional single power source failure will result in
 
station blackout.
SU1  Loss of all offsite AC power to Vital
 
4.16 KV busses  15 minutes Emergency Action Level(s):
NOTE: The SM / TSC Director / EOF 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 offsite AC power to Vital 4.16 KV busses 15 minutes.
1  2  3  4 1  2  3  4 1  2  3  4 1  2  3  4    to 0CAN121102
 
Page 24 of 30
 
GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT SYSTEM MALFUNCTION - Failure of Reactor Protection System SG3  Automatic trip and all manual
 
actions fail to shutdown the reactor
 
and indication of an extreme
 
challenge to the ability to cool the core exists Emergency Action Level(s):
: 1. a. An automatic trip failed to shutdown the reactor.
AND b. All manual actions do not shutdown the reactor as
 
indicated by reactor power 5%. AND c. Either of the following exist or have occurred due to
 
continued power generation:  CET temperatures at or approaching 1200 &deg;F.
OR  Feedwater flow rate less than: Unit 1: 430 gpm Unit 2: 485 gpm SS3  Automatic trip fails to shutdown the
 
reactor and manual actions taken
 
from the reactor control console are
 
not successful in shutting down the
 
reactor Emergency Action Level(s):
: 1. a. An automatic trip failed to shutdown the reactor.
AND b. Manual actions taken at panel C03 (Unit 1) or panels
 
2C03/2C14 (Unit 2) do not
 
shutdown the reactor as
 
indicated by reactor power 5%. SA3  Automatic trip fails to shutdown the
 
reactor and the manual actions
 
taken from the reactor control
 
console are successful in shutting
 
down the reactor Emergency Action Level(s):
: 1. a. An automatic trip failed to shutdown the reactor as
 
indicated by reactor power 5%. AND b. Manual actions taken at panel C03 (Unit 1) or panels
 
2C03/2C14 (Unit 2)
 
successfully shutdown the
 
reactor as indicated by
 
reactor power < 5%.
1  2 1  2 1  2    to 0CAN121102
 
Page 25 of 30
 
GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT SYSTEM MALFUNCTION - Loss of DC Power SS4  Loss of all Vital DC power 15 minutes Emergency Action Level(s):
NOTE: The SM / TSC Director / EOF 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. < 105 volts on all Vital DC busses  15 minutes.
1  2  3  4    to 0CAN121102
 
Page 26 of 30
 
GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT SYSTEM MALFUNCTION - Loss of Annunciators SS6  Inability to monitor a SIGNIFICANT TRANSIENT in progress Emergency Action Level(s):
NOTE: The SM / TSC Director / EOF 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. UNPLANNED loss of >
approximately 75% of the 
 
following  15 minutes:  Control Room annunciators associated with safety systems. OR  Control Room safety system indication.
AND b. A SIGNIFICANT TRANSIENT in progress.
AND c. Compensatory indications are unavailable.
SA6  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 Emergency Action Level(s):
NOTE: The SM / TSC Director / EOF Director should not wait until the applicable time has elapsed, but should declare the ev ent as soon as it is determined that the condition has exceeded, or will likely exceed, the applicable time.
: 1. a. UNPLANNED loss of >
approximately 75% of the following  15 minutes:  Control Room annunciators associated with safety systems. OR  Control Room safety system indication.
AND b. Either of the following:  A SIGNIFICANT TRANSIENT is in progress OR  Compensatory indications are unavailable SU6  UNPLANNED loss of safety system
 
annunciation or indication in the
 
Control Room for  15 minutes Emergency Action Level(s):
NOTE: The SM / TSC Director / EOF 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 >
approximately 75% of the following  15 minutes: a. Control Room annunciators associated with safety systems. OR b. Control Room safety system indication.
1  2  3  4 1  2  3  4 1  2  3  4    to 0CAN121102
 
Page 27 of 30
 
GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT SYSTEM MALFUNCTION - RCS Leakage SU7  RCS leakage Emergency Action Level(s):
: 1. Unidentified or pressure boundary leakage > 10 gpm.
OR 2. Identified leakage > 25 gpm.
1  2  3  4    to 0CAN121102
 
Page 28 of 30
 
GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT SYSTEM MALFUNCTION - Loss of Communications SU8  Loss of all onsite or offsite
 
communications capabilities Emergency Action Level(s):
: 1. Loss of all Table M1 onsite communications methods
 
affecting the ability to perform
 
routine operations.
Table M1 Onsite Communications Methods Station radio system Plant paging system In-plant telephones Gaitronics OR 2. Loss of all Table M2 offsite communications methods
 
affecting the ability to perform
 
offsite notifications.
Table M2 Offsite Communications Methods All telephone lines (commercial and microwave)
ENS    1  2  3  4    to 0CAN121102
 
Page 29 of 30
 
GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT SYSTEM MALFUNCTION - Fuel Clad Degradation SU9  Fuel clad degradation Emergency Action Level(s):
: 1. Failed Fuel Iodine radiation monitor reading indicates fuel
 
clad degradation > Technical
 
Specification allowable limits:
Unit 1: RI-1237S reads
> 1.3 x 10 5 cpm Unit 2: 2RITS-4806B reads
> .65 x 10 5 cpm OR 2. RCS sample activity value indicating fuel clad degradation
 
> Technical Specification
 
allowable limits:  > 1.0 uCi/gm Dose Equivalent I-131 for more than 48 hours OR  Unit 1:  60 uCi/gm Dose Equivalent I-131 Unit 2: > 60 uCi/gm Dose Equivalent I-131 OR  1  2  3  4    to 0CAN121102
 
Page 30 of 30
 
GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT SYSTEM MALFUNCTION - Fuel Clad Degradation SU9 (continued)
Unit 1: > 2200 &#xb5;Ci/gm Dose Equivalent Xe-133
 
for more than
 
48 hours Unit 2: > 3100 &#xb5;Ci/gm Dose Equivalent Xe-133
 
for more than
 
48 hours SYSTEM MALFUNCTION - Inadvertant Criticality SU10  Inadvertent criticality Emergency Action Level(s):
: 1. An UNPLANNED sustained positive startup rate observed on
 
nuclear instrumentation.
SYSTEM MALFUNCTION - Failure to Shutdown SU11  Inability to reach required operating
 
mode within Technical Specification
 
limits Emergency Action Level(s):
: 1. A Plant is not brought to required operating mode within Technical
 
Specifications LCO action
 
statement time.
3  4 1  2  3  4 
 
Attachment 5 to 0CAN121102 Supporting Referenced Document Pages    to 0CAN121102
 
Page 1 of 21
 
Arkansas Nuclear One (ANO) Units 1 and 2 NEI 99-01 Revision 5 EAL Supporting Document Table ANO lC/EAL Subject Supporting Document Page Number(s)
FCBI, FCB5, RCB4, CNB6 Fuel cladding radiation monitoring system/survey
 
readings Calculation 03-E-0002-01 22, 27, 38, 42, 54, 46, 49, 60 FCB4 Reactor vessel levels Calculation 90-E-01 16-01 Calculation 84-EQ-0080-02
 
ULD-1-SYS-24
 
ULD-2-SYS-24 238 16 11 9 RCB2 Subcooling margin Calculation 90-E-0116-07 Calculation 90-E-0116-01 74, 75 196, 197 SGI Loss of offsite power Calculation 85-E-0072-02 Calculation 85-E-0072-01 1 1      to 0CAN121102
 
Page 2 of 21
 
Attachment 1, Calculation No. 90-E-0116-01, Page 238 OP Setpoint No.:
R.3 Revision:  10  Parameter:
RVLMS  Setpoint Value:
RVLMS LVL 06 or higher elevation indicates WET Applicability:
Associated System/Component:
RVLMS
 
== Description:==
 
RVLMS level which indicates the core is covered with coolant. This value is used to verify RCS
 
inventory control.
 
Key Assumptions:
Basis:  The lowest reactor vessel level sensor is #7, and is 47 inches above the top of the core (ref. 1). 
 
Sensor #7 corresponds to RVLMS LVL 06 on the ICC monitoring panel 2C388 (ref. 2).
 
The only instrument uncertainty associated with the RVLMS is the response time of the level
 
probes and the discrete layout of the sensors (ref. 1). Since this value is used during relatively
 
steady state conditions no uncertainty was incorporated
 
- Note, this is not an instrument alarm or actuation setpoint but a value to be used by the operator in the control room to determine required actions within the EOPs or adequacy of
 
equipment operation.
 
==References:==
: 1) Calculation No. 84-EQ-0080-01, Rev. 2, Determination of Reactor Vessel Level Measurement Uncertainties.
: 2) TM T068.0040, Technical Manual For Reactor Ve ssel Monitoring System ANO Unit 2, Rev.
2, 1/2/91.
 
to 0CAN121102
 
Page 3 of 21
 
84-EQ-0080-02 PAGE 16 of 93 REV 2  DRAINDOWN 1 2 3 4 5 6 7 8 9 SENSOR # AND TYPE SENSOR LOC REF. RVL (FEET)
REGION DRN RATE @ SNS FT/S COLL LVL @
SNS SEC. MAN. ERR. FT. MAN. DEL.
SEC. TC ERR DEGF TC OUT DELAY SEC. 2 SLOW 36.50 DOME 0.105 342.86 0.33 3.14 10 101.79 3 SLOW 34.88 DOME 0.105 358.28 0.33 3.14 10 101.79 4 SLOW 33.13 DOME 0.105 374.89 0.33 3.14 10 101.79 5 SLOW 31.38 DOME 0.105 391.50 0.33 3.14 10 101.79 6 SLOW 29.63 PLENUM 0.105 408.11 0.42 4.01 10 101.79 7 SLOW 27.88 PLENUM 0.025 428.52 0.42 16.84 10 101.79 8 SLOW 26.13 PLENUM 0.025 498.36 0.42 16.84 10 101.79 9 SLOW 24.29 PLENUM 0.025 571.51 0.42 16.84 10 101.79 10 SLOW 22.46 PLENUM 0.025 644.71 0.42 16.84 10 101.79 10 11 12 13 14 15 16 17 18 DAS ERR DEGF DAS ERR DEL SEC. DAS UNCOV TIME SEC. LVL AT DAS UNC FT. DAS UNC UNCERT FT.
SPDS DEL.
SEC. SPDS UNCOV TIME SEC.
LVL @ SPDS UNC FT. SPDS UNC. UNCERT FT. 8.96 43.93 491.72 26.30 -10.20 30 521.72 25.55 -10.95 8.96 43.93 507.14 25.91 -8.97 30 537.14 25.16 -9.72 8.96 43.93 523.75 25.49 -7.63 30 553.75 24.74 -8.38 8.96 43.93 540.36 25.08 -6.30 30 570.36 24.33 -7.05 8.96 43.93 557.83 24.64 -4.98 30 587.83 23.89 -5.73 8.96 43.93 591.08 23.81 -4.06 30 621.08 23.06 -4.81 8.96 43.93 660.92 22.07 -4.06 30 690.92 21.32 -4.81 8.96 43.93 734.07 20.24 -4.06 30 764.07 19.49 -4.81 8.96 43.93 807.27 18.41 -4.05 30 837.27 17.66 -4.80 2    to 0CAN121102
 
Page 4 of 21
 
ARKANSAS NUCLEAR ONE UPPER LEVEL DOCUMENT ANO-1 INADEQUATE CORE COOLING SYSTEM NO.: ULD-1-SYS-24 REV. NO.: 4
 
PAGE: 11  The ATC sensor in the RLI just under the r eactor head provides a unique benefit. The sensor monitors temperature at the metal-to-liquid interf ace of the reactor head, which is influenced by the sensible heat stored in the reactor head. Use of this reactor vessel head fluid temperature measurement by operators helps to prevent i nadvertent steam bubble formation during natural circulation cooldown and during RCS depressurization.  (REF. 21, 24, 63)
B. The RGT sensors installed in the reactor ve ssel dome and upper plenum regions will give an early warning of the approach to inadequate core cooling.
The sensors are axially located to provide optimum resolution in the areas of most concern. (REF. 21, 25, 26)
The collapsed liquid level indication provided by t he RGT sensors of the RVLMS portion of ICCMDS furnishes the operators with reactor coolant in ventory trend information during the approach to an ICC event and the recovery from an ICC event.
The level information allows the operators to determine if reactor vessel inventor y is increasing or decreasing. (REF. 21, 27)
Reactor Coolant Pumps (RCP) status is used to va lidate hot leg water level and reactor vessel level, in that with any of the four RCPs running, flow induced turbulence produced by the pumps results in inaccurate sensed level. Therefore, when any of the RCPs are running the hot leg water level displays and the reactor vessel level sensors in t he plenum are invalid. Inventory measurements in the upper head region are not seriously affected by the operation of the RCPs and so will be available to give advanced warning of an approach to ICC. (REF. 19, 27, 28, 29, 73)
This operation is confirmed by surveillance testing utilizing simulated RCP field inputs. (REF. 49)  C. The RLls presently installed in ANO-1 extend from the top of the reactor vessel to the fuel alignment plate. The above-core level indication provided by the RLIs is used to monitor the approach to and recovery from ICC conditions. The CET portion of ICCMDS is utilized to identify core uncovery (fuel rod cladding temperature) when reactor coolant level drops below the top of the core. (REF. 8, 19, 27, 30)  D. The 24 radially distributed CETs (six per each core quadrant) provide indication of the temperature rise across representative regions of the core.
These CETs are part of and are located in the upper portion of the incore neutron detector assembly. (REF. 7, 17, 18) Note, temporary CETs, with associated raceways, are installed in the reacto r building. These temporary CET5 are used to monitor core conditions via ICCMDS during refueling conditions. (REF. 77)  GL 88-17 requirements are met as follows:
A. Two independent indications of average CET temperatur e information is readily available for display at any time via the LMD and is logged every hour during cold shutdown. CET temperature information is also continuously available on the Mimic Display Monitor. Additionally, contingency steps exist for monitoring CET temperatures should LMD indications become unavailable. (REF. 59, 60, 62, 63) to 0CAN121102
 
Page 5 of 21
 
ARKANSAS NUCLEAR ONE UPPER LEVEL DOCUMENT ANO-2 INADEQUATE CORE COOLING SYSTEM NO.: ULD-2-SYS-24 REV. NO.: 2
 
PAGE: 9  The ATC sensor in the RLI just under the r eactor head provides a unique benefit. The sensor monitors temperature at the metal-to-liquid interf ace of the reactor head, which is influenced by the sensible heat stored in the reactor head. This temperature measurement provides valuable information concerning the influence of this sensible heat on coolant conditions at the metal-to-liquid interface. Use of these temperature measurement s by operators helps to pr event inadvertent steam bubble formation during natural circulation cooldow n and during repressurization after a small break LOCA. (REF. 1, 25)
B. The RGT sensors installed in the reactor vesse l head will give an early warning of the approach to inadequate core cooling. The sensors are axially lo cated to provide optimum resolution in the areas of most concern. (REF. 1, 6, 24)
The collapsed liquid level indication provided by the RGT sensors of the RVLMS furnishes the operators with reactor coolant inventory trend in formation during the approach to an ICC event and the recovery from an ICC event. The level inform ation allows the operators to determine if reactor vessel inventory is increasing or decreasing. (REF. 1, 6, 20)
Process inputs to DAS consists of "slow" and "fast" RGT sensors.
These inputs are differential temperature signals which correspond to the tem perature difference across the argon gas annulus in a "slow" sensor or the differ ence in surface and interior temperat ures in the "fast" sensor. The slow and fast sensors are used to determine "WET", "DRY" and "QUENCH" states for output to the level display. These signals also generate a displa y alarm on a "DRY" state for either sensor type. (REF. 1, 6, 20)
When the Reactor Coolant Pumps (RCPs) are r unning, all except the upper head region (dome) sensors are interlocked (REF. 16, 26) to read "INVALID" due to flow induced turbulence that may offset the sensor outputs (REF. 1, 20, 24). Inventory measurements in the upper head region are not seriously affected by the operation of t he RCPs and so will be available to give advanced warning of an approach to ICC.
C. The RLls presently installed in ANO-2 extend from the top of the reactor vessel to the bottom of the core. However, for ICC monitoring, only the portions of the RLls above the co re' (from the top of the reactor vessel to the absolute thermocouples located at the core exit) are c onsidered part of the ICC Monitoring System. The above-core portions of the RLls are used to monitor the approach to and recovery from ICC conditions. The CETs are ut ilized to estimate core uncovery (fuel rod cladding temperature) when reactor coolant le vel drops below the top of the core. The in-core portions of the RLls are used for core heat-transfer trending, DT measurement across the core, and local power monitoring. (REF. 1, 6, 20, 21)
D. The 42 radially distributed CETs provide indication of the temper ature rise across representative regions of the core. These CETs are part of and are located in the upper portion of the in-core neutron detectors. (REF. 1, 17, 27)
The ATC sensors of the RVLMS can satisfy one of the requirements for continuous temperature indication of core exit conditions whenever the RCS is in a reduced inventory condition and the head is on (REF. 48, 49). to 0CAN121102
 
Page 6 of 21
 
CALCULATION 03-E-0002-01 Rev. 0 Page 22 of 63
 
A plot of the response is shown below.
 
Figure 1:  Unit 1 RB @ 300 &#xb5;Ci/gm DEQ I-131 & 5% Clad Failure
 
For simplicity of presentation to the Shift Manager acting as the Site Emergency Coordinator, a
 
high range radiation monitor reading of at least 1000 rem/hr between 15 minutes and 2 hours
 
following reactor shutdown would adequately represent 2-5% cladding failure. The two hour
 
point is chosen because it allows ample time for the transfer of Site Emergency Coordinator
 
duties to outside the control room. According to reference 23, during the initial fifteen minutes
 
after a thermal event inside containment, the high range radiation monitor readings are
 
considered invalid due to possibility of a transient thermally induced current.
 
This value is therefore recommended as the EAL setpoint for containment high range radiation
 
monitors indicating a loss of the fuel cladding barrier.
 
1.0E+04 1.0E+03 1.0E+02 R/hr hrs 0 4 8 12 5% CF 300 &#xb5;Ci/gm    to 0CAN121102
 
Page 7 of 21
 
CALCULATION 03-E-0002-01 Rev. 0 Page 27 of 63
 
A plot of the response is shown below.
 
Figure 2:  Unit 2 RB @ 300 &#xb5;Ci/gm DEQ I-131 & 5% Clad Failure
 
For simplicity of presentation to the Shift Manager acting as the Site Emergency Coordinator, a
 
high range radiation monitor reading of at least 1000 rem/hr between 15 minutes and 2 hours
 
after reactor shutdown would adequately represent 2-5% cladding failure. The two hour point is
 
chosen because it allows ample time for the transfer of Site Emergency Coordinator duties to
 
outside the control room. According to reference 23, during the initial fifteen minutes after a
 
thermal event inside containment, the high range radiation monitor readings are considered
 
invalid due to possibility of a transient thermally induced current.
 
This value is therefore recommended as the EAL setpoint for containment high range radiation
 
monitors indicating a loss of the fuel cladding barrier.
 
1.0E+04 1.0E+03 1.0E+02 R/hr hrs 0 4 8 12 5% CF 300 &#xb5;Ci/gm    to 0CAN121102
 
Page 8 of 21
 
CALCULATION 03-E-0002-01 Rev. 0 Page 38 of 63
 
A plot of the response is shown below.
 
Figure 5:  Unit 1 RB After RCS Blowdown at 60 &#xb5;Ci/gm
 
A high range radiation monitor reading of  100 R/hr between 15 minutes and 2 hours following reactor shutdown would adequately represent the release of reactor coolant of 60 &#xb5;Ci/gm DEQ
 
I-131 due to iodine spiking into the containment. This reading is an order of magnitude lower
 
than that specified for the Fuel Clad Barrier EAL. This corresponds to the intent of NEI 99-01
 
PWR RCS Barrier EAL #4.
 
1.0E+03 1.0E+02 1.0E+01 R/hr hrs 0 4 8 12 Average R/hr    to 0CAN121102
 
Page 9 of 21
 
CALCULATION 03-E-0002-01 Rev. 0 Page 42 of 63 Figure 6:  Unit 2 RB After RCS Blowdown at 60 &#xb5;Ci/gm
 
A high range radiation monitor reading of 100 R/hr between 15 minutes and 2 hours following
 
reactor shutdown would adequately represent the release of reactor coolant of 60 &#xb5;Ci/gm DEQ
 
I-131 due to iodine spiking into the containment. This reading is an order of magnitude lower
 
than that specified for the Fuel Clad Barrier EAL. This corresponds to the intent of NEI 99-01
 
PWR RCS Barrier EAL #4.
 
The two hour point is chosen because it allows ample time for the transfer of Site Emergency
 
Coordinator duties to outside the control room. According to reference 23, during the initial
 
fifteen minutes after a thermal event inside containment, the high range radiation monitor
 
readings are considered invalid due to possibility of a transient thermally induced current.
 
1.0E+02 1.0E+01 1.0E+00 R/hr hrs 0 4 8 12 60 &#xb5;Ci/gm 1.0E+03    to 0CAN121102
 
Page 10 of 21
 
CALCULATION 03-E-0002-01 Rev. 0 Page 46 of 63 Figure 7:  Unit 1 RB After RCS Blowdown @ 20% Clad Failure
 
For simplicity of presentation to the Shift Manager acting as the Site Emergency Coordinator, a
 
high range radiation monitor reading of at least 4000 rem/hr between 15 minutes and 2 hours
 
after reactor shutdown would adequately represent 20% cladding failure. The two hour point is
 
chosen because it allows ample time for the transfer of Site Emergency Coordinator duties to
 
outside the control room. According to reference 23, during the initial fifteen minutes after a
 
thermal event inside containment, the high range radiation monitor readings are considered
 
invalid due to possibility of a transient thermally induced current. While the above plot shows
 
the radiation levels dropping below 4000 R/hr at about 1.5 hours, this setpoint is judged to be
 
adequate, for the purpose of the EAL, which is to indicate significant fuel damage well in excess
 
of the EAL associated with loss of Fuel Clad, and as a decision point for declaration of a
 
General Emergency. The situation will likely be diagnosed from radiation levels in less than
 
1.5 hours.
 
1.0E+04 1.0E+03 R/hr hrs 0 4 8 12 Average    to 0CAN121102
 
Page 11 of 21
 
CALCULATION 03-E-0002-01 Rev. 0 Page 49 of 63 Figure 8:  Unit 2 RB After RCS Blowdown @ 20% Clad Failure
 
For simplicity of presentation to the Shift Manager acting as the Site Emergency Coordinator, a
 
high range radiation monitor reading of at least 4000 rem/hr between 15 minutes and 2 hours
 
after reactor shutdown would adequately represent 20% cladding failure. This EAL setpoint is
 
used to indicate significant fuel damage well in excess of the EAL associated with loss of Fuel
 
Clad, and as a decision point for declaration of a General Emergency.
 
The two hour point is chosen because it allows ample time for the transfer of Site Emergency
 
Coordinator duties to outside the control room. According to reference 23, during the initial
 
fifteen minutes after a thermal event inside containment, the high range radiation monitor
 
readings are considered invalid due to possibility of a transient thermally induced current.
 
1.0E+04 1.0E+03 R/hr hrs 0 4 8 12 Average    to 0CAN121102
 
Page 12 of 21
 
CALCULATION 03-E-0002-01 Rev. 0 Page 54 of 63 Figure 9:  SA-229 Dose Rate at 300 &#xb5;Ci/gm and 5% Clad Failure in Unit 1 RCS
 
For simplicity of presentation to the Shift Manager acting as the Site Emergency Coordinator, a
 
reading of at least 1000 mRem/hr within 2 hours after reactor shutdown measured at 1 foot from
 
SA-229, six inches from the end of the piping, would adequately represent 2-5% cladding
 
failure. The two hour point is chosen because it allows ample time for the transfer of Site
 
Emergency Coordinator duties to outside the control room. This value is therefore
 
recommended as the EAL setpoint indicating a loss of the fuel cladding barrier.
 
1.0E+04 1.0E+03 1.0E+02 mR/hr hrs 0 4 8 12 SA229 at 5% CF SA229 at 300 &#xb5;Ci/gm    to 0CAN121102
 
Page 13 of 21
 
CALCULATION 03-E-0002-01 Rev. 0 Page 54 of 63 Figure 11: 2TCD-19 Dose Rate at 300 &#xb5;Ci/gm DEQ I-131 and 5% Clad Failure in Unit 2 RCS
 
For simplicity of presentation to the Shift Manager acting as the Site Emergency Coordinator, a
 
reading of > 1000 mR/hr within 2 hours after reactor shutdown measured at 1 foot from the
 
midpoint of 2TCD-19 would adequately represent 2-5% cladding failure. The two hour point is
 
chosen because it allows ample time for the transfer of Site Emergency Coordinator duties to
 
outside the control room. This value is therefore recommended as the EAL setpoint indicating a
 
loss of the fuel cladding barrier.
 
1.0E+04 1.0E+03 1.0E+02 mR/hr hrs 0 4 8 12 2TCD-19 at 5% CF 2TCD-19 at 300 &#xb5;Ci/gm    to 0CAN121102
 
Page 14 of 21
 
CALCULATION 90-E-0116-07 Rev. 4, Attachment 1 Page 74 of 238 OP Setpoint No.:
B.19 Revision:  3  Parameter:
MINIMUM SUBCOOLING MARGIN Setpoint Value:
EOP Figure 1 Applicability:
Associated System/Component:
RCS
 
== Description:==
 
Minimum Subcooling Margin according to RCS pressure.
Key Assumptions:
NONE  Basis:  The basis for this curve (Figure 1, curve 1) in the EOP is the instrument error for accident conditions associated with CET temperature indications. During natural circulation, CET
 
subcooling margin is utilized, while the T-hot indications are utilized when RCP's are running. 
 
CET subcooling margin errors were utilized, becaus e they are slightly more conservative than those incorporating T-hot. These errors were extracted from Reference 1, pages 32-36 for
 
pressures < 500 psig and Reference 3 for pressures > 500 psig. The data points for the curve
 
in Figure 1 are given below by the "Bounding Error" Column:
RCS Pressure (psig) RCS Temperature (&deg;F)
Bounding Error RCS Temperature (&deg;F)
Actual Error 2500 639 653 652  2000 607 620 619  1500 567 578 577  1000 516 (496) 522 500 420 433 432  400 398 408 350 386 (366) 392 300 352 375 200 318 328 150 296* 296 100 268* 251 50 228* 178
* 70 &deg;F subcooling margin is incorporated below 150 psig, since the actual error is too prohibitive. See below for further explanation. to 0CAN121102
 
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At 1000 psig and 350 psig the instrument error applied to the saturation curve changes to a higher value (indicated by number in parentheses). The bounding error to 1000 psig is 30 &deg;F, 50&deg;F to 350 psig, and 70 &deg;F to just below 150 psig. The calculated accident condition error
 
becomes too prohibitive to include below 150 psig, therefore 70 &deg;F will be assumed below this
 
pressure until boiling is no longer a possibility below 200 &deg;F (since the maximum instrument
 
error associated with the CETs is less than 10 &deg;F from Ref 3 and the boiling point at
 
atmospheric pressure is 212 &deg;F). At 150 psig the combined accident error in the positive
 
direction applied to the minimum subcooling margin curve and the accident error in the negative
 
direction applied to the maximum 200 &deg;F subcooling margin PTS curve would allow for little to
 
no maneuverability. Also, below 150 psig, DHR will most likely have been initiated and
 
temperature indications from the DHR loop along with an alarm for DH pump suction
 
temperature greater than 280 &deg;F will be avail able providing corroboration (Reference 2).
 
Note: Since this curve incorporates accident error, it is very conservative for non-accident conditions. However, it will be used for both abnormal and accident conditions in the
 
interest of simplicity.
 
==References:==
: 1) Engineering Report 91-R-1011-01, Rev. 11 12, "Instrumentation Error Evaluation for ANO-1 EOP Setpoint Verification Project per IRF No. 6051, 6078 and 6095."
: 2) Drawing M-418, sheet 2, Rev. 16, "Functional Description And Logic Diagram - Decay Heat Removal System."
: 3) Engineering calculation 88-EQ-0006-01, Rev. 1, including DRN 05
-2991 6-2677, "ICC Tsat Loop Accuracy."
to 0CAN121102
 
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Attachment 1, CALCULATION 90-E-0116-01, Page 196 OP Setpoint No.:
M.1 Revision:  11  Parameter:
RCS SUBCOOLED MARGIN Setpoint Value:
30 &deg;F  Applicability:
Associated System/Component:
RCS/SMM
 
== Description:==
 
Margin to saturation, taken from 2XI-4612-3,-4, 2XR-4612, or SPDS with or without forced
 
coolant flow, with the saturation margin monitor in Tsat mode, which indicates that the RCS is
 
subcooled.
 
Key Assumptions:
Basis:  This value is based on providing assurance of adequate RCS subcooling with consideration
 
given to the instrument inaccuracy of the above instruments. The worst case uncertainties, with a harsh containment environment, of the instruments mentioned above occur on 2XR-4612 and are tabulated below (ref. 1 and 2):
 
Inaccuracy (&deg;F) Pressure (psia) 13.50 2500 17.41 1500 21.74 1000 26.90 700 29.62 600 31.28 550 33.28 500 38.52 400 46.90 300 The historical value used for this setpoint is 30 &deg;F. The instrument inaccuracy exceeds this
 
value at just below 600 psia. Therefore the value of 30 &deg;F bounds the uncertainty over most of
 
the range except below approximately 600 psia. This is acceptable as most operator actions
 
based on this setpoint occur at higher RCS pressures and utilize criteria in addition to margin to
 
saturation. For instance, HPSI termination and RCP restart criteria include, among other things, margin to saturation greater than 30 &deg;F. Below 1000 psia or so the RCP NPSH limits for one
 
RCP or two RCPs in opposite loops in operation are more restrictive anyway. The Safety    to 0CAN121102
 
Page 17 of 21
 
Function Status Check (SFSC) uses this value to ve rify RCS inventory control. It is not the only parameter monitored to check inventory control as pressurizer level and reactor vessel level are
 
also monitored. The reactor vessel level monitoring system should provide sufficient
 
corroborative measures to counteract the higher uncertainties in margin to saturation at lower
 
pressures.
 
In FRP, IC-2 there is a step which directs the operator to reduce RCS pressure to less than
 
600 psia to attempt to discharge the SITs if MTS is less than 30 &deg;F. However, whenever the
 
pressure is below 600 psia the 30 &deg;F may not be bounding depending on the actual accident
 
conditions. Other corroborative measures are present which will support the use of this setpoint
 
value. This step will also restore the MTS. In blackout there is a step which directs the operator
 
to reduce RCS temperature if MTS is less than 30 &deg;F (due to RCS pressure decay). Again, this
 
action is taken at higher RCS pressures, hence the 30 &deg;F setpoint should still be bounding.
 
Note: SPDS instrument uncertainty for subcooled margin based on CETs or Thot is bounded by the uncertainty discussed above (ref. 1).
 
==References:==
: 1) Calculation No. 85-EQ-0004-21, Rev. 7(1), Loop Error Analysis for Subcooling Margin Monitor. 
: 2) ER980574 I204 Rev. 0 "EOP"
 
to 0CAN121102
 
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CALCULATION COVER SHEET FORM 203F2 Rev. 7/15/85 Page 1 of 1
 
Proj. No. - Calc. No.:
85-E-00072-02 Plant/Unit:
ANO-1    Calc. Title:
Time from Loss of All AC Power to Loss of Subcooling    Proj. Title:
Q Non-Q Seis. I Seis. II Non-Seis Fire Other  Calc. Type (Chk. & Provide Information) Calc. Status (Chk. & Provide Info.)
ANO Piping    T  W  S  O  HGR Line Class:
ANO General - Log Sect:
Nuclear Engineering Foss. Piping    T  W  S  O  HGR
____________________________
Foss. General - Log Sect:  New Calc.
Supercedes Calc #  __________
Amends Calc #  __________
Voids Calc #  __________
Verification Method:        Design Review Alternate Calcs.
Qual. Testing Purpose: To determine how long following loss of all AC till charging of the RCS must be accomplished to avoid losing subcooling.
Results: In the above described scenario, subcooling will not be lost if charging of the RCS is accomplished within 41/2 hours.
Revision No. 0 Pgs. Rev'd or added    By/Init./date (signature on file)
Chk/Init./date Rvw/Init./date (signature on file)
Orig. Displn. Apv/Init./date (signature on file)
Pgs. affected By/Init./date Chk/Init./date Rvw/Init./date Supp. Displn. Apv/Init./date to 0CAN121102
 
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CALCULATION COVER SHEET FORM 203F2 Rev. 7/15/85 Page 1 of 1
 
Proj. No. - Calc. No.:
85-E-00072-01 Plant/Unit:
ANO-2    Calc. Title:
Time from Loss of All AC Power to Loss of Subcooling    Proj. Title:
Q Non-Q Seis. I Seis. II Non-Seis Fire Other  Calc. Type (Chk. & Provide Information) Calc. Status (Chk. & Provide Info.)
ANO Piping    T  W  S  O  HGR Line Class:
ANO General - Log Sect:
Nuclear Engineering Foss. Piping    T  W  S  O  HGR
____________________________
Foss. General - Log Sect:  New Calc.
Supercedes Calc #  __________
Amends Calc #  __________
Voids Calc #  __________
Verification Method:        Design Review Alternate Calcs.
Qual. Testing Purpose: To determine how long following loss of all AC till charging of the RCS must be accomplished to avoid losing subcooling.
Results: In the above described scenario, subcooling will not be lost if charging of the RCS is accomplished within 61/2 hours and appropriate operator actions are taken.
Revision No. 0 Pgs. Rev'd or added    By/Init./date (signature on file)
Chk/Init./date Rvw/Init./date (signature on file)
Orig. Displn. Apv/Init./date (signature on file)
Pgs. affected By/Init./date Chk/Init./date Rvw/Init./date Supp. Displn. Apv/Init./date to 0CAN121102
 
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CROSS REFERENCE MATRIX FROM NEI EAL NUMBER TO ENTERGY EAL NUMBER ANO IC NEI 99-01 IC Diff. Doc. Page Clean Basis Page No AU1 AU1 5 10  AU2 AU2 8 13 AA1 AA1 11 15 AA2 AA2 14 18 AA3 AA3 16 20 AS1 AS1 18 21 AG1 AG1 21 23 CU1 CU1 23 26 CU2 CU2 24 27 CU3 CU4 26 29 CU5 CU3 27 30 CU6 CU7 28 31 CU7 CU8 29 32 CU8 CU6 30 33 CA1 CA1 32 34 CA3 CA4 35 36 CA5 CA3 36 38 CS1 CS1 38 39 CG1 CG1 41 41 E-HU1 E-HU1 43 44 FU1 FU1 44 N/A FA1 FA1 45 N/A FS1 FS1 46 N/A FG1 FG1 47 N/A Fuel Clad EALs Fuel Clad EALs 49 47 RCS EALs RCS EALs 53 53 Containment EALs Containment EALs 58 57 HU1 HU4 62 66 HU2 HU5 63 68 HU4 HU4 65 69    to 0CAN121102
 
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CROSS REFERENCE MATRIX FROM NEI EAL NUMBER TO ENTERGY EAL NUMBER ANO IC NEI 99-01 IC Diff. Doc. Page Clean Basis Page No HU5 HU3 67 71  HU6 HU1 68 72 HA1 HA4 71 76 HA2 HA6 72 78 HA3 HA5 73 79 HA4 HA2 75 80 HA5 HA3 77 82 HA6 HA1 79 83 HS1 HS4 82 87 HS2 HS3 83 88 HS3 HS2 84 89 HG1 HG1 85 90 HG2 HG2 86 91 SU1 SU1 87 93 SU6 SU3 89 94 SU7 SU5 90 96 SU8 SU6 91 97 SU9 SU4 93 98 SU10 SU8 95 100 SU11 SU2 96 101 SA1 SA5 98 102 SA3 SA2 99 103 SA6 SA4 101 104 SS1 SS1 102 106 SS3 SS2 103 107 SS4 SS3 104 108 SS6 SS6 106 109 SG1 SG1 108 111 SG3 SG2 110 113}}

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