View source for ONS-2016-008, Response to Request for Additional Information Regarding the License Amendment Request to Change to Emergency Plan to Upgrade Emergency Action Levels Based on NEI 99-01, Revision 6

{{Adams
| number = ML16041A237
| issue date = 02/04/2016
| title = Response to Request for Additional Information Regarding the License Amendment Request to Change to Emergency Plan to Upgrade Emergency Action Levels Based on NEI 99-01, Revision 6
| author name = Batson S
| author affiliation = Duke Energy Carolinas, LLC
| addressee name = 
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000269, 05000270, 05000287
| license number = DPR-038, DPR-047, DPR-055
| contact person = 
| case reference number = 2015-04, ONS-2016-008
| document type = Legal-Affidavit, Response to Request for Additional Information (RAI)
| page count = 260
| project = 
| stage = Response to RAI
}}

=Text=
{{#Wiki_filter:SCOTT L. BATSON DUI<EVice                                                                     President Oconee Nuclear Station ENERGYS                                                                                   Duke Energy ONOI VP I 7800 Rochester Hwy Seneca, SC 29672 o: 864-873-3274
: f. 864-873-4208 Scott.Batson@duke-energy.com ONS-201 6-008 10 CFR 50.90 10 CFR 50 Appendix E February 4, 2016 Attn: Document Control Desk U. S. Nuclear Regulatory Commission 11555 Rockville Pike Rockville, MD 20852 Duke Energy Carolinas, LLC (Duke Energy)
Oconee Nuclear Station, Units 1, 2, and 3 Docket Numbers 50-269, 50-270, and 50-287 Renewed Operating License Nos. DPR-38, DPR-47, and DPR-55

==Subject:==
Response to Request for Additional Information Regarding the License Amendment Request (LAR) to Change the Oconee Nuclear Station (ONS)
Emergency Plan to Upgrade ONS Emergency Action Levels Based on NEI 99-01, Revision 6 License Amendment Request No. 2015-04 By letter dated June 23, 2015 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML15183A060), Duke Energy requested approval of a proposed change to the Emergency Action Levels (EALs) used at Oconee Nuclear Station (ONS). Duke Energy proposes to revise their current ONS EAL scheme to one based upon Nuclear Energy Institute (NEI) document NEI 99-01, "Development of Emergency Action Levels for Non-Passive Reactors," Revision 6 (ADAMS Accession No. ML12326A805).
The NRC staff reviewed the request and determined that additional information is needed to complete their review. A letter requesting additional information was sent on December 15, 2015 (ADAMS Accession No. ML15345A398). Enclosure 1Ito this letter provides the ONS response to the request for additional information. Enclosure 2 provides the ONS Emergency Action Level Technical Bases Document, Revision 0, that incorporates changes in response to the request for additional information and additional changes deemed necessary, as described in Enclosure 1.
This letter makes no new commitments or changes to any existing commitments. Should you have any questions regarding this request, please contact Chris Wasik, Regulatory Affairs Manager, at (864) 873-5789.

U.S. Nuclear Regulatory Commission February 4, 2016 Page 2 I declare under penalty of perjury that the foregoing is true and correct. Executed on the 4th day of February, 2016.
Sincerely, Scott L. Batson Vice President Oconee Nuclear Station I

==Enclosures:==
: 1. Response to Request for Additional Information
: 2. Oconee Nuclear Station Emergency Action Level Technical Bases Document, "Emergency Action Level Technical Bases" (Clean Version), Revision 0

U.S. Nuclear Regulatory Commission February 4, 2016 Page 3 cc w/ 

==Enclosures:==

Ms. Catherine Haney, Regional Administrator U. S. Nuclear Regulatory Commission - Region II Marquis One Tower 245 Peachtree Center Ave., NE, Suite 1200 Atlanta, GA 30303-1257 Mr. James Hall, Senior Project Manager (by electronic mail only)
Off ice of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission 11555 Rockville Pike Mail Stop 0-8 G9A Rockville, MD 20852 Mr. Jeffrey Whited, Project Manager (by electronic mail only)
Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission 11555 Rockville Pike Mail Stop 0-8 B1A Rockville, MD 20852 Mr. Eddy Crowe NRC Senior Resident Inspector Oconee Nuclear Site Ms. Susan Jenkins, Manager Radioactive & Infectious Waste Management Division of Waste Management South Carolina Department of Health and Environmental Control 2600 Bull St.
Columbia, SC 29201

U.S. Nuclear Regulatory Commission February 4, 2016 Page 4 cc w/o 

==Enclosures:==

Mr. Scott Krein, Director Oconee County Emergency Preparedness 415 South Pine Street Walhalla, SC 29691 Ms. Denise Kwiatek, Deputy Director Pickens County Emergency Preparedness 1509 Walhalla Highway Pickens, SC 29671 Mr. Kim Stenson, Director South Carolina Emergency Management Division 2779 Fish Hatchery Road West Columbia, SC 29172

ONS-201 6-008 Enclosure 1 ENCLOSURE I RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION LICENSE AMENDMENT REQUEST NO. 2015-04 OCONEE NUCLEAR POWER STATION, UNIT 1, 2, AND 3 DOCKET NOS 50-269, 50-270, and 50-287 RENEWED LICENSE NOS. DPR-38, DPR-47, and DPR-55 14 Pages Follow

ONS-201 6-008 Enclosure 1 Page 1 of 14 By letter dated June 23, 2015 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML15183A060), Duke Energy requested approval of a proposed change to the Emergency Action Levels (EALs) used at Oconee Nuclear Station (ONS) which would revise the current ONS EAL scheme to one based upon Nuclear Energy Institute (NEI) document NEI 99-01, "Development of Emergency Action Levels for Non-Passive Reactors," Revision 6 (ADAMS Accession No. ML12326A805). The NRC staff reviewed the request and determined that additional information is needed to complete their review. A letter requesting additional information was sent on December 15, 2015 (ADAMS Accession No. ML15345A398).
Duke Energy provides the following response to the request for additional information (RAI) regarding the License Amendment Request to revise the ONS EAL Scheme to one based on NEI 99-01, Revision 6. Changes to the ONS Technical Bases Document as a result of this RAI are identified within the ONS RAI Response in the table below. Changes were also made to the ONS Technical Bases Document that were not the result of an RAI. These additional changes are described in a table below the RAl response, with an explanation of why each change was deemed necessary. The revised ONS EAL Technical Bases Document is provided in Enclosure 2 with revision bars indicating changes from the original submittal.
RAI-ONS-#                                  Question                                             ONS Response Section 4.3, "Instrumentation Used for EALs," to NEI 99-01,    ONS has confirmed that setpoints and indications used in the proposed EAL Revision 6, states, in part, that: "Scheme developers should   scheme are within the calibrated range(s) of the stated instrumentation and ensure that specific values used as EAL setpoints are within   that the resolution of the instrumentation is appropriate for the the ca/ibratedrange of the referenced instrumentation."        setpoint/indication.
01        Please confirm that all setpoints and indications used in the proposed EAL scheme are within the calibrated range(s) of the stated instrumentation and that the resolution of the instrumentation is appropriate for the setpoint/indication.
Section 2.5, 'Technical Bases Information," of Duke's LAR     The ONS site specific and NEI 99-01    generic bases sections have been states, in part, that: "A Plant-specific basis section that   combined into a single bases section   for each EAL. Section 2.5 "Technical provides ONS-relevant information concerning the EAL.         Bases Information" has been revised    accordingly.
This is followed by a Generic basis section that provides a   Reudnbaswreplibehvbeneet.
description of the rationale for the EAL as provided in NEI   Rdnatbss              hr plcbe           aebedltd 02        99-01 Rev. 6." Due to the high probability that EAL decision-makers may be confused between these two sections when the information appears to be inconsistent, please justify the rational for two sections when it is acceptable to just have one basis section that is specific to the plant, or revise accordingly to eliminate potential confusion by user.

ONS-201 6-008 Enclosure 1 Page 2 of 14 RAI.ONS-#                               Question                                             ONS Response In Section 5.0, "Definitions," of Duke's LAR, the definition  Site Area Emergency was verified consistent with the NEI 99-01, Revision 6 for Site Area Emergency is not consistent with the definition definition.
in the NRC-endorsed guidance. Please revise accordingly, or provide further justification for inconsistency with the   Also, after further review, determined that the following definitions should be 03     NRC-endorsed guidance.                                        included in Section 5.0 to be consistent with NEI 99-01 Revision 6:
* Emergency Action Level (E-AL),          --
* Emergency Classification Level (ECL),
* Fission Product Barrier Threshold, and
* Initiating Condition For the following EALs, please explain why the listed         Deleted Note 3 from EALs RAI1.2, RS1 .2 and RG1 .2 04     NOTEs were included, or revise accordingly..,
* RAI.2- NOTE-3
* RSI.2-NOTE-3
_______          RG1.2  - NOTE-3

0NS-201 6-008 Enclosure 1 Page 3 of 14 RAI-ONS-#                           Question                                              ONS Response For EALs RU1, RA1, RS1 and RG1, please explain why the       The proposed ONS RGI .1, RSI .1 and RA1 .1 EAL threshold values are based values and thresholds developed for these EALs are           on NEI 99-0 1 Rev 6 and utilize the site specific URI/RASCAL dose inconsistent with what has already been approved for ONS,    assessment model (v2.0.0.1) for their development.
or revise accordingly.                                       The proposed ONS RUI.1 EAL threshold value is based on NEI 99-01 Rev 6 and utilizes 2 times the 0DCM limit for its development.
The current ONS RG1 .1 and RSI .1 EALs are based on NUMARC/NESP-007 Rev 2.
05 The current ONS RAI.1 EAL threshold value is-based on NUMARC/NESP-007 Rev 2 and utilizes 200 times the 0DCM limit for its development.
The current ONS RUI.1 EAL threshold value is based on NUMARC/NESP-007 Rev 2 and utilizes 2 times the 0DCM limit for its development.
The table below provides a comparison between the current and proposed ONS gaseous effluent EAL thresholds for the GE, SAE, Alert and Unusual Event emergency classification levels:
Monitor             GE                   SAE                      Alert                Unusual Event RA-6(p)NEP:            3.OOE+5  NEI:           3.O0E+4   NEI:          3.OOE+3            N/A RAI46(cm)NUMARC:       2.09E+I6 NUJMARC:       2.09E+5   NUMARC: 2.09E+4 RAI-45 (cpm)         N/A                   N/A                      N/A             NEI:          1.41E+5 NUJMARC:      9.35E+5

ONS-201 6-008 Enclosure 1 Page 4 of 14 RAI-ONS-#  Question                    ONS Response RGI.1 Differences The NEI 99-01 Rev 60ONS RGI.1 EAL threshold value utilizes the RASCAL based dose assessment model assuming a reactor accident source term and meteorology reflective of conditions consistent with a design basis accident and definition of the General Emergency classification level. The NUMARC/NESP-007 Rev 20ONS RGI .1 EAL threshold value utilizes an annual average X/Q in accordance with the NRC endorsed "Methodology for Development of Emergency Action Levels NUMARC/NESP-007 Revision 2 Questions and Answers," dated June 1993. The X/Q input alone accounts for a difference of more than one order of magnitude. Differences in assumed source term mix also contribute to the difference in threshold values. The basis and development methodologies for RGI.1 are completely different between 05,             NUMARC/NESP-007 Rev 2 and NEI 99-01 Rev 6.
continued RSI .1 Differences Both NEI 99-01 Rev 6 and the NUMARC/NESP-007 Rev 2 RS1I.1 EAL threshold values are established as 1/10 the RGI.1 threshold values. Thus, the differences described above for RGI.1 are the same for RS1.1.
RAI .1 Differences NEI 99-01 Rev 6 established the basis for RAI.1 as 1/10 the RSI.1 threshold value. NUMARC/NESP-007 Rev 2 established the basis for RA1.1 as 200 times the ODCM limit. The basis and development methodologies for RAI .1 are completely different between NUMARC/NESP-007 Rev 2 and NEI 99-01 Rev 6.

ONS-201 6-008 Enclosure 1 Page 5 of 14 RAI-ONS-#                             Question .ONS                                              Response RU1. 1 Differences Both the NEI 99-01 Rev 6 and the NUMARC/NESP-007 Rev 20ONS RUI .1 EAL threshold values are derived from the same the ODCM equation:
S~obody (cpm) =   4r2 f         *i(i 500*     ~ C          +bkg
                                                                                                                                    +_3
* Eq, RU 1.1 Differences, continued 05,                                                                                                         --
continued                                                                Both the NEI 99-01 Rev 6 and the NUMARC/NESP-007 Rev 20ONS RUI .1 EAL threshold values use the same input values with the exception of source term.
* The NUMARC/NESP-007 Rev 20ONS RUI.1 EAL threshold value is based on Xe-133.
* The NEI 99-01 Rev 60ONS RU1.1 EAL threshold value is based on NUREG-1 940 Table 1-2 normal RCS Noble Gas term source mix.
Neither NUMARC/NESP-007 Rev 2 nor NEI 99-01 Rev 6 provides source term guidance for the RU1.1 EAL threshold value. NUREG-1940 is used for the ONS RAI.1, RSI.1 and RG1.1 radiological effluent EAL threshold source term basis. NUREG-1940 was selected as the source term basis for the RUI.1 radiological effluent EAL threshold to maintain consistency of guidance document bases across the emergency classification thresholds.
For EAL RA2.2, the information in the NEI 99-01 Basis          Re-instated the following text to the RA2.2 bases:
section does  not contain to thisallparticular of the actual EAL.information from 06    NE! 99-01  germane                              Please explain "This EAL applies to irradiated fuel that is licensed for dry storage up to the why this information was omitted, or revise accordingly,       point that the loaded storage cask is sealed. Once sealed, damage to a loaded cask causing loss of the CONFINEMENT BOUNDARY is classified in accordance with EAL EU1.1 ."

0NS-201 6-008 Enclosure 1 Page 6 of 14 RAI-ONS-#                               Question                                             ONS Response For EAL RG2.I, the level value in the EAL, -23 feet, is       Corrected RG2.I to read, 'Spent fuel pooi level cannot be restored to at least inconsistent with the stated level in the ONS Basis, -23.5     -23.5 ft. for ->60 mai. (Note 1).'
feet. Please correct the discrepancy and ensure the correct 07    ONS Level 3 value is used in this EAL, or provide justification for this difference.
For EALs RA3.2 and HA5.1, both of these EALs are               ONS will maintain two tables, R-2 and H-2 for end-user usability.
applicable to the same areas; therefore, it is not required to NR-nosdgiacidcteththeiespcfcltoflntrmsr have two separate tables (Table R-2 for EAL RA3.2, and         NR-nosdgiacincteththeiespcfcltoflntrmsr Table H-2 for EAL HA5.I). In addition, the listed areas are    areas with entry-related mode applicability should include those rooms or 08    too all-encompassing and should be refined to just those       areas that contain equipment which require a manual/local action as specified specific areas where access is required as described in the    in operating procedures used for normal plant operation, cooldown and NRC ndosedguiance Plaserevse te aeas orshutdown.               EAL Technical Bases Document, Attachment 3, Safe Operation &
provide further justification.                                 Shutdown Rooms/Areas Tables R-3 & H-2 Bases, provides analysis for the areas listed. The Oconee plant is constructed with essentially a common Turbine Building and a common Auxiliary Building for all three units. Very little separation is provided for key areas joined by common hallways. For this reason, the areas provided in Table R-2/H-2 are necessarily all-encompassing.
For the description of Category E-Independent Spent            Revised the ISFSI category introduction:
09     Fuel Storage installation (ISFSI), please add wording related to how security events at the ISFSI are declared.      "The ONS ISFSI is contained wholly within the plant Protected Area. Therefore a security event related to the ISFSI would be applicable to EALs HUI .1, HAl.land HSI.I."

ONS-201 6-008 Enclosure 1 Page 7 of 14 RAI-ONS-# JQuestion                                                                      ONS Response For EALs CU2.1 and SA1.1, please confirm that the         The CU2.1 and SAI.1 listed AC Power Sources include:
listed alternating current (AC) power sources are timely and reliable (i.e., will they be available when needed?), Offsite:
and If not, please remove them from the list.                  0     Unit Normal Transformer (backcharged)
S Unit Startup Transformer (SWYD) a     Another Unit Startup Transformer (aligned) (SWYD)
S CT5 (Central/energizing Standby Bus)
Emergency:
* Unit Startup Transformer (Keowee)
* Another Unit Startup Transformer (aligned) (Keowee)
* CT4
* CT5 (dedicated line/energizing Standby Bus) 10                                                              The timeliness of the listed AC power sources is dependent upon the required alignment as indicated by the parenthetical following the source. For example, Unit Normal Transformer (backcharged) is only credited as a viable offsite source while it is aligned and the associated unit is on backcharge such that it is capable of powering the essential busses within 15 minutes, whether or not the buses are currently powered from it.
Routine maintenance and surveillance programs are in place to ensure the reliability of the offsite and emergency power paths. The important functions of these systems are monitored under the 10 CFR 50.65 Maintenance Rule program. In addition, Operations performs routine rounds to verify these paths are operable and/or available. While the Central switchyard is outside the scope of the Maintenance Rule, it is subject to the North American Electric Reliability Corporation (NERC) reliability standards.

ONS-201 6-008 Enclosure 1 Page 8 of 14 RAI-ONS-#                              Question                                            ONS Response The intent of EALs CA2.1, 551.1, SG1.1 and SG1.2 is to       Removed table references from CA2.1, SS1.1,-SGI.1, and SG1.2 as make the appropriate emergency classification upon a         suggested.
loss of ALL power sources. However, the list as 11    developed eliminates the possibility of alternative AC power sources energizing an essential bus, thus negating the need for declaring the applicable EAL.
Please remove the table of AC power sources, or provide further justification.
For EALs CU3.1 and CA3.1, the guidance in the ONS            Added new Note 10 to CU3.1 and CA3.1:
basis related to the process to follow When core exit              'oe1:I         h bec             frlal    C eprtr          niaincue 12 thermocouples (CETCs) are unavailable/unreliable                   byt 1:Ilossnc the dea of etremvlcaablityS           clpeassreifdication should sed 12 should be carried over as a NOTE for this EAL to ensure            basdo timeos    tofboily deata."      vlcpbliy lsiiato           hudb this information is presented to EAL decision-makers (i.e., on the EAL wallboard).
For EALs CU5.1 and SU7.1, both of these EALs are             The two separate tables are provided for end-user usability.
applicable to the same areas; therefore, it is not required  EOF Phones are not available for Onsite communications. Removed EOF to have two separate tables (Table C-5 for EAL CU5.1, and Table S-4 for EAL SU7.1). Please provide basis for       PoeSse             rmOst oun listing separate tables, or revise accordingly. In addition, Dedicated satellite phones are available for Offsite communications only.
please explain the following, and if necessary, revise       Removed Satellite phones from Onsite and NRC columns.
13    teeEsacodny:There                                                  are two dedicated satellite phones. One is in the EOF Offsite Agency
* How the EOF phone system can suffice for             Communicator Area and one is in the Offsite Monitoring Area.
onsite communications?
                  *Are dedicated satellite phones available for onsite, offsite, and NRC communications?
                  *How many dedicated satellite phones are there, as it may be likely that the NRC phone will need to be dedicated for NRC communications?

ONS-201 6-008 Enclosure 1 Page 9 of 14 RAI-ONS-#                             Question                                               ONS Response For EAL HU2.1, please explain why the use of the                Revised HU2 IC and EAL to use "aBE" vs. "DBE."
abbreviation DBE (design basis earthquake) and OBE (opratng eathqake asi ar inonsstetlyuse. Wile         Removed the detailed information related to the Strong Motion Accelerometer and Tendon Gallery Peak Acceleration Recorders, as they cannot be used to the guidance eventually states that these two terms are the     clsiyaemrny(15iutidcto).
same for ONS, please revise to be consistent with theclsiyaemrny(>5iutincto).
guidance or consistent with the ONS use of the equivalent term(s), or provide further justification. Also, please explain 14    the need for the detailed information related to the Strong Motion Accelerometer and Tendon Gallery Peak Acceleration Recorders, as it could be implied that, pending approval, the staff agrees that these recorders will suffice for determining this EAL, or revise accordingly.
For EAL HU3.2, please explain why EAL SA9.1 is not              Added escalation pathway of EAL SA9.1 to EAL HU3.2 bases.
also listed, along with EAL CA6.1, to ensure that the           During review of the escalation pathways on the Wallboards, it was noted that 15  escaion pathcordi   ally prtn.mdsaeadrseo                      HA2, HA3, and HA4 incorrectly referenced HA6.1 instead of CA6.1. This has revie acordigly.been                                corrected in the Wallboards.
For EAL HU3.5, please explain, in greater detail, what          The Jocassee Dam is located upstream of the Oconee Nuclear Station.
this EAL is intended for and how it meets the definition of     Historically, Condition B associated with the failure of the Jocassee Dam was an Unusual Event. In particular, please explain what            included in the Emergency Plan because it could cause the turbine building Condition B is, and how it related to the site radiological     basement to flood, disabling important equipment (e.g., main and emergency emergency plan                                                  feedwater pumps). Escalation of the event to a higher category was based on 16                                                                    the ability to maintain core cooling or shutdown functions. With the migration to NEI 99-01, Rev 6 guidance, symptom-based effects from flooding on the radiological emergency plan would be addressed through the loss of power EALs and the Natural/Technical Hazard EALs, based on actual losses of power or safety-related equipment. FERC Emergency Planning maintains evacuation protocols for the public based on dam failures; therefore, HU3.5 was removed.

ONS-201 6-008 Enclosure I Page 10 of 14 RAI-ONS-#                           Question                        ]ONS                       Response For EAL HS3.1, please explain, in greater detail, what     In the ONS submittal to adopt the NUMARC/NESP-007, Rev. 2, this EAL is intended to do and why it is at the Site Area  Classification Scheme, dated May 5, 1994, Duke Energy identified the Emergency classification level. Has there been an NRC      concerns associated with the failure of the Keowee Hydro Dam and commitment made related to ONS dam failures?               characterized this failure as a Site Area Emergency. A failure of Keowee Hydro would result in loss of the emergency AC power supply and the potential to lose the ultimate heat sink source. Some flooding of the site may result. Evaluation of the plant status following failure of the dam would determine the need to escalate emergency classifications. More recently, in a January 15, 2010 letter to the NRC, Duke Energy communicated measures in place to address postulated external flood threat issues. Within that letter, Duke Energy referred to the Jocassee EAP, which identifies two conditions related to the status of the dam: Condition A- Failure is Imminent or Has Occurred; Condition B- Potentially Hazardous Situation is Developing. Condition A initiates a call tree that notifies offsite agencies to implement specific actions to protect/warn the public as well as notifications to the ONS Unit 2 Control Room. The letter further communicated that once notification has been received at ONS, the ONS Emergency Plan and associated response procedures are implemented. Existing plant 17                                                               procedures included guidance to trip each of the ONS units and bring the units to a safe shutdown condition.
Although no specific commitments related to ONS dam failures were identified, previous versions of the ONS Emergency Plan and ongoing flood hazard discussions have included existing actions within the Site and Hydro Emergency Plans. NRC approval of this submittal would supersede previous Emergency Plans. Symptom-based affects from flooding on the radiological emergency plan would be addressed through the loss of power EALs and the Natural/Technical Hazard EALs, based on actual losses of power or safety-related equipment. The Federal Energy Regulatory Commission (FERC), as regulator for both Jocassee and Keowee dams, has Emergency Planning requirements and oversight for evacuation protocols for the public safety based on dam failures. Therefore, HS3.1 has been removed.

ONS-201 6-008 Enclosure 1 Page 11 of 14 RAI-ONS-#                             Question                                           ONS Response For EALs HU4.1 and HU4.2, the areas listed in Table H-     Table H-I Fire Areas are based on the ONS Fire Protection design basis 1 seem to be vague or too all-encompassing. Please         specification. Table H-i Fire Areas include those structures containing explain if the listed areas are all the areas that contain functions and systems required for safe operation, shutdown and cooldown 18    equipment needed for safe operation, safe shutdown and     of the plant (SAFETY SYSTEMS).
safet cool-downiand ifothese aea~s.cnb ietndt              A balance must be established between defining major plant structures limi fo thse  E~s.containing cosideatin                                     safe shutdown equipment as fire areas versus a detailed list of areas for every safety system component location. The Table H-I list of fire areas achieves that balance in support of timely and accurate emergency classification for the end-user.
For EALs HU4.3 and HU4.4, please confirm that the          The ONS ISFSI is contained wholly within the plant Protected Area. Therefore 19 Independent Spent Fuel Storage Installation (ISFSI)        the ISFSI would be applicable to EALs HU4.3 and HU4.4 for fires within the 19 would be an area applicable to these EALs, or revise        plant Protected Area accordingly.
For EAL HS6.1, please consider adding operating mode        Revised HS6.I mode applicability from ALL to Modes 1 - 6.
speifiityto ey afey fnctonslised he               n te EL. Revised HS6.1 based upon an assessment of applicable modes for each of the listed safety function as follows:
                                                                     'An event has resulted in plant control being transferred from the Control Room to the Auxiliary Shutdown Panels or Standby Shutdown Facility (SSF) 20                                                                       AND Control of any of the following key safety functions is not reestablished within 15 min. (Note 1):
* Reactivity (Modes 1, 2 and 3 only)
* Core Cooling
                                                                            *RCS heat removal"

ONS-201 6-008 Enclosure 1 Page 12 of 14 RAI-ONS-#                            Question                                        ONS Response Frevalsae the. andudesig plase ctonfir that  di    NSa  A conference call between ONS and NRC personnel was conducted on evaluatedntheio uniqed adesign apelcabltsof thesdigiALs  1/19/2016 to gain further understanding of this question. As a result of this call, instumetaton sed an th appicailiy o thse A~s      it was determined that the question specifically pertains to the digital Reactor tonfrtheapiabl faditnluE~re md.Ino wadditinepleased        Protective System (RPS) at Oconee and whether the design of this digital cponfirmthat additiontaltEAn arenot warrnted bsdsystem           introduced failure modes which affect applicability of EALs or warrant upondigtalinsrumntaton nd ontol.additional                  EALs.
As part of the RPS and Engineered Safeguards Protective System (ESPS) replacement project, a Failure Modes and Effects Analysis (FMEA) was performed. The FMEA, reviewed by NRC staff as part of the licensing submittal for the RPS/ESPS replacement [ML100220016, Ref 72], consisted of a 21                                                            systematic qualitative analysis of the RPS and ESPS design with the primary objectives of identifying all credible failure modes, evaluating the consequence and effects of failures, and verifying that the design satisfies single-failure criterion applicable to the replacement RPS/ESPS.
While the design of the Oconee RPS is now digital, the functions remain the same. Credible failures within the system have been analyzed. These failures cause the system to fail in a predefined safe state, as was true for the original analog RPS system. Based on an assessment of the FMEA analysis there is no impact on the scope of SU3.1 or SA3.1. The EALs for Oconee that the RPS is applicable to still apply as written, and no new initiating criteria are deemed to be warranted.

ONS-201 6-008 Enclosure 1 Page 13 of 14 RAI-ONS-#                           Question                                           ONS Response 22    For Table F-2 in the Fission Barrier Matrix, please       The values presented in Table F-2 were derived from Table 2 of the Oconee provide the information related to how this table was     Core Damage Assessment Guidelines (OSC-5283) as follows:
developed.                                                FC Loss Column FC Loss represents the expected Containment High Range Radiation Monitor (RIA-57 & RIA-58) response based on a LOCA for various periods of time after SID with 4% clad damage. The values in Table F-2 in this column were calculated first by multiplying the 100% clad damage values in Table 2 of OSC-5283 by .04 and then rounded to account for variation in monitor readings due to decay.
CMT Potential Loss Column CMT Potential Loss represents the expected Containment High Range Radiation Monitor (RIA-57 & RIA-58) response based on a LOCA for various periods of time after SID with 20% clad damage. The values in Table F-2 in this column were calculated first by multiplying the 100% clad damage values in Table 2 of OSC-5283 by 0.2 and then rounded to account for variation in monitor readings due to decay.
Under the Fission Product Barrier (FPB) Matrix, the cited As per response to RAI-ONS-02, the ONS site specific and NEI 99-01 NEI 99-01 Basis sections for several of the FPB criteria  Revision 6 bases have been unified.
23    are not from NRC-endorsed guidance (NEI 99-01, Revision 6). Please either revise to what has actually been endorsed, or (depending on respond to RAI-02) consider unifying the basis sections into one.

ONS-201 6-008 Enclosure 1 Page 14 of 14 Summary of EAL Changes Not Associated with RAI Response The table below summarizes the changes that have been incorporated into the EAL Technical Bases Document contained in Enclosure 2 that are not involved in the NRC RAI response.
EAL #                                                                         Description HUI .1      Separated HU1 .1 into HU1 .1, HU1 .2 and HU1 .3 to remove the "OR" qualifier in the NEI 99-01, Revision 6 template, as follows:
                        "HU1 .1 A SECURITY CONDITION that does not involve a HOSTILE ACTION as reported by the Site Security Supervision" "HU1 .2 Notification of a credible security threat directed at the site" "HUl1.3 A validated notification from the NRC providing information of an aircraft threat" HA1 .1      Separated HA1 .1 into HA1 .1 and HA1 .2 to remove the "OR" qualifier in the NEI 99-01, Revision 6 template, as follows:
                       "HAl .1 A HOSTILE ACTION is occurring or has occurred within the OWNER CONTROLLED AREA as reported by the Security Shift Supervision" "HA1 .2 A validated notification from NRC of an aircraft threat within 30 mai. of the site" RAl1.4      Administrative change to correct the definition for Site Boundary by changing "DPC" to "Duke Energy".
RG1.3       Administrative change to update ONS Basis Reference 1Ito be AD-EP-ALL-0203 which has superseded SH/0/B/2005/002.
SAl.1       Administrative change to correct the Mode Applicability to read "4-Hot Shutdown" instead of "3-Hot Shutdown".
SA6.1/SS6.1     Administrative change to correct the Subcategory to read "6 - RPS Failure" instead of "2 - RPS Failure".
HA7.1/HS7.1/    Administrative change to update the Operations Shift Manager acronym to be "SM" instead of "OSM".
HG7.1

ENCLOSURE 2 ONS EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT, REVISION 0 OCONEE NUCLEAR POWER STATION, UNIT 1, 2, AND 3 DOCKET NOS 50-269, 50-270, and 50-287 RENEWED LICENSE NOS. DPR-38, DPR-47, and DPR-55 240 Pages Follow

*,*DUKE         OCONEE NCERSTATION EMERGENCY ACTION LEVEL TECHNICAL BASES Revision 0 2/4/16 S[Document No.1         Rev. 0       Page 1 of 240

Table of Contents SECTION                                                                                                     PAGE 1.0  PURPOSE........................................................................................... 3 2.0 DISCUSSION ........................................................................................       3 2.1 Background........................................................................................      3 2.2 Fission Product Barriers .......................................................................... 4 2.3 Fission Product Barrier Classification Criteria...................................................       4 2.4 EAL Organization .................................................................................      5 2.5 Technical Bases Information......................................................................       7 2.6 Operating Mode Applicability ....... .............................................................       8 3.0 GUIDANCE ON MAKING EMERGENCY CLASSIFICATIONS .................................. 9 3.1 General Considerations........................................................................... 9 3.2 Classification M ethodology.........          J........................................10

==4.0 REFERENCES==
 ....................................  ................................................        13 4.1 Developmental................................................................................... 13 4.2 Implementing..................................................................................... 13 5.0 DEFINITIONS, ACRONYMS & ABBREVIATIONS............................................... 14 5.1 Definitions ........................................................................................ 14 5.2 Acronyms and Abbreviations .................................................................... 19 6.0  ONS TO NEI 99-01 Rev. 6 EAL CROSS-REFERENCE........................................ 22 7.0  ATTACHMENTS................................................................................... 26 1      Emergency Action Level Technical Bases .......................................... :.....27 Cateaorv R Abnormal Rad Release I Rad Effluent .............................. 27 Catecqorv E ISFSI................................................................... 63 Catecqorv C Cold Shutdown / Refueling System Malfunction.................... 66
_Cateaqorv H Hazards .............................................................. 101
_Cate~qorv S System Malfunction ................................................. 145 Cate~Qorv F Fission Product Barrier Degradation............................... 184 2      Fission Product Barrier Loss / Potential Loss Matrix and Bases ........................................................................ 189 3     Safe Operation & Shutdown Rooms/Areas Tables R-2 & H-2 Bases ............... 235 l[Document No.]                        1Rev.                  0                  [Page                 2 of 240

1.0      PURPOSE This document provides an explanation and rationale for each Emergency Action Level (EAL) included in the EAL Upgrade Project for Oconee Nuclear Station (ONS). It should be used to facilitate review of the ONS EALs and provide historical documentation for future reference.
Decision-makers responsible for implementation of RPI0/A/1 000/001, Emergency Classification, may use this document as a technical reference in support of EAL interpretation.
This information may assist the Emergency Coordinator in making classifications, particularly those involving judgment or multiple events. The basis information may also be useful in training and for explaining event classifications to off-site officials.
The expectation is that emergency classifications a~re to be made as soon as conditions are present and recognizable for the classification, but within 15 minutes or less in all cases of conditions present. Use of this document for assistance is not intended to delay the emergency classification.
Because the information in a basis document can a~ffect emergency classification decision-making (e.g., the Emergency Coordinator refers to it during an event), the NRC staff expects that changes to the basis document will be evaluated in accordance with the provisions of 10 CFR 50.54(q).
2.0      DISCUSSION 2.1      Background EALs are the plant-specific indications, conditions or instrument readings that are utilized to classify emergency conditions defined in the ONS Emergency Plan.
In 1992, the NRC endorsed NUMARC/NESP-007 "Methodology for Development of Emergency Action Levels" as an alternative to NUREG-0654 EAL guidance.
NEI 99-01 (NUMARC/NESP-007) Revisions 4 and 5 were subsequently issued for industry implementation. Enhancements over earlier revisions included:
* Consolidating the system malfunction initiating conditions and example emergency action levels which address conditions that may be postulated to occur during plant shutdown conditions.
* Initiating conditions and example emergency action levels that fully address conditions that may be postulated to occur at permanently Defueled Stations and Independent Spent Fuel Storage Installations (ISFSIs).
* Simplifying the fission product barrier EAL threshold for a Site Area Emergency.
Subsequently, Revision 6 of NEI 99-01 has been issued which incorporates resolutions to numerous implementation issues including the NRC EAL Frequently Asked Questions (FAQs).
Using NEI 99-01 Revision 6, "Methodology for the Development of Emergency Action Levels for Non-Passive Reactors," November 2012 (ADAMS Accession Number ML12326A805) (ref.
4.1.1), ONS conducted an EAL implementation upgrade project that produced the EALs discussed herein.
I[Document No.]                        IRev.0                                           Page 3of 240

2.2     Fission Product Barriers Fission product barrier thresholds represent threats to the defense in depth design concept that precludes the release of radioactive fission products to the environment. This concept relies on multiple physical barriers, any one of which, if maintained intact, precludes the release of significant amounts of radioactive fission products to the environment.
Many of the EALs derived from the NEI methodology are fission product barrier threshold based. That is, the conditions that define the EALs are based upon thresholds that represent the loss or potential loss of one or more of the three fission product barriers. "Loss" and "Potential Loss" signify the relative damage and threat of damage to the barrier. A "Loss" threshpld means the barrier no longer assures contain mentlof radioactive materials. A "Potential Loss" threshold implies an increased probability of barrier loss and decreased certainty of maintaining the barrier.
The primary fission product barriers are:
A. Fuel Clad (FC): The Fuel Clad Barrier consists of the cladding material that contains the fuel pellets.
B. Reactor Coolant System (RCS): The RCS Barrier includes the RCS primary side and its connections up to and including the pressurizer safety and relief valves, and other connections up to and including the primary isolation valves.
C. Containment (CMT): The Containment (Reactor Building) Barrier includes the Reactor Building and connections up to and including the outermost containment isolation valves. This barrier also includes the main steam, feedwater, and blowdown line extensions outside the Reactor Building up to and including the outermost secondary side isolation valve. Containment Barrier thresholds are used as criteria for escalation of the Emergency Classification Level (ECL) from Alert to a Site Area Emergency or a General Emergency.
2.3     Fission Product Barrier Classification Criteria The following criteria are the bases for event classification related to fission product barrier loss or potential loss:
Alert:
Any loss or any potential loss of either Fuel Clad or RCS Barrier Site Area Emergqency:
Loss or potential loss of any two barriers General Emergency:
Loss of any two barriersand loss or potentialloss of the third barrier S[Document No.]                                 Rev. 0                                  Page 4 of 240

2.4     EAL Organization The ONS EAL scheme includes the following features:
* Division of the EAL set into three broad groups:
o EALs applicable under any plant operating modes - This group would be reviewed by the EAL-user any time emergency classification is considered.
o EALs applicable only under hot operating modes - This group would only be reviewed by the EAL-user when the plant is in Hot Shutdown, Hot Standby, Startup, or Power Operation mode.
o E'ALs applicable only under col operating modes - This group would only be reviewed by the EAL-user when the plant is in Cold Shutdown, Refueling mode or No Mode.
The purpose of the groups is to avoid review of hot condition IEALs when the plant is in a cold condition and avoid review of cold condition EALs when the plant is in a hot condition. This approach significantly minimizes the total number of EALs that must be reviewed by the EAL-user for a given plant condition, reduces EAL-user reading burden and, thereby, speeds identification of the EAL that applies to the emergency.
* Within each group, assignment of EALs to categories and subcategories:
Category and subcategory titles are selected to represent conditions that are operationally significant to the EAL-user. The ONS EAL categories are aligned to and represent the NEI 99-01 "Recognition Categories." Subcategories are used in the ONS scheme as necessary to further divide the EALs of a category into logical sets of possible emergency classification thresholds. The ONS EAL categories and subcategories are listed below.
I[Document No.]                                   Rev. 0                               Page 5 of 240

EAL Groups, Categories and Subcategories EAL Group/Category                               EAL Subcategory Any Operating Mode:
R - Abnormal Rad Levels / Rad Effluent   1 - Radiological Effluent 2- Irradiated Fuel Event 3- Area Radiation Levels H - Hazards and Other Conditions         1 - Security, Affecting Plant Safety               2- Seismic Ekent 3 - Natural or Technological Hazard 4- Fire 5 - Hazardous Gas 6 -Control Ro~om Evacuation 7 - Emergency Coordinator Judgment E - Independent Spent Fuel Storage       1 - Confinement Boundary Installation (ISFSI)
Hot Conditions:
S - System Malfunction                   1 - Loss of Essential AC Power 2 - Loss of Vital DC Power 3 - Loss of Control Room Indications 4-  RCS  Activity 5-  RCS  Leakage 6-  RPS  Failure 7-  Loss of Communications 8 - Containment Failure 9 - Hazardous Event Affecting Safety Systems F - Fission Product Barrier Degradation  None Cold Conditions:
C - Cold Shutdown I Refueling System     1 - RCS Level Malfunction                          2 - Loss of Essential AC Power 3-  RCS Temperature 4-  Loss of Vital DC Power 5-  Loss of Communications 6-  Hazardous Event Affecting Safety Systems The primary tool for determining the emergency classification level is the EAL Classification Matrix. The user of the EAL Classification Matrix may (but is not required to) consult the EAL Technical Bases in order to obtain additional information concerning the EALs under classification consideration. The user should consult Section 3.0 and Attachments 1 & 2 of this document for such information.
I[Document No.]                           I        Rev. 0                                    Page 6 of 240

2.5     Technical Bases Information EAL technical bases are provided in Attachment 1 for each EAL according to EAL group (Any, Hot, Cold), EAL category (R, C, H, 5, E and F) and EAL subcategory. A summary explanation of each category and subcategory is given at the beginning of the technical bases discussions of the EALs included in the category. For each EAL, the following information is provided:
Category Letter & Title Subcateqiory Number & Title Initiating Condition (IC)
Site-specific description of the generic IC given in NEI 99-01 Rev. 6.
EAL Identifier (enclosed in rectanqle)
Each EAL is assigned a unique identifier to support accurate communication of the emergency classification to onsite and offsite personnel. Four characters define each EAL identifier:
: 1. First character (letter): Corresponds to the EAL category as described above (R, C, H, S, EorF)
: 2. Second character (letter): The emergency classification (G, 5, A or U)
G = General Emergency S = Site Area Emergency A = Alert U = Unusual Event
: 3. Third character (number): Subcategory number within the given category.
Subcategories are sequentially numbered beginning with the number one (1). If a category does not have a subcategory, this character is assigned the number one (1).
: 4. Fourth character (number): The numerical sequence of the EAL within the EAL subcategory. If the subcategory has only one EAL, it is given the number one (1).
Classification (enclosed in rectangle):
Unusual Event (U), Alert (A), Site Area Emergency (S) or General Emergency (G)
EAL (enclosed in rectanqle)
Exact wording of the EAL as it appears in the EAL Classification Matrix Mode Applicability One or more of the following plant operating conditions comprise the mode to which each EAL is applicable: 1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown, 5   -
Cold Shutdown, 6 - Refueling, NM      - No Mode, or Any. (See Section 2.6 for operating mode definitions)
Definitions:
If the EAL wording contains a defined term, the definition of the term is included in this section. These definitions can also be found in Section 5.1.
[Document No.]                                   Rev. 0                               Page 7 of 240

Basis:
A basis section that provides ONS-relevant information concerning the EAL as well as a description of the rationale for the EAL as provided in NEI 99-01 Rev. 6.
ONS Basis Reference(s):
Site-specific source documentation from which the EAL is derived 2.6     Operating Mode Applicability (ref. 4.1.6) 1     Power Operation Keff>~ 0.99 and reactor thermal power > 5%
2     Startup      i Keff > 0.99 and reactor thermal power < 5%
3     Hot Standby Keff < 0.99 and average coolant temperature > 250&deg;FI 4     Hot Shutdown Keff < 0.99 and average coolant temperature 250&deg;F    > Tavg > 2000 F and all reactor vessel head closure bolts fully tensioned 5     Cold Shutdown Keff < 0.99 and average coolant temperature < 200&deg;F and all reactor vessel head closure bolts fully tensioned 6     Refuelingq One or more reactor vessel head closure bolts less than fully tensioned NM No Mode Reactor vessel contains no irradiated fuel The plant operating mode that exists at the time that the event occurs (prior to any protective system or operator action being initiated in response to the condition) should be compared to the mode applicability of the EALs. If a lower or higher plant operating mode is reached before the emergency classification is made, the declaration shall be based on the mode that existed at the time the event occurred.
I[Document No.]                         IRev.         0            IPage                        8 of 240

3.0     GUIDANCE ON MAKING EMERGENCY CLASSIFICATIONS 3.1     General Considerations When making an emergency classification, the Emergency Coordinator must consider all information having a bearing on the proper assessment of an Initiating Condition (IC3). This includes the Emergency Action Level (EAL) plus the associated Operating Mode Applicability, Notes, and the informing basis information. In the Recognition Category F matrices, EALs are based on loss or potential loss of Fission Product Barrier Thresholds.
3.1.1    Classification Timeliness NRC3 regulations require the licensee to establish and maintain the capability to assess, classify, and declare an emergency condition within 15 minutes after the availability of indications to plant operators that an emergency action level has been exceeded and to promptly declare the emergency condition as soon as possible following identification of the appropriate emergency classification level. The NRC3 staff has provided guidance, on implementing this requirement in NSIR/DPR-ISG-01, "Interim Staff Guidance, Emergency Planning for Nuclear Power Plants" (ref. 4.1.9).
3.1.2 Valid Indications All emergency classification assessments shall be based upon valid indications, reports or conditions. A valid indication, report, or condition, is one that has been verified through appropriate means such that there is no doubt regarding the indicator's operability, the condition's existence, or the report's accuracy.
An indication, report, or condition is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment.
3.1 .3 Imminent Conditions For 1I3s and EALs that have a stipulated time duration (e.g., 15 minutes, 30 minutes, etc.), the Emergency Coordinator should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the condition has exceeded, or will likely exceed, the applicable time. If an ongoing radiological release is detected and the release start time is unknown, it should be assumed that the release duration specified in the 1IC/EAL has been exceeded, absent data to the contrary.
3.1.4 Planned vs. Unplanned Events A planned work activity that results in an expected event or condition which meets or exceeds an EAL does not warrant an emergency declaration provided that: 1) the activity proceeds as planned, and 2) the plant remains within the limits imposed by the operating license. Such activities include planned work to test, manipulate, repair, maintain or modify a system or component. In these cases, the controls associated with the planning, preparation and execution of the work will ensure that compliance is maintained with all aspects of the operating license provided that the activity proceeds and concludes as expected. Events or conditions of this type may be subject to the reporting requirements of 10 &sect; CFR 50.72 (ref.
4.1.4).
I[Document No.]                        *.          Rev. 0                                 Page 9 of 240

3.1 .5 Classification Based on Analysis The assessment of some EALs is based on the results of analyses that are necessary to ascertain whether a specific EAL threshold has been exceeded (e.g., dose assessments, chemistry sampling, RCS leak rate calculation, etc.). For these EALs, the EAL wording or the associated basis discussion will identify the necessary analysis. In these cases, the 15-minute declaration period starts with the availability of the analysis results that show the threshold to be exceeded (i.e., this is the time that the EAL information is first available). The NRC expects licensees to establish the capability to initiate and complete EAL-related analyses within a reasonable period of time (e.g., maintain the necessary expertise on-shift).
3.1.6 Emergency Coordinator Judgment While the EALs have been developed to address a full spectrum of possible events and conditions which may warrant emergency classification, a provision for classification based on operator/management experience and judgment is still necessary. The NEI 99-01 EAL scheme provides the Emergency Coordinator with the ability to classify events and conditions based upon judgment using EALs that are consistent with the ECL definitions (refer to Category H).
The Emergency Coordinator will need to determine if the effects or consequences of the event or condition reasonably meet or exceed a particular ECL definition. A similar provision is incorporated in the Fission Product Barrier Tables; judgment may be used to determine the status of a fission product barrier.
3.2     Classification Methodology To make an emergency classification, the user will compare an event or condition (i.e., the relevant plant indications and reports) to an EAL(s) and determine if the EAL has been met or exceeded. The evaluation of an EAL must be consistent with the related Operating Mode Applicability and Notes. If an EAL has been met or exceeded, the associated IC is likewise met, the emergency classification process "clock" starts, and the ECL must be declared in accordance with plant procedures no later than fifteen minutes after the process "clock" started.
When assessing an EAL that specifies a time duration for the off-normal condition, the "clock" for the EAL time duration runs concurrently with the emergency classification process "clock."
For a full discussion of this timing requirement, refer to NSlRIDPR-ISG-01 (ref. 4.1.9).
3.2.1   Classification of Multiple Events and Conditions When multiple emergency events or conditions are present, the user will identify all met or exceeded EALs. The highest applicable ECL identified during this review is declared. For example:
* If an Alert EAL and a Site Area Emergency EAL are met, whether at one unit or at two different units, a Site Area Emergency should be declared.
There is no "additive" effect from multiple EALs meeting the same ECL. For example:
* If two Alert EALs are met, whether at one unit or at two different units, an Alert should be declared.
Related guidance concerning classification of rapidly escalating events or conditions is provided in Regulatory Issue Summary (RIS) 2007-02, Clarificationof NRC Guidance for Emergency Notifications During Quickly Changing Events (ref. 4.1.2).
[Document No.]                                    Rev. 0                               Page 10. of 240

3.2.2 Consideration of Mode Changes During Classification The mode in effect at the time that an event or condition occurred, and prior to any plant or operator response, is the mode that determines whether or not an IC is applicable. If an event or condition occurs, and results in a mode change before the emergency is declared, the emergency classification level is still based on the mode that existed at the time that the event or condition was initiated (and not when it was declared). Once a different mode is reached, any new event or condition, not related to the original event or condition, requiring emergency classification should be evaluated against the ICs and EArs applicable to the operating mode at the time of the new event or condition.
For events that occur in Cold Shutdo~vn or Refueling, escalation is via EALs that are appliq*able in the Cold Shutdown or Refueling modes, even if Hot Shutdown (or a higher mode) is entered during the subsequent plant response. In particular, the fission product barrier EALs are applicable only to events that initiate in the Hot Shutdown mode or higher.
3.2.3 Classification of Imminent Cornditions Although EALs provide specific thresholds, the Emergency Coordinator must remain alert to events or conditions that could lead to meeting or exceeding an EAL within a relatively short period of time (i.e., a change in the ECL is IMMINENT). If, in the judgment of the Emergency Coordinator, meeting an EAL is IMMINENT, the emergency classification should be made as if the EAL has been met. While applicable to all emergency classification levels, this approach is particularly important at the higher emergency classification levels since it provides additional time for implementation of protective measures.
3.2.4 Emergency Classification Level Upgrading and Downgrading An ECL may be downgraded when the event or condition that meets the highest IC and EAL no longer exists, and other site-specific downgrading requirements are met. If downgrading the ECL is deemed appropriate, the new ECL would then be based on a lower applicable IC(s) and EAL(s). The ECL may also simply be terminated.
As noted above, guidance concerning classification of rapidly escalating events or conditions is provided in RIS 2007-02 (ref. 4.1.2).
3.2.5 Classification of Short-Lived Events Event-based ICs and EALs define a variety of specific occurrences that have potential or actual safety significance. By their nature, some of these events may be short-lived and, thus, over before the emergency classification assessment can be completed. If an event occurs that meets or exceeds an EAL, the associated ECL must be declared regardless of its continued presence at the time of declaration. Examples of such events include an earthquake or a failure of the reactor protection system to automatically trip the reactor followed by a successful manual trip.
3.2.6 Classification of Transient Conditions Many of the ICs and/or EALs employ time-based criteria. These criteria will require that the IC/EAL conditions be present for a defined period of time before an emergency declaration is warranted. In cases where no time-based criterion is specified, it is recognized that some transient conditions may cause an EAL to be met for a brief period of time (e.g., a few seconds to a few minutes). The following guidance should be applied to the classification of these conditions.
I[Document No.]                                   Rev. 0                                Page11o24

EAL    momentarily met during           met durinq an expected         expected (normal)        plant response - In instances where an EAL is briefly plant response, an emergency declaration is not warranted provided that associated systems and components are operating as expected, and operator actions are performed in accordance with procedures.
EAL momentarily met but the condition is corrected prior to an emergency declaration     -  Ifan operator takes prompt manual action to address a condition, and the action is successful in correcting the condition prior to the emergency declaration, then the applicable EAL is not considered met and the associated emergency declaration is not required. For illustrative purposes, consider the following example:
An ATWS occurs and the high pressur* ECCS systems fail to automatically start.
Reactor vessel level rapidly decreases and the plant enters an inadequate core cooling condition (a potential loss of both the Fuel Clad and RCS Barriers). If an operator manually starts a high pressure ECCS system in accordance with an EOP step and clears the inadequate core cooling condition prior to an emergency declaration, then the classification should be based on the ATWS only.
It is important to stress that the 15-minute emergency classification assessment period (process clock) is not a "grace period" during which a classification may be delayed to allow the performance of a corrective action that would obviate the need to classify the event.
Emergency classification assessments must be deliberate and timely, with no undue delays.
The provision discussed above addresses only those rapidly evolving situations when an operator is able to take a successful corrective action prior to the Emergency Coordinator completing the review and steps necessary to make the emergency declaration. This provision is included to ensure that any public protective actions resulting from the emergency classification are truly warranted by the plant conditions.
3.2.7 After-the-Fact Discovery of an Emergency Event or Condition In some cases, an EAL may be met but the emergency classification was not made at the time of the event or condition. This situation can occur when personnel discover that an event or condition existed which met an EAL, but no emergency was declared, and the event or condition no longer exists at the time of discovery. This may be due to the event or condition not being recognized at the time or an error that was made in the emergency classification process.
In these cases, no emergency declaration is warranted; however, the guidance contained in NUREG-1 022 (ref. 4.1.3) is applicable. Specifically, the event should be reported to the NRC in accordance with 10 CFR &sect; 50.72 (ref. 4.1.4) within one hour of the discovery of the undeclared event or condition. The licensee should also notify appropriate State and local agencies in accordance with the agreed upon arrangements.
3.2.8 Retraction of an Emergency Declaration Guidance on the retraction of an emergency declaration reported to the NRC is discussed in NUREG-1022 (ref. 4.1.3).
S[Document No.]                                    Rev. 0                             Page 12 of 240

==4.0   REFERENCES==

4.1   Developmental 4.1.1    NEI 99-0 1 Revision 6, Methodology for the Development of Emergency Action Levels for Non-Passive Reactors, ADAMS Accession Number ML12326A805 4.1.2  'RIS 2007-02 Clarification of NRC Guidance for Emergency Notifications During Quickly Changing Events, February 2, 2007.
4.1.3    NUREG-1022 Event Reporting Guidelines: 10CFR5O.72 and 50.73 4.1.4    10 &sect; CFR 50.72 Immediate Notification Requirements for Operating Nuclear Power Reactors 4.1.5    10 &sect; CFR 50.73 License Event Report System 4.1.6   Technical Specifications Table 1.1-1 Modes 4.1.7    OP/i1,2 ,3/A/1502/000 Containment Clo~sure Control 4.1.8    Procedure Writer's Manual, Revision 012 4.1.9    NSIR/DPR-ISG-01 Interim Staff Guidance; Emergency Planning for Nuclear Power Plants 4.1.10   Oconee Nuclear Site Emergency Plan 4.1.11   S.D.1.3.5 Shutdown Protection Plan 4.1.12   Duke Energy Physical Security Plan for ONS 4.2   Implementing 4.2.1 RPIOIA/1 000/00 1 Emergency Classification 4.2.2 NEI 99-01 Rev. 6 to ONS EAL Comparison Matrix 4.2.30ONS EAL Matrix
[Document No.]                              Rev. 0                             Page 13 of 240

5.0     DEFINITIONS, ACRONYMS & ABBREVIATIONS 5.1     Definitions (ref. 4.1 .1 except as noted)
Selected terms used in Initiating Condition and Emergency Action Level statements are set in all capital letters (e.g., ALL CAPS). These words are defined terms that have specific meanings as used in this document. The definitions of these terms are provided below.
Alert Events are in progress, or have occurred, which involve an actual or potential substantial degradation of the level of safety of the plant or a security event that involves probable life threatenirqg risk to site personnel or damage to site equipment *ecause of hostile action. Any releases are expected to be small fractions of the EPA Protective Action Guideline exposure levels.
Confinement Barrier The barrier(s) between spent fuel and the environment once the spent fuel is processed for dry storage. As related to the ONS ISFSI, Confinement Boundary is comprised of the DSC (dry shielded canister) shell, inner bottom cover plate, inner top cover plate, siphon & vent block, siphon & vent port cover plate, and the welds that join them together.
Containment Closure The action to secure containment and its associated structures, systems, and components as a functional barrier to fission product release under-all plant conditions up to and including a loss of decay heat removal or fuel handling accident inside containment (ref. 4.1 .11).
As applied to ONS, Containment Closure is established when the requirements of OP/1,2,3/A/1502/000, Containment Closure Control, are met (ref. 4.1.7).
Emergency Action Level (EAL)
A pre-determined, site-specific, observable threshold for an Initiating Condition that, when met or exceeded, places the plant in a given emergency classification level.
Emergency Classification Level (ECL)
One of a set of names or titles established by the US Nuclear Regulatory Commission (NRC) for grouping off-normal events or conditions according to (1) potential or actual effects or consequences, and (2) resulting onsite and offsite response actions. The emergency classification levels, in ascending order of severity, are:
* Unusual Event (UE)
* Alert
* Site Area Emergency (SAE)
* General Emergency (GE)
EPA PAGs Environment Protection Agency Protective Action Guidelines. The EPA PAGs are expressed in terms of dose commitment: 1 Rem TEDE or 5 Rem CDE Thyroid. Actual or projected offsite exposures in excess of the EPA PAGs requires ONS to recommend protective actions for the general public to offsite planning agencies.
I[Document     NO.]                                Rev. 0          IPage                      14 of 240

Explosion A rapid, violent and catastrophic failure of a piece of equipment due to combustion, chemical reaction or overpressurization. A release of steam (from high energy lines or components) or an electrical component failure (caused by short circuits, grounding, arcing, etc.) should not automatically be considered an explosion. Such events require a post-event inspection to determine if the attributes of an explosion are present.
Faulted The term applied to a steam generator that has a steam leak on the secondary side of sufficient size to cause an uncontrolled drop in steam generator pressure or the steam generator to becom~e completely depressurized.
Fire Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred but is NOT required if large quantities of smoke and heat are observed.
Fission Product Barrier Threshold A pre-determined, site-specific, observable threshold indicating the loss or potential loss of a fission product barrier.
Flooding A condition where water is entering a room or area faster than installed equipment is capable of removal, resulting in a rise of water level within the room or area.
General Emergency Events are in progress or have occurred which involve actual or imminent substantial core degradation or melting with potential for loss of containment integrity or hostile actions that result 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.
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 ONS 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 ONS. 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.
[Document No.]                                    Rev.O0                              Page15 &deg;f240O

Imminent The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions.
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.
Impede(d)
Personnel access to a room or area is hindered to an extent that extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., requiring use of protective equipment, such as SCBAs, that is not routinely employed).
Initiating Condition (IC)
An event or condition that aligns with the definition of one of the four emergency classification levels by virtue of the potential or actual effects or consequences.
Intrusion The act of entering without authorization. Discovery of a bomb in a specified area is indication of intrusion into that area by a hostile force.
Maintain Take appropriate action to hold the value of an identified parameter within specified limits.
Normal Levels As applied to radiological IC/EALs, the highest reading in the past twenty-four hours excluding the current peak value.
Owner Controlled Area Area outside the PROTECTED AREA fence that immediately surrounds the plant. Access to this area is generally restricted to those entering on official business.
Projectile An object directed toward a Nuclear Power Plant that could cause concern for its continued operability, reliability, or personnel safety.
Protected Area That part of the nuclear station property consisting of the Reactor, Auxiliary, Turbine, and Service Building and grounds, contained within the owner controlled security fence (ref.
4.1.10).
RCS Intact The RCS should be considered intact when the RCS pressure boundary is in its normal condition for the cold shutdown mode of operation (e.g., no freeze seals or nozzle dams, pressurizer manway and safeties installed).
Reduced Inventory Condition with fuel in the reactor vessel and the level lower than approximately three feet below the reactor vessel flange (RCS level < 50" on LT-5) (ref. 4.1.11 ).
S[Document No.]                                   Rev. 0                              Page 16 of 240

Refueling Pathway The spent fuel pool and/or fuel transfer canal comprise the refueling pathway.
Restore Take the appropriate action required to return the value of an identified parameter to the applicable limits.
Ruptured The condition of a steam generator in which primary-to-secondary leakage is of sufficient magnitude to require a safety injection.
Safety SystemII A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2):I Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:
(1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.
Security Condition Any security event as listed in the approved security contingency plan that constitutes a threat/compromise to site security, threat/risk to site personnel, or a potential degradation to the level of safety of the plant. A security condition does not involve a hostile action.
Site Area Emergency Events are in progress or have occurred which involve actual or likely major failures of plant functions needed for protection of the public or hostile actions that result 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 Guidelines exposure levels beyond the site boundary.
Site Boundary That area, including the PROTECTED AREA, in which Duke Energy has the authority to control all activities including exclusion or removal of personnel and property (1 mile radius) from the center of Unit 2. (ref. 4.1.10).
Unisolable An open or breached system line that cannot be isolated, remotely or locally.
Unplanned A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.
((Document No.]                                   Rev. 0           IPage                       17 &deg;f240

Unusual Event Events are in progress or have occurred which indicate a potential degradation in the level of safety of the plant or indicate a security threat to facility protection has been initiated. No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of safety systems occurs.
Valid An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the condition's existence, or the report's accurac~y is removed, Implicit in this definition is the need for timely assessment.
Visible Damage Damage to a component or structure that is' readily observable without measurements, testing, or analysis. The visual impact of the damage is sufficient to cause concern regarding the operability or reliability of the affected component or structure.
.[Document NO.]                                   Rev. 0                                  Page 18 of 240

5.2      Acronyms and Abbreviations 0F .......................................................................... Degrees Fahrenheit
 &deg;..................................................................................... Degrees AC .......................................................................... Alternating Current AP............................................................. Abnormal Operating Procedure ATWS................................................... Anticipated Transient Without Scram BWST ........................................................... Borated Water Storage Tank CETC ...............................................                         4................. Core Exit Thermocouple CDE............................................................... Committed Dose Equivalent CFR ............................................................. Code of Federal Regulations CMT ...........................................                              ..................................... Containment DBA..................................................................... Design Basis Accident DBE.................................................................. Design Basis Earthquake DC ................................................................................                               Direct Current DSC ..................................................................... Dry Shielded Canister EAL ...........           i........................................................ Emergency Action Level ECCS....................................................... Emergency Core Cooling System ECL........................................................... Emergency Classification Level EOF ........................................................... Emergency Operations Facility EOP........................................................ .Emergency Operating Procedure EPA ........................................................ Environmental Protection Agency ERG.......................................................... Emergency Response Guideline EPIP .............................................. Emergency Plan Implementing Procedure ESF ................................................................ Engineered Safety Feature FAA............................                   ,............................... Federal Aviation Administration FBI............................................................ Federal Bureau of Investigation FEMA ............................................. Federal Emergency Management Agency GE......................................................................... General Emergency HPI..................................................................... High Pressure Injection IC............................................................................ Initiating Condition IPEEE............. Individual Plant Examination of External Events (Generic Letter 88-20)
ISFSI ........................................... Independent Spent Fuel Storage Installation Keff ...................................................... Effective Neutron Multiplication Factor LCO.............................................................. Limiting Condition of Operation
[Document No.]                                                    Rev. 0                     *IPage                             19 of 240

LEO ................................................................. Law Enforcement Center LER...................................................................... Licensee Event Report LOCA ................................................................ Loss of Coolant Accident LWR ....................................................................... Light Water Reactor MPC ......................... Maximum Permissible Concentration/Multi-Purpose Canister mR, mRem, mrem, mREM.................................. milli-Roentgen Equivalent Man MSL ........................................................................... Main Steam Line iMW............................................                      *....................................... Megawatt NEI.................................................................... Nuclear Energy Institute NESP................................................. National Environmental Studies Project SNM ............................................                   !......................................... No Mode NPP....................................................................... Nuclear Power Plant NRC.......................................................... Nuclear Regulatory Commission NORAD..................................... North American Aerospace Defense Command (NO)UE.......................................................... Notification of Unusual Event OBE .............................................................. Operating Basis Earthquake OCA .....................................................  .........             ..... Owner Controlled Area ODCM....................................................... Off-site Dose Calculation Manual ORO........................................................... Offsite Response Organization PA ...............................................................................                     Protected Area PAG................................................................ Protective Action Guideline PRA ............................................................ Probabilistic Risk Assessment PSA .. ....................................................... Probabilistic Safety Assessment PWR............................................................... Pressurized Water Reactor PSIG .......................................................... Pounds per Square Inch Gauge PSW.................................................................. Protected Service Water R .......................................................................................                     Roentgen RCS .................................................................. Reactor Coolant System Rem, rem, REM ................................................... Roentgen Equivalent Man Rep CET........................................... Representative Core Exit Thermocouples RETS ......................................... Radiological Effluent Technical Specifications RPS ................................................................ Reactor Protective System RV................................................................................ Reactor Vessel RVLIS ............................................... Reactor Vessel Level Indicating System I[Document NO.]                                            Rev. 0                                                 Page 20 of 240

SAR ...................................................................... Safety Analysis Report SBO.............................................................................. Station Blackout SCBA..................................................... Self-Contained Breathing Apparatus SG.............................................................................. Steam Generator SLCO............................................................ Selected License Commitment SPDS........................................................ Safety Parameter Display System SRO .................................................................... Senior Reactor Operator TEDE . .. .L..........................................................................Total Effective Dose Equivalent TSC................................................................... Technical Support Center UFSAR.................................................. Updated Final Safety Analysis Report S[Document No.].                                                     Rev. 0                                    Page 2l of 240

6.0   ONS-TO-NEI 99-01 Rev. 6 EAL CROSS-REFERENCE This cross-reference is provided to facilitate association and location of an ONS EAL within the NEI 99-0 1 IC/EAL identification scheme. Further information regarding the development of the ONS EALs based on the NEI guidance can be found in the EAL Comparison Matrix.
ONS           NEI 99-01 Rev. 6 EAL            IC        Example EAL RU1.1           AU1          1, 2 RU1.2          AU1           3 RU2.1          AU2           1 RA1.1          AA1           1 RA1.2          AA1           2 RA1.3          AA1           3 RA1.4          AA1           4 RA2.1          AA2           1 RA2.2          AA2           2 RA2.3          AA2           3 RA3.1          AA3           1 RA3.2          AA3           2 RSI.1          AS1           1 RS1.2          AS1           2 RS1.3          AS1           3 RS2.1          AS2           1 RGI.1           AG1           1 RG1.2           AG1           2 RG1.3           AG1           3 RG2.1           AG2           1 CU1.1           CU1           1 CU1.2           CU1           2 S[Document No.]                                  Rev. 0                            Page 22 of 240

ONS      NEI 99-01 Rev. 6 EAL  IC       Example EAL ICEAL CU2.1     CU2           1 CU3.1     CU3           1 CU3.2     CU3          2 CU4.1     CU4           1 CU5.1     CU5        1,2, 3 CAl.1     CA1           1 CA1.2     CAl          2 CA2.1     CA2           1 CA3.1     CA3         1, 2 CA6.1     CA6           1 CS1.1     CS1          3 CGI.1     CG1          2 FAI.1     FA1           1 FSI.1     FS1           1 FGI.1     FG1           1 HUI.1     HU1           1 HU1.2     HU1          2 HU1.3     HlU1          3 HU2.1     HU2           1 HU3.1     HU3           1 HU3.2     HU3          2 HU3o3     HU3           3 HU3.4     HU3          4 HU4.1     HU4           1 HU4.2     HU4          2
[Document No.]            Rev. 0           Page 23 of 240

ONS        NEI 99-01 Rev. 6 EAL  IC       Example EAL ICEAL HU4.3      HU4           3 HU4.4      HU4           4 HU7.1      HU7           1 HAl.1      HA1           1 HA1.2      HA1           2 HA*5.1     HA5           1 HA6.1      HA6           1 HA7.1      HA7           1 HSl.1      HSl           1 HS6.1      HS6           I HS7.1      HS7           1 HG1.1      HG1           1 HG7.1      HG7           1 SUI.1      SU1           1 SU3.1      SU2           1 SU4.1      SU3           2 SU5.1      SU4        1, 2, 3 SU6.1      SU5           1 SU6.2      SU5           2 SU7.1      SU6        1,2, 3 SU8.1      SU7          1,2 SA1.1      SA1           1 SA3.1      SA2           1 SA6.1      SA5           1 SA9.1      SA9           1 S[Document No.]             Rev. 0            Page 24 of 24.0

ONS      NEI 99-01 Rev. 6 EAL  IC       Example EAL ICEAL SS1.1     SS11 SS2.1     SS8          1 SS6.1     SS5          1 SG1.1     SG1          1 SG1.2     SG8          1 EUI.1     EU1          1
*I[Document No.]           Rev. 0          IPage 25 of 240

7.0   ATTACHMENTS 7.1   Attachment 1, Emergency Action Level Technical Bases 7.2   Attachment 2, Fission Product Barrier Matrix and Ba'sis 7.3   Attachment 3, Safe Operation & Shutdown Areas Tables R-2 & H-2 Bases S[Document No.]                             Rev. 0                         Page 26 of 240

ATTACHMENT 1 EAL Bases Category R - Abnormal Rad Release / Rad Effluent EAL Group:       ANY (EALs in this category are applicable to any plant condition, hot or cold.)
Many EALs are based on actual or potential degradation of fission product barriers because of the elevated potential for offsite radioactivity release. Degradation of fission product barriers though is not always apparent via non-radiological symptoms. Therefore, direct indication of elevated radiological effluents or area radiation levels are appropriate symptoms for emlergency classification.
At lower levels, abnormal radioactivity releases may be indicative of a failure of containment systems or precursors to more significant releases. At higher release rates, offsite radiological conditions may result which require offsite protective actions. Elevated area radiation levels in the plant may also be indicative of the failure of containment systems or preclude access to plant vital equipment necessary to ensure plant safety.
Events of this category pertain to the following subcategories:
: 1. Radiological Effluent Direct indication of effluent radiation monitoring systems provides a rapid assessment mechanism to determine releases in excess of classifiable limits. Projected offsite doses, actual offsite field measurements or measured release rates via sampling indicate doses or dose rates above classifiable limits.
: 2. Irradiated Fuel Event Conditions indicative of a loss of adequate shielding or damage to irradiated fuel may preclude access to vital plant areas or result in radiological releases that warrant emergency classification.
: 3. Area Radiation Levels Sustained general area radiation levels which may preclude access to areas requiring continuous occupancy also warrant emergency classification.
[Document No.]                                  Rev. 0                             Page 27 of 240

ATTACHMENT 1 EAL Bases Category:                          R - Abnormal Rad Levels / Rad Effluent Subcategory:                       1 - Radiological Effluent Initiating Condition:              Release of gaseous or liquid radioactivity greater than 2 times the SLCITS limits for 60 minutes or longer EAL:
RUI.1                Unusual Event Reading on an~y Table R-1 effluent radiation monitor > column "UE"ifor --60 min.
(Notes 1,2, 3)
Note 1:       The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
Note 2:       If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.
Note 3:       If the effluent flow past an effluent monitor is known to have stopped, indicating that the release path is isolated, the effluent monitor reading is no longer VALID for classification purposes.
Table R-.1      Effluent Monitor Classification Thresholds_____]
Release Point            [Monitor [            GE        [       SAE         [     Alert       ]        UE     ]
  =, Unit 112/3 Plant Vent              RIA-45            .......                                            1.41 E+5 cpm 0
Unit 1/2/3 Plant Vent             RIA-46       3.O0E+5 cprm       3.00E+4 cpm       3.00E+3 cpm              ..
  ,m. Liquid Radwaste Discharge        RIA-33            .......                                            4.79E+5 cpm Mode Applicability:
All Definition(s):
None Basis:
The column "UE" release values in Table R-1 represent two times the appropriate SLC and Technical Specification release rate and concentration limits associated with the specified monitors (ref. 1, 2,3, 4, 5,6).
Gaseous Releases Instrumentation that may be used to assess this EAL: (ref. 1):
      . Unit 1/2/3 Plant Vent Noble Gas Low Monitor                -  RIA-45(L)
S[Document No.]                                              Rev. 0                                  Page 28 of 240

ATTACHMENT 1 EAL Bases Liquid Releases Instrumentation that may be used to assess this EAL: (ref. 1):
      *Liquid Radwaste Discharge Monitor     -  RIA-33 (batch release)
This IC addresses a potential decrease in the level of safety of the plant as indicated by a low-level radiological release that exceeds regulatory commitments for an extended period of time (e.g., an uncontrolled release). It includes any gaseous or liquid radiological release, monitored or un-monitored, including those for which a radioactivity discharge permit is normally prepared.        II Nuclear power plants incorporate design features intended to control the release of radioactive effluents to the environment. Further, there are administrative controls established to prevent unintentional releases, and to control and monitor intentional releases. The occurrence of an extended, uncontrolled radioactive release to the environment is indicative of degradation in these features and/or controls.
Radio logical effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.
Classification based on effluent monitor readings assumes that a release path to the environment is established. Ifthe effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.
Releases should not be prorated or averaged. For example, a release exceeding 4 times release limits for 30 minutes does not meet the EAL.
This EAL addresses normally occurring continuous radioactivity releases from monitored gaseous or liquid effluent pathways.
Escalation of the emergency classification level would be via IC RAl.
ONS Basis Reference(s):
: 1. UFSAR Section 11.5, Process and Effluent Radiological Monitoring and Sampling Systems
: 2. Oconee Nuclear Station Units 1, 2 and 30Offsite Dose Calculation Manual
: 3. ONS-SLC 16.11.1 Radioactive Liquid Effluents
: 4. ONS-SLC 16.11.2 Radioactive Gaseous Effluents
: 5. EP-EALCALC-ONS-1401 ONS Radiological Effluent EAL Values, Rev. 0
: 6. Technical Specification Section 5.5.5
: 7. NEI 99-01 AU1
[Document No.]                                 Rev. 0                            Page 29 of 240

ATTACHMENT 1 EAL Bases Category:                    R  - Abnormal Rad Levels / Rad Effluent Subcategory:                 1 - Radiological Effluent Initiating Condition:        Release of gaseous or liquid radioactivity greater than 2 times the SLCITS limits for 60 minutes or longer.
EAL:
RUI.2            Unusual Event Sample analysis for a gaseous or liquid release indicates a concentration or release rate
> 2 x SLC/TS limits for > 60 mmn. (Notes 1, 2)
Note 1:   The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.I Note 2:   If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.
Mode Applicability:
All Definition(s):
None Basis:
This IC addresses a potential decrease in the level of safety of the plant as indicated by a low-level radiological release that exceeds regulatory commitments for an extended period of time (e.g., an uncontrolled release). It includes any gaseous or liquid radiological release, monitored or un-monitored, including those for which a radioactivity discharge permit is normally prepared.
Nuclear power plants incorporate design features intended to control the release of radioactive effluents to the environment. Further, there are administrative controls established to prevent unintentional releases, and to control and monitor intentional releases. The occurrence of an extended, uncontrolled radioactive release to the environment is indicative of degradation in these features and/or controls.
Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.
Releases should not be prorated or averaged. For example, a release exceeding 4 times release limits for 30 minutes does not meet the EAL.
This EAL addresses uncontrolled gaseous or liquid releases that are detected by sample analyses or environmental surveys, particularly on unmonitored pathways (e.g., spills of radioactive liquids into storm drains, heat exchanger leakage in river water systems, etc.).
Escalation of the emergency classification level would be via IC RAI.
[Document No.1                                        Rev. 0             IPage                       30 of 2401

ATTACHMENT 1 EAL Bases ONS Basis Reference(s):
: 1. UFSAR Section 11.5, Process and Effluent Radiological Monitoring and Sampling Systems
: 2. Oconee Nuclear Station Units 1, 2 and 30Offsite Dose Calculation Manual
: 3. ONS-SLO 16.11.1 Radioactive Liquid Effluents
: 4. ONS-SLC 16.11.2 Radioactive Gaseous Effluents
: 5. AD-RP-ALL-2003 Investigation of Unusual Radiological Occurrences
: 6. NEI 99-01 AU1
[Document No.]                            Rev. 0                           Page 31 of 240.

ATTACHMENT 1 EAL Bases Category:                        R - Abnormal Rad Levels / Rad Effluent Subcategory:                     1 - Radiological Effluent Initiating Condition:            Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid ODE EAL:
RA1.1              Alert Reading on any Table R-1 effluent radiation monitor > column "ALERT" for > 15 min.
(Notes 1, 2, 3,4)
Note 1:     The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
Note 2:     If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.
Note 3:     If the effluent flow past an effluent monitor is known to have stopped, indicating that the release path is isolated, the effluent monitor reading is no longer VALID for classification purposes.
Note 4      The pre-calculated effluent monitor values presented in EALs RAI.1, RSI.1 and RGI.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.
Table R-1       Effluent Monitor Classification Thresholds                 _____
        - Release Paint            ]Monitor              GE        (       SAE        ]      Alert       ]        UE
  = Unit 1/2/3 Plant Vent              RIA-45            .........                                          1.41 E+5 cpm
    *=Unit 1/2/3 Plant Vent            RIA-46       3.00E+5 cpm        3.00E+4 cpm       3.00E+3 cpm              ..
  *- Liquid Radwaste Discharge         RIA-33            .........                                           4.79E+5 cpm Mode Applicability:
All Definition(s):
None Basis:
This EAL addresses gaseous radioactivity releases, that for whatever reason, cause effluent radiation monitor readings corresponding to site boundary doses that exceed either:
* 10mRem TEDE
* 50 mRem ODE Thyroid I[Document No.]                                             Rev. 0           IPage                          32 of 240I

ATTACHMENT 1 EAL Bases The column "ALERT" gaseous effluent release values in Table R-1 correspond to calculated doses of 1% (10% of the SAE thresholds) of the EPA Protective Action Guidelines (TEDE or ODE Thyroid) (ref. 1, 2, 3, 4).
Instrumentation that may be used to assess this EAL: (ref. 1):
* Unit 1/2/3 Plant Vent Noble Gas Medium Monitor- RIA-46(M)
This IC addresses a release of gaseous or liquid radioactivity that results in projected or actual offsite doses greater than or equal to 1% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude represent an actual or potential substantial degradation of the level of safety of the plant as indicated by a radiological release that significantly exceeds regulatory limits (e.g., a significant uncontrolled release).
Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.
The TEDE dose is set at 1% of the EPA PAG of 1,000 mrem while the 50 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid ODE.
Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.
Escalation of the emergency classification level would be via IC RSI.
ONS Basis Reference(s):
: 1. UFSAR Section 11.5 Process and Effluent Radiological Monitoring and Sampling Systems
: 2. Oconee Nuclear Station Units 1, 2 and 30Offsite Dose Calculation Manual
: 3. EP-EALCALC-ONS-1401 ONS Radiological Effluent EAL Values, Rev. 0
: 4. SDQA-70400-COM, "Unified RASCAL Interface (URI)"
: 5. NEI 99-01 AA1 I[Document No.]                        I        Rev. 0                               Page 33 of 240I

ATTACHMENT 1 EAL Bases Category:                    R - Abnormal Rad Levels / Rad Effluent Subcategory:                 1 - Radiological Effluent Initiating Condition:        Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid CDE EAL:
RA1.2           Alert Dose assessment using actual meteorology indicates dos~es > 10 mrem TEDE or 50 mrem thyroid ODE at or beyond the SITE BOUNDARY (Note 4)
Note 4:    The pre-calculated effluent monitor values presented in EALs RAIl,; RSI.l and RGI.1 should be used Sfor emergency classification assessments until the results from a dose assessment using actual meteorology are available.
Mode Applicability:
All Definition(s):
SITE BOUNDARY- That area, including the PROTECTED AREA, in which Duke Energy has the authority to control all activities including exclusion or removal of personnel and property (1 mile radius) from the center of Unit 2.
Basis:
Dose assessments are performed by computer-based methods (ref. 1, 2).
This IC addresses a release of gaseous or liquid radioactivity that results in projected or actual offsite doses greater than or equal to 1% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude represent an actual or potential substantial degradation of the level of safety of the plant as indicated by a radiological release that significantly exceeds regulatory limits (e.g., a significant uncontrolled release).
Radiological effluent EAL~s are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.
The TEDE dose is set at 1% of the EPA PAG of 1,000 mrem while the 50 mrem thyroid ODE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid ODE.
Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.
Escalation of the emergency classification level would be via IC RSI.
S[Document Np.]                                       Rev. 0                               Page 34 of 240

ATTACHMENT 1 EAL Bases ONS Basis Reference(s):
: 1. RP/0/A/1000/001 Emergency Classification
: 2. AD-EP-ALL-0202 Emergency Response Offsite Dose Assessment
: 3. NEI 99-01 AA1
((Document No.]                           Rev. 0              Page 35 of 240

ATTACHMENT 1 EAL Bases Category:                    R - Abnormal Rad Levels / Rad Effluent Subcategory:                 1 - Radiological Effluent Initiating Condition:        Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid ODE EAL:
RA1.3            Alert Analysis of a liquid effluent sample indicates a concentration or release ratle that would result in doses > 10 mrem TEDE or 50 mrem thyroid ODE at or beyond the SITE BOUNDARY for 60 mai. of exposure (Notes 1, 2)
Note 1:   The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
Note 2:   If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.
Mode Applicability:
All Definition(s):
SITE BOUNDARY - That area, including the PROTECTED AREA, in which Duke Energy has the authority to control all activities including exclusion or removal of personnel and property (1 mile radius) from the center of Unit 2.
Basis:
Dose assessments based on liquid releases are performed per Offsite Dose Calculation Manual (ref. 1).
This IC addresses a release of gaseous or liquid radioactivity that results in projected or actual offsite doses greater than or equal to 1% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude represent an actual or potential substantial degradation of the level of safety of the plant as indicated by a radiological release that significantly exceeds regulatory limits (e.g., a significant uncontrolled release).
Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.
The TEDE dose is set at 1% of the EPA PAG of 1,000 mrem while the 50 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.
Classification based on effluent monitor readings assumes that a release path to the environment is established. Ifthe effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.
[Document No.]                                       Rev. 0                                  Page.36 of 240I

ATTACHMENT 1 EAL Bases Escalation of the emergency classification level would be via IC RS1.
ONS Basis Reference(s):
: 1. Oconee Nuclear Station Units 1, 2 and 30Offsite Dose Calculation Manual
: 2. NEI 99-01 AA1 S[Document No.]                               Rev. 0                        Page 37 of 24.0

ATTACHMENT 1 EAL Bases Category:                     R - Abnormal Rad Levels / Rad Effluent Subcategory:                  1 - Radiological Effluent Initiating Condition:         Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid CDE EAL:
RA1.4             Alert Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY:
* Closed window dose rates > 10 mR/hr expected to continue for > 60 min.
* Analyses of field survey samples indicate thyroid CDE > 50 mrem for 60 min. of inhalation.
(Notes 1,2)
Note 1:   The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
Note 2:    If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.
Mode Applicability:
All Definition(s):
SITE BOUNDARY- That area, including the PROTECTED AREA, in which Duke Energy has the authority to control all activities including exclusion or removal of personnel and property (1 mile radius) from the center of Unit 2.
Basis:
AD-EP-ALL-0203 Protocol for the Field Monitoring Coordinator During Emergency Conditions provides guidance for emergency or post-accident radiological environmental monitoring (ref.
1).
This IC addresses a release of gaseous or liquid radioactivity that results in projected or actual offsite doses greater than or equal to 1% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude represent an actual or potential substantial degradation of the level of safety of the plant as indicated by a radiological release that significantly exceeds regulatory limits (e.g., a significant uncontrolled release).
Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.
The TEDE dose is set at 1% of the EPA PAG of 1,000 mrem while the 50 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.
S[Document No.]                                       Rev. 0                                 Page 38 of 240I

ATTACHMENT 1 EAL Bases Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.
Escalation of the emergency classification level would be via IC RSI.
ONS Basis Reference(s):
: 1. AD-EP-ALL-0203 Protocol for the Field Monitoring Coordinator During Emergency Conditions
: 2. NEI 99-01 AA1 S[Document No.1                                 Rev. 0                            Page  39 of 2401

ATTACHMENT 1 EAL Bases Category:                        R - Abnormal Rad Levels / Rad Effluent Subcategory:                     1 - Radiological Effluent Initiating Condition:            Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem thyroid CDE EAL:
RSI.1              Site Area Emergency Reading on any Table R-1 effluent radiation monitor > column "SAE" for ->15 min.
(Notes 1, 2,3, 4)
Note 1:     The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
Note 2:     If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.
Note 3:     If the effluent flow past an effluent monitor is known to have stopped, indicating that the release path is isolated, the effluent monitor reading is no longer VALID for classification purposes.
Note 4:     The pre-calculated effluent monitor values presented in EALs RAI.1, RSI.1 and-RG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.
Table R-1        Effluent Monitor Classification Thresholds_____
__   Release Point            IMonitor I             GE          I     SAE       I       Alert      f        UE
    =Unit 1/2/3 Plant Vent             RIA-45             .........                                         1.41 E+5 cpm t*Unit 1/2/3 Plant Vent            RIA-46        3.00E+5 cpm        3.00E+4 cpm       3.00E+3 cpm--
 "0
  ~.mLiquid Radwaste Discharge         RIA-33             ..........-                                       4.79E+5 cpm Mode Applicability:
All Definition(s):
None I[Do~cument No.]                                             Rev. 0           IPage                         40 of 240

ATTACHMENT 1 EAL Bases Basis:
This EAL addresses gaseous radioactivity releases, that for whatever reason, cause effluent radiation monitor readings corresponding to site boundary doses that exceed either:
* 100mRemTEDE
* 500 mRem ODE Thyroid The column "SAE" gaseous effluent release value in Table R-1 corresponds to calculated doses of 10% of the EPA Protective Action Guidelines (TEDE or ODE Thyroid) (ref. 1, 3).
Instrumentation that may be used to assess this EAL: (ref. 2):
* Unit 1/2/3 Plant Vent Noble Gas Medium Monitor       - RIA-46(M)
This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to 10% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public.
Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.
The TEDE dose is set at 10% of the EPA PAG of 1,000 mrem while the 500 mrem thyroid ODE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid ODE.
Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.
Escalation of the emergency classification level would be via IC RG1.
ONS Basis Reference(s):
I. EP-EALCALC-ONS-1401 ONS Radiological Effluent EAL Values, Rev. 0
: 2. UFSAR Section 11.5 Process and Effluent Radiological Monitoring and Sampling Systems
: 3. SDQA-70400-COM, "Unified RASCAL Interface (URl)"
: 4. NEI 99-01 AS1
[Document No.]                                 Rev. 0                            Page 41of 240O

ATTACHMENT 1 EAL Bases Category:                   R - Abnormal Rad Levels / Rad Effluent Subcategory:                1 - Radiological Effluent Initiating Condition:       Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem thyroid CDE EAL:
Note 4:   The pre-calculated effluent monitor values presented in EALs RAI .1, RS1 .1 and RGI .1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.
Mode Applicability:
All Definition(s):
SITE BOUNDARY- That area, including the PROTECTED AREA, in which Duke Energy has the authority to control all activities including exclusion or removal of personnel and property (1 mile radius) from the center of Unit 2.
Basis:
Dose assessments are performed by computer-based methods (ref. 1, 2).
This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to 10% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public.
Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both planit condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.
The TEDE dose is set at 10% of the EPA PAG of 1,000 mrem while the 500 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.
Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.
Escalation of the emergency classification level would be via IC RG1.
S[Document No.]                                      Rev. 0                                 Page 42 of 240

ATTACHMENT 1 EAL Bases ONS Basis Reference(s):
: 1. RP/0/A/1 000/001 Emergency Classification
: 2. AD-EP-ALL-0202 Emergency Response Offsite Dose Assessment
: 3. NEI 99-01 AS1 l[Document No.]                            Rev. 0   1Page     43 of 240

ATTACHMENT 1 EAL Bases Category:                     R - Abnormal Rad Levels / Rad Effluent Subcategory:                  1 - Radiological Effluent Initiating Condition:         Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem thyroid ODE EAL:
RS1.3             Site Area Emergency Field survey results indicate IEITHER of the following at or beyond the SITE BOUNDARY:
* Closed window dose rates > 100 mR/hr expected to continue for > 60 min.
* Analyses of field survey samples indicate thyroid ODE > 500 mrem for 60 min. of inhalation.
(Notes 1,2)
Note 1:   The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
Note 2:    If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.
Mode Applicability:
All Definition(s):
SITE BOUNDARY- That area, including the PROTECTED AREA, in which Duke Energy has the authority to control all activities including exclusion or removal of personnel and property (1 mile radius) from the center of Unit 2.
Basis:
AD-EP-ALL-0203 Protocol for the Field Monitoring Coordinator During Emergency Conditions provides guidance for emergency or post-accident radiological environmental monitoring (ref.
1).
This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to 10% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public.
Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.
The TEDE dose is set at 10% of the EPA PAG of 1,000 mrem while the 500 mrem thyroid ODE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid ODE.
Escalation of the emergency classification level would be via IC RGI.
r[Document No.]                                        Rev. 0                                  Page 44 of 240

ATTACHMENT 1 EAL Bases ONS Basis Reference(s):
: 1. AD-EP-ALL-0203 Protocol for the Field Monitoring Coordinator During Emergency Conditions
: 2. NEI 99-01 AS1 S[Document No.]                           Rev. 0                         Page 45 of 240I

ATTACHMENT 1 EAL Bases Category:                             R - Abnormal Rad Levels / Rad Effluent Subcategory:                          1 - Radiological Effluent Initiating Condition:                 Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid ODE EAL:
RGI.1                   General Emergency Reading on any Table R-1 effluent radiation monitor > column "GE" for ->15 min.
(Notes 1, 2,3, 4)
Note 1:         The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
Note 2:          If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.
Note 3:          If the effluent flow past an effluent monitor is known to have stopped, indicating that the release path is isolated, the effluent monitor reading is no longer VALID for classification purposes.
Note 4:         The pre-calculated effluent monitor values presented in EALs RAI.1, RSI.1 and RGI.1 should be used for emergency classification assessments until the results from a dose assessment using actual.
Table R-1       Effluent Monitor Classification Thresholds                _____
         -Release       Paint          IMonitor ]            GE        [       SAE       I       Alert       ]        UE
   *oUnit 1/2/3 Plant Vent                 RIA-45            * .....                                            1.41 E+5 cpm uUnit      1/2/3 Plant Vent             RIA-46       3.00E+5 cpm        3.00E+4 cpm       3.00E+3 cpm              ..
  ~.Liquid Radwaste Discharge              RIA-33             ......                                             4.79E+5 cpm Mode Applicability:
All Definition(s):
None Basis:
This EAL addresses gaseous radioactivity releases, that for whatever reason, cause effluent radiation monitor readings corresponding to site boundary doses that exceed either:
* 1000 mRemTEDE
* 5000 mRem ODE Thyroid The column "GE" gaseous effluent release values in Table R-1 correspond to calculated doses of 100% of the EPA Protective Action Guidelines (TEDE or ODE Thyroid) (ref. 1, 3).
S[Document No.]                                                 Rev. 0                                  Page 46 of 240

ATTACHMENT 1 EAL Bases Instrumentation that may be used to assess this EAL: (ref. 2):
* Unit 1/2/3 Plant Vent Noble Gas Medium Monitor        - RIA-46(M)
This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude will require implementation of protective actions for the public.
Radiological effluent EALs are also included to provide a basis for classifying events and cqonditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.
The TEDE dose is set at the EPA PAG of 1,000 mrem while the 5,000 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.
Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.
ONS Basis Reference(s):
: 1. EP-EALCALC-ONS-1 401 ONS Radiological Effluent EAL Values, Rev. 0
: 2. UFSAR Section 11.5 Process and Effluent Radiological Monitoring and Sampling Systems
: 3. SDQA-70400-COM, "Unified RASCAL Interface (URI)"
: 4. NEI 99-01 AG1 S[Document No.]                                Rev. 0                             Page 47 of 240

ATTACHMENT 1 EAL Bases Category:                  R - Abnormal Rad Levels / Rad Effluent Subcategory:               1 - Radiological Effluent Initiating Condition:      Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid CDE EAL:
RG1.2          General Emergency Dose assessment using actual meteorology indicates doses > I1,000 mrem TEDE or 5,000 mrem thyroid CDE at or beyond the SITE BOUNDARY (Note 4)
Note 4:  The pre-calculated effluent monitor values presented in EALs RAI .1, RSI .1 and RGI .1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.
Mode Applicability:
All Definition(s):
SITE BOUNDARY- That area, including the PROTECTED AREA, in which Duke Energy has the authority to control all activities including exclusion or removal of personnel and property (1 mile radius) from the center of Unit 2.
Basis:
Dose assessments are performed by computer-based methods (ref. 1, 2).
This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude will require implementation of protective actions for the public.
Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.
The TEDE dose is set at the EPA PAG of 1,000 mrem while the 5,000 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.
Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.
I[Document No.]                                      Rev. 0                                 Page 48 of 240

ATTACHMENT 1 EAL Bases ONS Basis Reference(s):
: 1. RP/O/A/1 000/00 1 Emergency Classification
: 2. AD-EP-ALL-0202 Emergency Response Offsite Dose Assessment
: 3. NE! 99-01 AG1 S[Document No.]                            Rev. 0            Page 49 of 240

ATTACHMENT 1 EAL Bases Category:                    R  - Abnormal Rad Levels / Rad Effluent Subcategory:                 1 - Radiological Effluent Initiating Condition:        Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid ODE EAL:
RG1.3             General Emergency Field survey results indicate EITHER of the following at or beyond the *ITE BOUNDARY:
* Closed window dose rates > 1,000 mR/hr expected to continue for > 60 min.
* Analyses ofq field survey samples indicate thyroid ODE > 5,000 mrem*for 60 mmn. of inhalation.
(Notes 1,2)
Note 1:    The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
Note 2:    Ifan ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.
Mode Applicability:
All Definition(s):
SITE BOUNDARY- That area, including the PROTECTED AREA, in which Duke Energy has the authority to control all activities including exclusion or removal of personnel and property (1 mile radius) from the center of Unit 2.
Basis:
AD-EP-ALL-0203 Protocol for the Field Monitoring Coordinator During Emergency Conditions provides guidance for emergency or post-accident radiological environmental monitoring (ref.
1).
This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude will require implementation of protective actions for the public.
Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.
The TEDE dose is set at the EPA PAG of 1,000 mrem while the 5,000 mrem thyroid ODE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.
[Document No.]                                        Rev. 0                                 Page 50 of 240

ATTACHMENT 1 EAL Bases ONS Basis Reference(s):
: 1. AD-EP-ALL-0203 Protocol for the Field Monitoring Coordinator During Emergency Conditions
: 2. NEI 99-01 AGI
[Document No.]                            Rev. 0                          Page 51 of 240

ATTACHMENT 1 EAL Bases Category:                  R - Abnormal Rad Levels / Rad Effluent Subcategory:               2 - Irradiated Fuel Event Initiating Condition:      UNPLANNED loss of water level above irradiated fuel EAL:
RU2.1           Unusual Event UNPLANNED water level drop in the REFUELING PATHWAY as indicated by low water level alarm or indicationI AND UNPLANNED rise in corresponding area radiation levels as indicated by any of the following radiation monitors:
* RIA-3 RB Refueling Deck Shield Wall
* RIA-6 Spent Fuel Building Wall
* Portable area monitors on the main bridge or SFP bridge Mode Applicability:
All Definition(s):
UNPLANNED -. A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.
REFUELING PATHWAY- The spent fuel pool and/or fuel transfer canal comprise the refueling pathway.
Basis:
The spent fuel pool low water level alarm setpoint is actuated at -1.8 ft. below normal level (ref.
1). Water level restoration instructions are performed in accordance with Abnormal Operating Procedures (APs) (ref. 2).
The specified radiation monitors are those expected to see increase area radiation levels as a result of a loss of REFUELING PATHWAY inventory (ref. 3). Increasing radiation indications on these monitors in the absence of indications of decreasing water level are not classifiable under this EAL. Radiation levels in the Reactor Building refueling area are monitored by RIA-3. Radiation levels in the Spent Fuel Pool area are monitored RIA-6. When a fuel bridge is being used to handle fuel, radiation levels are monitored by a portable area monitor mounted on the bridge. (ref. 3, 4)
When the spent fuel pool and reactor cavity are connected, there could exist the possibility of uncovering irradiated fuel. Therefore, this EAL is applicable for conditions in which irradiated fuel is being transferred to and from the reactor vessel and spent fuel pool.
S[Document No.]                                 Rev. 0                            Page 52 of 240

ATTACHMENT 1 EAL Bases This IC addresses a decrease in water level above irradiated fuel sufficient to cause elevated radiation levels. This condition could be a precursor to a more serious event and is also indicative of a minor loss in the ability to control radiation levels within the plant. It is therefore a potential degradation in the level of safety of the plant.
A water level decrease will be primarily determined by indications from available level instrumentation. Other sources of level indications may include reports from plant personnel (e.g., from a refueling crew) or video camera observations (if available). A significant drop in the water level may also cause an increase in the radiation levels of adjacent areas that can be detected by monitors in those locations.        I The effects of planned evolutions should be considered. For example, a refueling bridge area radiation monitor reading may increase due to planned evolutions such as lifting of the reactor vessel head or movement of a fuel assembly. Note that this EAL is applicable only in cases where the elevated reading is due to an unplanned loss of water level.
A drop in water level above irradiated fuel within the reactor vessel may be classified in accordance Recognition Category C during the Cold Shutdown and Refueling modes.
Escalation of the emergency classification level would be via IC RA2.
ONS Basis Reference(s):
: 1. OP/I1/A/61 01/009 Alarm Response Guide 1SA-09, A-5; OP/2/A/61 02/009; OP/3/A/61 03/009
: 2. AP/I1-2,3/Ni1700/035 Loss of SPF Cooling and/or Level
: 3. UFSAR Table 12-3 Area Radiation Monitors
: 4. OP/I1,2,3/A/1 502/007, Enclosure 1, Defueling/Refueling Prerequisites
: 5. NEI 99-01 AU2 S[Document No.]                                   Rev. 0                              Page 53 of 240I

ATTACHMENT 1 EAL Bases Category:                 R - Abnormal Rad Levels / Rad Effluent Subcategory:              2 - Irradiated Fuel Event Initiating Condition:     Significant lowering of water level above, or damage to, irradiated fuel EAL:
RA2.1          Alert Uncovery of irradiated fuel in the REFUELING PATHWAY Mode Applicability:
All I Definition(s):
REFUELING PATHWAY- The spent fuel pool and/or fuel transfer canal comprise the refueling pathway.
Basis:
This IC addresses events that have caused imminent or actual damage to an irradiated fuel assembly, or a significant lowering of water level within the spent fuel pool. These events present radiological safety challenges to plant personnel and are precursors to a release of radioactivity to the environment. As such, they represent an actual or potential substantial degradation of the level of safety of the plant.
This EAL escalates from RU2.1 in that the loss of level, in the affected portion of the REFUELING PATHWAY, is of sufficient magnitude to have resulted in uncovery of irradiated fuel. Indications of irradiated fuel uncovery may include direct or indirect visual observation (e.g., reports from personnel or camera images), as well as significant changes in water and radiation levels, or other plant parameters. Computational aids may also be used (e.g., a boil-off curve). Classification of an event using this EAL should be based on the totality of available indications, reports and observations.
While an area radiation monitor could detect an increase in a dose rate due to a lowering of water level in some portion of the REFUELING PATHWAY, the reading may not be a reliable indication of whether or not the fuel is actually uncovered. To the degree possible, readings should be considered in combination with other available indications of inventory loss.
A drop in water level above irradiated fuel within the reactor vessel may be classified in accordance Recognition Category C during the Cold Shutdown and Refueling modes.
Escalation of the emergency classification level would be via IC RSI.
ONS Basis Reference(s):
: 1. AP/1 -2,3/NI1700/035 Loss of SPF Cooling and/or Level
: 2. NEI 99-01 AA2 S[Document No.]                                 Rev. 0                            Page 54 of 240

ATTACHMENT I EAL Bases Category:                R - Abnormal Rad Levels / Rad Effluent Subcategory:             2 - Irradiated Fuel Event Initiating Condition:    Significant lowering of water level above, or damage to, irradiated fuel EAL:
RA2.2          Alert Damage to irradiated fuel resulting in a release of radioactivity AND HIGH alarm on any of the following radiation monitors:
* RIA-3 RB Refueling Deck Shield Wall
* RIA-6 Spent Fuel Building Wall
* RIA-41 Spend Fuel Pool Gas
* RIA-49 RB Gas
* Portable area monitors on the main bridge or SEP bridge Mode Applicability:
All Definition(s):
None Basis:
The specified radiation monitors are those expected to see increase area radiation levels as a result of damage to irradiated fuel. Radiation levels in the Reactor Building refueling area are monitored by RIA-3. Radiation levels in the Spent Fuel Pool area are monitored RIA-6. When a fuel bridge is being used to handle fuel, radiation levels are monitored by a portable area monitor mounted on the bridge. (ref. 1, 2, 3)
The HIGH alarm for RIA-3 (containment area monitor) and RIA-49 (RB gaseous process monitor) corresponds to the setpoints established to assure that 10 CFR 20 limits are not exceeded.
The HIGH alarm setpoint for RIA-6 (SFP bridge area monitor) is designed to make operators aware of increased readings above 10 CFR 20 limits. The HIGH alarm setpoint for RIA-41 (Spent Fuel Pool gaseous atmosphere) is set to alarm if 4 times the limits of 10 CFR 20 are exceeded based upon Xe-I133. RIA-49 monitors the reactor building gas. Portable monitors are established during refueling outages and are located on the main bridge, and the spent fuel pool bridge.
((Document No.].                               Rev. 0                             Page 55 of 240[

ATTACHMENT 1 EAL Bases This IC addresses events that have caused imminent or actual damage to an irradiated fuel assembly, or a significant lowering of water level within the spent fuel pool. These events present radiological safety challenges to plant personnel and are precursors to a release of radioactivity to the environment. As such, they represent an actual or potential substantial degradation of the level of safety of the plant.
This EAL applies to irradiated fuel that is licensed for dry storage up to the point that the loaded storage cask is sealed. Once sealed, damage to a loaded cask causing loss of the CONFINEMENT BOUNDARY is classified in accordance with EAL EU1 .1.
Escalation of th* emergency would be based on either Recognition d~ategory R or C ICs.
This EAL addresses a release of radioactive material caused by mechanical damage to irradiated fuel. Damaging events may include the dropping, bumping or binding of an assembly, or dropping a heavy load onto an assembly. A rise in readings on radiation monitors should be considered in conjunction with in-plant reports or observations of a potential fuel damaging event (e.g., a fuel handling accident).
Escalation of the emergency classification level would be via IC RSI.
ONS Basis Reference(s):
: 1. OP/1/AI6101/O08, Alarm Response Guide 1SA-08 B-9; OP/2/AI6101/O08; OP/3/A/6101/008
: 2. AP/1 ,2,3/A/1 700/018, Abnormal Release of Radioactivity
: 3. OP/1,2,3/A/1502/007, Enclosure 1, Defueling/Refueling Prerequisites
: 4. NEI 99-01 AA2 S[Document No.]                                 Rev. 0                             page 56 of 240

ATTACHMENT 1 EAL Bases Category:                 R - Abnormal Rad Levels / Rad Effluent Subcategory:              2 - Irradiated Fuel Event Initiating Condition:     Significant lowering of water level above, or damage to, irradiated fuel EAL:
RA2.3          Alert Lowering of spent fuel Pool level to -13.5 ft.
Mode Applicability:
All Definition(s):I None Basis:
This EAL is applicable once the post-Fukushima level instrumentation becomes operational on its associated unit.
Post-Fukushima order EA-12-051 required the installation of reliable SFP level indication capable of identifying normal level (Level 1), SFP level 10 ft. above the top of the fuel racks (Level 2) and SEP level at the top of the fuel racks (Level 3).
SFP level instruments 1/2/3SFP0010 (primary) and 011 (backup) measure SFP level relative to normal water level (El. 840 ft.) from  + 1 ft. to -23.5 ft. (El. 816.4 ft).
For ONS Level 2 corresponds to an indicated water level of -13.5 ft. (El. 826.5 ft.) (ref. 1).
This IC addresses events that have caused imminent or actual damage to an irradiated fuel assembly, or a significant lowering of water level within the spent fuel pool. These events present radiological safety challenges to plant personnel and are precursors to a release of radioactivity to the environment. As such, they represent an actual or potential substantial degradation of the level of safety of the plant.
Escalation of the emergency would be based .on either Recognition Category R or C ICs.
Spent fuel pool water level at this value is within the lower end of the level range necessary to prevent significant dose consequences from direct gamma radiation to personnel performing operations in the vicinity of the spent fuel pool. This condition reflects a significant loss of spent fuel pool water inventory and thus it is also a precursor to a loss of the ability to adequately cool the irradiated fuel assembles stored in the pool.
Escalation of the emergency classification level would be via IC RSI.
ONS Basis Reference(s):
: 1. Engineering Change EC 105805 & 105806
: 2. NEI 99-01 AA2
[Document No.]                         I         Rev. 0            IPage                  57of 240

ATTACHMENT 1 EAL Bases Category:                 R - Abnormal Rad Levels / Rad Effluent Subcategory:              2 - Irradiated Fuel Event Initiating Condition:     Spent fuel pool level at the top of the fuel racks EAL:
RS2.1          Site Area Emergency Lowering of spent fuel pool level to -23.5 ft.
Mode Applicability:
All Definition(s):
None Basis:
This EAL is applicable once the post-Fukushima level instrumentation becomes operational on its associated unit.
Post-Fukushima order EA-1 2-051 required the installation of reliable SFP level indication capable of identifying normal level (Level 1), SEP level 10 ft. above the top of the fuel racks (Level 2) and SFP level at the top of the fuel racks (Level 3).
SEP level instruments 1/2/3SFP0010 (primary) and 011 (backup) measure SEP level relative to normal water level (El. 840 ft.) from + 1 ft. to -23.5 ft. (El. 816.4 ft).
For ONS Level 3 corresponds to an indicated water level of -23.5 ft. (El. 816.5 ft.) (ref. 1).
This EAL addresses a significant loss of spent fuel pool inventory control and makeup capability leading to IMMINENT fuel damage. This condition entails major failures of plant functions needed for protection of the public and thus warrant a Site Area Emergency declaration.
It is recognized that this IC would likely not be met until well after another Site Area Emergency IC was met; however, it is included to provide classification diversity.
Escalation of the emergency classification level would be via IC AGI or RG2.
ONS Basis Reference(s):
: 1. Engineering Change EC 105805 & 105806
: 2. NEI 99-01 AS2 i[Document No.]                                  Rev. 0                              Page 58 of 240'

ATTACHMENT 1 EAL Bases Category:                  R - Abnormal Rad Levels/IRad Effluent Subcategory:               2 - Irradiated Fuel Event Initiating Condition:      Spent fuel pool level cannot be restored to at least the top of the fuel racks for 60 minutes or longer EAL:
SRG2.1          General Emergency Spent fuel pool level cannot be restored t*o at least -23.5 ft. for > 60 min. (Note 1)I Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
Mode Applicability:
All Definition(s):
None Basis:
This EAL is applicable once the post-Fukushima level instrumentation becomes operational on its associated unit.
Post-Fukushima order EA-12-051 required the installation of reliable SFP level indication capable of identifying normal level (Level 1), SEP level 10 ft. above the top of the fuel racks (Level 2) and SEP level at the top of the fuel racks (Level 3).
SEP level instruments 1/2/3SFP0010 (primary) and 011 (backup) measure SFP level relative to normal water level (El. 840 ft.) from + 1 ft. to -23.5 ft. (El. 816.4 ft).
For ONS Level 3 corresponds to an indicated water level of -23.5 ft. (El. 816.5 ft.) (ref. 1).
This EAL addresses a significant loss of spent fuel pool inventory control and makeup capability leading to a prolonged uncovery of spent fuel. This condition will lead to fuel damage and a radiological release to the environment.
It is recognized that this IC would likely not be met until well after another General Emergency IC was met; however, it is included to provide classification diversity.
ONS Basis Reference(s):
: 1. Engineering Change EC 105805 & 105806
: 2. NEI 99-01 AG2
- [Document No.]                                      Rev. 0                              Page 59 of 240

ATTACHMENT 1 EAL Bases Category:                  R - Abnormal Rad Levels / Rad Effluent S ubcategory:             3 - Area Radiation Levels Initiating Condition:      Radiation levels that IMPEDE access to equipment necessary for normal plant operations, cooldown or shutdown EAL:
RA3.1           Alert Dose rates > 15 mR/hr in EITHER of the following areas:
* Control Room (RIA-1)
* Central Alarm Station (by survey)
Mode Applicability:
All Definition(s):
IMPEDE(D) - Personnel access to a room or area is hindered to an extent that extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., requiri~ng use of protective equipment, such as SCBAs, that is not routinely employed).
Basis:
Areas that meet this threshold include the Control Room (CR) and the Central Alarm Station (CAS). RIA-1 monitors the Control room for area radiation (ref. 1). The CAS is included in this EAL because of its importance to permitting access to areas required to assure safe plant operations.
There are no permanently installed area radiation monitors in the CAS that may be used to assess this EAL threshold. Therefore, this threshold is evaluated using local radiation survey for this area (ref. 1).
This IC addresses elevated radiation levels in certain plant rooms/areas sufficient to preclude or impede personnel from performing actions necessary to maintain normal plant operation, or to perform a normal plant cooldown and shutdown. As such, it represents an actual or potential substantial degradation of the level of safety of the plant. The Emergency Coordinator should consider the cause of the increased radiation levels and determine if another IC may be applicable.
Escalation of the emergency classification level would be via Recognition Category R, C or F ICs.
ONS Basis Reference(s):
: 1. UFSAR Table 12-3 Area Radiation Monitors
: 2. NEI 99-01 AA3
* [Document No.]                                  Rev. 0.                         Page 60 of 240I

ATTACHMENT 1 EAL Bases Category:                      R - Abnormal Rad Levels / Rad Effluent Subcategory:                   3 - Area Radiation Levels Initiating Condition:          Radiation levels that IMPEDE access to equipment necessary for normal plant operations, cooldown or shutdown EAL:
RA3.2            Alert An UNPLANNED event results in radiation levels tilat prohibit or IMPEDE access to any Table R-2 rooms or areas (Note 5)
Note 5:   Ifthe equipment in the listed room or area was already inoperable or out-of-service before the event occurred, then no emergency classification is warranted.I Ii         Table R-2           Safe Operation &Shutdown Rooms/Areas                                      1 Room/Area                                        Mode Applicability Turbine Building                                                                        1,2, 3 Equipment and Cable Rooms                                                               1,2, 3 Auxiliary Building                                                                   1, 2, 3,4, 5 Reactor Buildings                                                                       3, 4, 5 Mode Applicability:
All Definition(s):
IMPEDE(D) - Personnel access to a room or area is hindered to an extent that extraordinary*
measures are necessary to facilitate entry of personnel into the affected room/area (e.g., requiring use of protective equipment, such as SCBAs, that is not routinely employed).
UNPLANNED - A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.
Basis:
If the equipment in the listed room or area was already inoperable, or out-of-service, before the event occurred, then no emergency should be declared since the event will have no adverse impact beyond that already allowed by Technical Specifications at the time of the event.
The list of plant rooms or areas with entry-related mode applicability identified specify those rooms or areas that contain equipment which require a manual/local action as specified in operating procedures used for normal plant operation, cooldown and shutdown. Rooms or areas in which actions of a contingent or emergency nature would be performed (e.g., an action to address an off-normal or emergency condition such as emergency repairs, corrective measures or emergency operations) are not included. In addition, the list specifies the plant mode(s) during which entry would be required for each room or area (ref. 1).
I[Document No.]                                           Rev. 0                                        Page 61 of 240

ATTACHMENT 1 EAL Bases This IC addresses elevated radiation levels in certain plant rooms/areas sufficient to preclude or impede personnel from performing actions necessary to maintain normal plant operation, or to perform a normal plant cooldown and shutdown. As such, it represents an actual or potential substantial degradation of the level of safety of the plant. The Emergency Coordinator should consider the cause of the increased radiation levels and determine if another IC may be applicable.
For RA3.2, an Alert declaration is warranted if entry into the affected room/area is, or may be, procedlurally required during the plant operating mode in effect at the time of the elevated radiation levels. The emergency classification is not contingent upon whether entry is actually necessary at the time of the increased radiation levels. Access should be considered as impeded if extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., installing temporary shielding, requiring use of non-routine protective equipment, requesting an extension in dose limits beyond normal administrative limits).
An emergency declaration is not warranted if any of the following conditions apply:
* The plant is in an operating mode different than the mode specified for the affected room/area (i.e., entry is not required during the operating mode in effect at the time of the elevated radiation levels). For example, the plant is in Mode 1 when the radiation increase occurs, and the procedures used for normal operation, cooldown and shutdown do not require entry into the affected room until Mode 4.
* The increased radiation levels are a result of a planned activity that includes compensatory measures which address the temporary inaccessibility of a room or area (e.g., radiography, spent filter or resin transfer, etc.).
* The action for which room/area entry is required is of an administrative or record keeping nature (e.g., normal rounds or routine inspections).
* The access control measures are of a conservative or precautionary nature, and would not actually prevent or impede a required action.
* Ifthe equipment in the listed room or area was already inoperable, or out-of-service, before the event occurred, then no emergency should be declared since the event will have no adverse impact beyond that already allowed by Technical Specifications at the time of the event.
Escalation of the emergency classification level would be via Recognition Category R, C or F ICs.
ONS Basis Reference(s):
: 1. Attachment 3 Safe Operation & Shutdown Rooms/Areas Tables R-3 & H-2 Bases
: 2. NEI 99-01 AA3 l[Document No.]                                   Rev. 0                           Page 62 of 240

ATTACHMENT 1 EAL Bases Categlory E - Independent Spent Fuel Storage Installation (IS FSI)
EAL Group: ANY (EALs in this category are applicable to any plant condition, hot or cold.)
An independent spent fuel storage installation (ISESI) is a complex that is designed and constructed for the interim storage of spent nuclear fuel and other radioactive materials associated with spent fuel storage. A significant amount of the radioactive material contained within a canister must escape its packaging and enter the biosphere for there to be a significant environmental effect resulting from an accident involving tthe dry storage of spent nuclear fuel.
A Notification of Unusual Event is declared on the basis of the occurrence of an event of sufficient magnitude that a loaded cask confinement boundary is damaged or violated.
The ONS ISFSI is contained wholly within the plant Protected Area. Therefore a security event related to the ISESI would be applicable to EALs HU1 .1, HA1 .land HSI1.1 I[Document No.]                                 Rev. 0                             Page 63 of 240

ATTACHMENT 1 EAL Bases Category:                 ISFSI S ubcategory:             Confinement Boundary Initiating Condition:     Damage to a loaded cask CONFINEMENT BOUNDARY EAL:
EUI,1          Unusual Event Damage to a loaded canister CONFINEMENT BOUNDARY as indicated by an on-contact radiation reading on the surface of a loaded spent fuel cask > any Table E-1* ISESI dose limit S    Table E-1   ISFSI Dose Limits Location                 24PHB             37PTH             69BTH HSM front bird screen       1,050 mremfhr     1,050 mrem/hr      500 mrem/hr Outside HSM door              40 mrem/hr        4 mrem/hr         4 mrem/hr End shield wall exterior     550 mrem/hr        8 mrem/hr         8 mrem/hr Mode Applicability:
All Definition(s):
CONFINEMENT BOUNDARY - The barrier(s) between spent fuel and the environment once the spent fuel is processed for dry storage. As related to the ONS ISFSI, Confinement Boundary is comprised of the DSC (dry shielded canister) shell, inner bottom cover plate, inner top cover plate, siphon & vent block, siphon & vent port cover plate, and the welds that join them together.
INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI) - A complex that is designed and constructed for the interim storage of spent nuclear fuel and other radioactive materials associated with spent fuel storage.
Basis:
The ONS ISFSI utilizes the NUHOMS System dry spent fuel storage system for dry spent fuel storage.
The Standardized NUHOMS System is a horizontal canister system composed of a steel dry shielded canister (DSC), a reinforced concrete horizontal storage module (HSM), and a transfer cask (TC). The welded DSC provides confinement and criticality control for the storage and transfer of irradiated fuel. The concrete module provides radiation shielding while allowing cooling of the DSC and fuel by natural convection during storage (ref. 1, 2). The ONS ISESI utilizes the 24PHB, 37PTH and 69BTH DSC designs.
I[Document No.]                                Rev. 0         IPage                   64 of 240

ATTACHMENT 1 EAL Bases Confinement boundary is defined as the barrier(s) between areas containing radioactive substances and the environment. Therefore, damage to a confinement boundary must be a confirmed physical breach between the spent fuel and the environment for the TSC.
The Table E-1 values shown are 2 times the limits specified in the ISESI Certificate of Compliance Technical Specifications for radiation external to the applicable loaded DSC (ref.
1,2).
This IC addresses an event that results in damage to the CONFINEMENT BOUNDARY of a storage cask containing spent fuel. It applies to irradiated fuel that is licensed for dry storage beginning at the point that the 15aded storage cask is sealed. The issues of concernP are the creation of a potential or actual release path to the environment, degradation of one or more fuel assemblies due to environmental factors, and configuration changes which could cause challenges in removing the cask or fuel from storage.
The existence of "damage" is determined by radiological survey. The technical specification multiple of "2 times", which is also used in Recognition Category R IC RU1, is used here to distinguish between non-emergency and emergency conditions. The emphasis for this classification is the degradation in the level of safety of the spent fuel cask and not the magnitude of the associated dose or dose rate. It is recognized that in the case of extreme damage to a loaded cask, the fact that the "on-contact" dose rate limit is exceeded may be determined based on measurement of a dose rate at some distance from the cask.
Security-related events for ISFSIs are covered under l~s HU1 and HA1.
ONS Basis Reference(s):
: 1. USNRC Certificate of Compliance for Spent Fuel Storage Casks, No. 1004, Amendment 13, Attachment A, Technical Specifications for Transnuclear, Inc., Standardized NUHOMS Horizontal Modular Storage System
: 2. OSC-8716, Oconee ISFSI Dose Rate Evaluations, Rev. 0 (4/29/05)
: 3. NEI 99-01 E-HU1
{[Document No.]                        1Rev.          0                             Page 65 of 240.1

ATTACHMENT 1 EAL Bases Category C - Cold Shutdown / Refueling System Malfunction EAL Group: Cold Conditions (RCS temperature -<200&deg;F); EALs in this category are applicable only in one or more cold operating modes.
Category C EALs are directly associated with cold shutdown or refueling system safety functions. Given the variability of plant configurations (e.g., systems out-of-service for maintenance, containment open, reduced AC power redundancy, time since shutdown) during these periods, the consequences of any given initiating event can vary greatly. For example, a loss of decay heat removal capability thMt occurs at the end of an extended outage has less significance than a similar loss occurring during the first week after shutdown. Compounding these events is the likelihood that instrumentation necessary for assessment may also be inoperable. The cold shutdown and refueling system malfunction EALs are based on performance capability to the extent possible with consideration given to RCS integrity, containment closure, and fuel clad integrity for the applicable operating modes (5 - Cold Shutdown, 6 - Refueling, NM - No Mode).
The events of this category pertain to the following subcategories:
: 1. RCS Level RCS water level is directly related to the status of adequate core cooling and, therefore, fuel clad integrity.
: 2. Loss of Essential AC Power Loss of essential plant electrical power can compromise plant safety system operability including decay heat removal and emergency core cooling systems which may be necessary to ensure fission product barrier integrity. This category includes loss of onsite and offsite power sources for 41 60V AC essential buses.
: 3. RCS Temperature Uncontrolled or inadvertent temperature or pressure increases are indicative of a potential loss of safety functions.
: 4. Loss of Vital DC Power Loss of emergency plant electrical power can compromise plant safety system operability including decay heat removal and emergency core cooling systems which may be necessary to ensure fission product barrier integrity. This category includes loss of power to or degraded voltage on the 125V DC vital buses.
: 5. Loss of Communications Certain events that degrade plant operator ability to effectively communicate with essential personnel within or external to the plant warrant emergency classification.
: 6. Hazardous Event Affectincq Safety Systems Certain hazardous natural and technological events may result in visible damage to or degraded performance of safety systems warranting classification.
S[Document No.]                                  Rev. 0                            Page 66 of 240I

ATTACHMENT 1 EAL Bases Category:                  C - Cold Shutdown I Refueling System Malfunction Subcategory:               1 - RCS Level Initiating Condition:      UNPLANNED loss of RCS inventory for 15 minutes or longer EAL:
Note 1:   The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
Mode Applicability:
5 - Cold Shutdown, 6 - Refueling Definition(s):
UNPLANNED - A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.
Basis:
RCS water level less than a required lower limit is meant to be less than the lower end of the level control band being procedurally maintained for the current condition or evolution.
RCS water level instrumentation requirements to begin an RCS inventory reduction with fuel in the core to below 80" (lowered inventory) or 50" (reduced inventory) are the following (ref. 1):
* Both channels of LT-5 prior to reducing ROS inventory below 80".
* Both channels of LT-5 and both hot leg and cold leg ultrasonic monitors prior to reducing RCS inventory below 50".
This IC addresses the inability to restore and maintain water level to a required minimum level (or the lower limit of a level band), or a loss of the ability to monitor RCS level concurrent with indications of coolant leakage. Either of these conditions is considered to be a potential degradation of the level of safety of the plant.
Refueling evolutions that decrease RCS water inventory are carefully planned and controlled.
An UNPLANNED event that results in water level decreasing below a procedurally required limit warrants the declaration of an Unusual Event due to the reduced water inventory that is available to keep the core covered.
This EAL recognizes that the minimum required RCS level can change several times during the course of a refueling outage as different plant configurations and system lineups are implemented. This EAL is met if the minimum level, specified for the current plant conditions, cannot be maintained for 15 minutes or longer. The minimum level is typically specified in the applicable operating procedure but may be specified in another controlling document.
S[Document No.]                                       Rev. 0                             Page 67 of 240

ATTACHMENT 1 EAL Bases The 15-minute threshold duration allows sufficient time for prompt operator actions to restore and maintain the expected water level. This criterion excludes transient conditions causing a brief lowering of water level.
Continued loss of RCS inventory may result in escalation to the Alert emergency classification level via either IC CA1 or CA3.
ONS Basis Reference(s):
: 1. S. D. 1.3.5 Shutdown Protection Plan, Section 5.2.7
: 2. NEI 99-01 CU1 S[Document No.]                             Rev. 0                             Page 68 of 240I

ATTACHMENT I EAL Bases Category:                 C - Cold Shutdown / Refueling System Malfunction Subcategory:              1 - RCS Level Initiating Condition:     UNPLANNED loss of RCS inventory for 15 minutes or longer EAL:
CU1.2          Unusual Event RC3S level cannot be monitored AND EITHER
* UNPLANNED increase in any Table C3-I sump/tank level due to loss of RC3S inventory
* Visual observation of UNISOLABLE RCS leakage Table C-I     Sumps / Tanks
* RB Normal Sumps
* RB Emergency Sumps
* Core Flood Tank
* Quench Tank
* Low Activity Waste Tank
* High Activity Waste Tank
* Miscellaneous Waste Holdup Tank
* LPI Room Sumps Mode Applicability:
5 - Cold Shutdown, 6 - Refueling Definition(s):
UNISOLABLE - An open or breached system line that cannot be isolated, remotely or locally.
UNPLANNED - A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.
Basis:
A RC3S leak should be considered UNISOLABLE if the leak cannot be isolated within 15 min.
In Cold Shutdown mode, the RC3S will normally be intact and standard RC3S level monitoring means are available.
In the Refuel mode, the RC3S is not intact and reactor vessel level may be monitored by different means, including the ability to monitor level visually.
In this EAL, all water level indication is unavailable and the RC3S inventory loss must be detected by indirect leakage indications (Table C3-I). Level increases must be evaluated S[Document No.]                                 Rev. 0                           Page 69 of 240

ATTACHMENT 1 EAL Bases against other potential sources of leakage such as cooling water sources inside the containment to ensure they are indicative of RCS leakage. If the make-up rate to the RCS unexplainably rises above the pre-established rate, a loss of RCS inventory may be occurring even if the source of the leakage cannot be immediately identified. Visual observation of significant leakage from systems connected to the RCS that cannot be isolated could also be indicative of a loss of RCS inventory (ref. 1, 2).
This IC addresses the inability to restore and maintain water level to a required minimum level (or the lower limit of a level band), or a loss of the ability to monitor RCS level concurrent with indications of' coolant leakage. Either of these conditions is considered to be a potential degradation of the level of safety of the plant.
Refueling evolutions that decrease RCS water inventory are carefully planned and controlled.
' An UNPLANNED event that results in water level decreasing below a procedurally required limit warrants the declaration of an Unusual Event due to the redu~ced water inventory that is available to keep the core covered.
This EAL addresses a condition where all means to determine level have been lost. In this condition, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels (Table C-I). Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the RCS.
Continued loss of RCS inventory may result in escalation to the Alert emergency classification level via either IC CAl or CA3.
ONS Basis Reference(s):
: 1. AP/1 ,2,3/A/1 700/002 Excessive RCS Leakage
: 2. AP/1 -2,3/NI1700/030 Auxiliary Building Flood
: 3. NEI 99-01 CU1                7 l[Document No.]                                  Rev. 0                              Page 70 of 240I

ATTACHMENT 1 EAL Bases Category:                  C - Cold Shutdown / Refueling System Malfunction Subcategory:               1 - RCS Level Initiating Condition:      Loss of RCS inventory EAL:
CA1.1          Alert Loss of RCS inventory as indicated by RCS level       < 10" (LT-5)
Mode Applicability:
5 - Cold Shutdown, 6 - Refueling Definition(s):           I None Basis:
RCS water level of 10" as indicated on LT-5 is the lowest level for continued operation of LPI pumps for decay heat removal (ref. 1). Two LPI pumps and two coolers normally perform the' decay heat removal function for each unit (ref. 2).
The threshold was chosen because a loss of suction to decay heat removal systems may occur. The inability to restore and maintain level after reaching this setpoint infers a failure of the RCS Barrier.
This IC addresses conditions that are precursors to a loss of the ability to adequately cool irradiated fuel (i.e., a precursor to a challenge to the fuel clad barrier). This condition represents a potential substantial reduction in the level of plant safety.
For this EAL, a lowering of RCS water level below 10 in. indicates that operator actions have not been successful in restoring and maintaining RCS water level. The heat-up rate of the coolant will increase as the available water inventory is reduced. A continuing decrease in water level will lead to core uncovery.
Although related, this EAL is concerned with the loss of RCS inventory and not the potential concurrent effects on systems needed for decay heat removal (e.g., loss of a Decay Heat Removal suction point). An increase in RCS temperature caused by a loss of decay heat removal capability is evaluated under IC CA3.
If RCS water level continues to lower, then escalation to Site Area Emergency would be via IC CSI.
ONS Basis Reference(s):
: 1. AP/1 ,2,3/A/1700/026 Loss of Decay Heat Removal
: 2. UFSAR Section 9.3.3 Low Pressure Injection System
: 3. NEI 99-01 CA1 i[Document No.]                                  Rev. 0                              Page. 71 of 240I

ATTACHMENT 1 EAL Bases Category:                     C - Cold Shutdown / Refueling System Malfunction Subcategory:                  1 - RCS Level Initiating Condition:         Loss of RCS inventory EAL:
CA1.2           Alert RCS level cannot be monitored for >- 15 mai.         (Note 1)
AND EITHER
* UNPLANNED increase in any Table C-i Sump / Tank level due to a loss of RCS inventory
* Visual observation of UNISOLABLE ROS leakage exceeded, or will likely be exceeded.
Table C-I      Sumps I Tanks
* RB Normal Sumps
* RB Emergency Sumps
* Core Flood Tank
* Quench Tank
* Low Activity Waste Tank
* High Activity Waste Tank
* Miscellaneous Waste Holdup Tank
* LPI Room Sumps Mode Applicability:
5 - Cold Shutdown, 6 - Refueling Definition(s):
UNISOLABLE - An open or breached system line that cannot be isolated, remotely or locally.
UNPLANNED - A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.
Basis:
A RCS leak should be considered UNISOLABLE if the leak cannot be isolated within 15 minutes.
In Cold Shutdown mode, the RCS will normally be intact and standard RCS level monitoring means are available.
I[Document No.]                            I         Rev. 0                     Page 72o&deg;240

ATTACHMENT 1 EAL Bases In the Refuel mode, the RCS is not intact and RCS level may be monitored by different means, including the ability to monitor level visually.
In this EAL, all RCS water level indication would be unavailable for greater than 15 minutes, and the RCS inventory loss must be detected by indirect leakage indications (Table C-1). Level increases must be evaluated against other potential sources of leakage such as cooling water sources inside the containment to ensure they are indicative of RCS leakage. If the make-up rate to the RCS unexplainably rises above the pre-established rate, a loss of RCS inventory may be occurring even if the source of the leakage cannot be immediately identified. Visual observation of significant leakage from systems connected to the RCS that cannot be isolated could also be indicative of a loss of RCS inventory (ref. 1, 2).
This IC addresses conditions that are precursors to a loss of the ability to adequately cool irradiated fuel (i.e., a precursor to a ch~allenge to the fuel clad barrier). This condition represents a potential substantial reduction in the level of plant safety.
For this EAL, the inability to monitor RCS level may be caused by instrumentation and/or power failures, or water level dropping below the range of available instrumentation. Ifwater level cannot be monitored, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the RCS.
The 15-minute duration for the loss of level indication was chosen because it is half of the EAL duration specified in IC CSI.
If the RCS inventory level continues to lower, then escalation to Site Area Emergency would be via IC CSI.
ONS Basis Reference(s):
: 1. AP/1 ,2,3/A/1700/002 Excessive RCS Leakage
: 2. AP/1 -2,3/NI1700/030 Auxiliary Building Flood
: 3. NEI 99-01 CA1
((Document No.]                                   Rev. 0                              Page 73 of 240I

ATTACHMENT 1 EAL Bases Category:                       C - Cold Shutdown / Refueling System Malfunction Subcategory:                    1 - RCS Level Initiating Condition:           Loss of RCS inventory affecting core decay heat removal capability EAL:
CSl.1           Site Area Emergency RCS level cannot be monitored for > 30 min. (Note I)
AND Core uncovery is indicated by any of the following:
* UNPLANNED increase in any Table C-I sump/tank level
* Visual observation of UNISOLABLE RCS leakage
* High alarm on RIA-3 RB Refueling Deck Shield Wall
* Erratic Source Range Monitor Indication Note 1:   The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
Table C-I     Sumps / Tanks
* RB Normal Sumps
* RB Emergency Sumps
* Core Flood Tank
* Quench Tank
* Low Activity Waste Tank
* High Activity Waste Tank
* Miscellaneous Waste Holdup Tank
* LPI Room Sumps Mode Applicability:
5 -  Cold Shutdown, 6 - Refueling Definition(s):
UNISOLABLE - An open or breached system line that cannot be isolated, remotely or locally.
UNPLANNED - A parameter change or an event that is not I) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.
Basis:
A RCS leak should be considered UNISOLABLE if the leak cannot be isolated within 15 minutes.
[Dcmn o]Rev.                                        0                                     Page 74 of 240

ATTACHMENT 1 EAL Bases In Cold Shutdown mode, the RCS will normally be intact and standard RCS level monitoring means are available.
In the Refueling mode, the RCS is not intact and RCS level may be monitored by different means, including the ability to monitor level visually.
In this EAL, all RCS water level indication would be unavailable for greater than 30 minutes, and the RCS inventory loss must be detected by indirect leakage indications (Table C-I). Level increases must be evaluated against other potential sources of leakage such as cooling water sources inside the containment to ensure they are indicative of RCS leakage. If the make-up rafe to the RCS unexplainably rises above the pre-established rate, a loss of RCS inventory may be occurring even if the source of the leakage cannot be immediately identified. Visual observation of significant leakage from systems connected to the RCS that cannot be isolated could*also be indicative of a loss of RCS inventory (ref.! 1, 2).
Sump or tank level increases should be of a magnitude that correlates to a volume sufficient to indicate fuel has been uncovered or uncovery is imminent.
The Reactor Vessel inventory loss may be detected by a reduction in water shielding that causes a high alarm on the Refueling Deck Shield Wall area radiation monitor (ref. 3).
Post-TM I accident studies indicated that the installed PWR nuclear instrumentation will operate erratically when the core is uncovered and that this should be used as a tool for making such determinations (ref. 4, 5, 6).
This IC addresses a significant and prolonged loss of reactor vessel/ROS inventory control and makeup capability leading to IMMINENT fuel damage. The lost inventory may be due to a RCS component failure, a loss of configuration control or prolonged boiling of reactor coolant.
These conditions entail major failures of plant functions needed for protection of the public and thus warrant a Site Area Emergency declaration.
Following an extended loss of core decay heat removal and inventory makeup, decay heat will cause reactor coolant boiling and a further reduction in reactor vessel level. If RCS level cannot be restored, fuel damage is probable.
The 30-minute criterion is tied to a readily recognizable event start time (i.e., the total loss of ability to monitor level), and allows sufficient time to monitor, assess and correlate reactor and plant conditions to determine if core uncovery has actually occurred (i.e., to account for various accident progression and instrumentation uncertainties). It also allows sufficient time for performance of actions to terminate leakage, recover inventory control/makeup equipment and/or restore level monitoring.
The inability to monitor RCS level may be caused by instrumentation and/or power failures, or water level dropping below the range of available instrumentation. Ifwater level cannot be monitored, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the RCS.
This EAL addresses concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal; SECY 91-283, Evaluation of Shutdown and Low Power Risk Issues; NUREG-1 449, 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.
S[Document.No.]                                 Rev. 0                             Page 75 of 240

ATTACHMENT 1 EAL Bases Escalation of the emergency classification level would be via IC OG1 or RG1.
ONS Basis Reference(s):
: 1. API/1,2,3/A/1700/002 Excessive RCS Leakage
: 2. AP/1-2,3/A/1700/030 Auxiliary Building Flood
: 3. UFSAR Table 12-3 Area Radiation Monitors
: 4. UFSAR Section 7.4.1 Nuclear Instrumentation
: 5. OP/1,2,3/A/5102/002 Alarm Response Guide 1,2,3SA-02, A-6
: 6. Nuclear Safety Analysis Center (NSAC), 1980, "Analysis of Three Mile Island  - Unit 2 Accident," NSAC-1
: 7. NEI 99-01 CS1 S[Document No.]                               Rev. 0                          Page 76 of 240

ATTACHMENT 1 EAL Bases Category:                   C - Cold Shutdown / Refueling System Malfunction Subcategory:                1 - RCS Level Initiating Condition:       Loss of RCS inventory affecting fuel clad integrity with containment challenged EAL:
CGI.1           General Emergency RCS level cannot be monitored for ->30 min. (Note 1)
AND Core uncovery is indicated by any of the following:
* UNPLANNED increase in any Table C-i sump/tank level due to a loss of RCS inventory
* Visual observation of UNISOLABLE RCS leakage
* High alarm on RIA-3 RB Refueling Deck Shield Wall
* Erratic Source Range Monitor Indication AND Any Containment Challenge indication, Table C-2 Note 1:   The been Emergency exceeded, Coordinator    should declare the event promptly upon determining that time limit has or will likely be exceeded.
Note 6:   If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, declaration of a General Emergency is not required.
Table C-I       Sumps I Tanks
* RB Normal Sumps
* RB Emergency Sum ps
* Core Flood Tank
* Quench Tank
* Low Activity Waste Tank
* High Activity Waste Tank
* Miscellaneous Waste Holdup Tank
* LPI Room Sumps I[Document No.]                                        Rev. 0                                 Page 77 of 240

ATTACHMENT 1 EAL Bases TablC-2   Containment Challenge TablC-2    Indications
* CONTAINMENT CLOSURE not established (Note 6)
* Containment hydrogen concentration
                                  > 4%
                                *,Un planned rise in containment pressure Mode Applicability:
5 - Cold Shutdown, 6 - Refueling Definition(s):
CONTAINMENT CLOSURE - The action to secure containment and its associated structures, systems, and components as a functional barrier to fission product release under all plant conditions up to and including a loss of decay heat removal or fuel handling accident inside containment.
As applied to ONS, Containment Closure is established when the requirements of OP/i1,2,3/A/1502/009, Containment Closure Control, are met.
UNISOLABLE - An open or breached system line that cannot be isolated, remotely or locally.
UNPLANNED - A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the paraimeter change or event may be known or unknown.
Basis:
A RCS leak should be considered UNISOLABLE if the leak cannot be isolated within 15 minutes.
In Cold Shutdown mode, the RCS will normally be intact and standard RCS level monitoring means are available.
In the Refueling mode, the RCS is not intact and RCS level may be monitored by different means, including the ability to monitor level visually.
In this EAL, all RCS water level indication would be unavailable for greater than 30 minutes, and the RCS inventory loss must be detected by indirect leakage indications (Table C-I). Level increases must be evaluated against other potential sources of leakage such as cooling water sources inside the containment to ensure they are indicative of RCS leakage. If the make-up rate to the RCS unexplainably rises above the pre-established rate, a loss of RCS inventory may be occurring even if the source of the leakage cannot be immediately identified. Visual observation of significant leakage from systems connected to the RCS that cannot be isolated could also be indicative of a loss of RCS inventory (ref. 1, 2).
S[Document No.]                                 Rev. 0                           Page 78 of 2401

ATTACHMENT 1 EAL Bases Sump or tank level increases should be of a magnitude that correlates to a volume sufficient to indicate fuel has been uncovered or uncovery is imminent.
The Reactor Vessel inventory loss may be detected by a reduction in water shielding that causes a high alarm on the Refueling Deck Shield Wall area radiation monitor (ref. 3).
Post-TMI accident studies indicated that the installed PWR nuclear instrumentation will operate erratically when the core is uncovered and that this should be used as a tool for making such determinations (ref. 4, 5, 6).
Three conditions are associated witl] a challenge to Containment integrity:
: 1. CONTAINMENT CLOSURE not established - The status of containment closure is tracked if plant conditions change that could raise the risk of a fission product release as a result of a loss of decay heat removal (ref. 7). If containment closure is re-established prior to exceeding the 30 minute core uncovery time limit then escalation to GE would not occur.
: 2. Containment hydrogen > 4% - The 4% hydrogen concentration threshold is generally considered the lower limit for hydrogen combustion. ONS is equipped with a Containment Hydrogen Monitoring System (CHMS) that provides continuous indication of hydrogen concentration in the containment atmosphere. The measurement capability is provided over the range of 0% to 10%. A continuous indication of the hydrogen concentration is not required in the control room at all times during normal operation. If continuous indication of the hydrogen concentration is not available at all times, continuous indication and recording shall be functioning within 90 minutes of the initiation of the safety injection. (ref. 8, 9)
: 3. UNPLANNED rise in containment pressure - An unplanned pressure rise in containment while in cold shutdown or refueling modes can threaten Containment Closure capability and thus containment potentially cannot be relied upon as a barrier to fission product release.
This IC addresses the inability to restore and maintain reactor vessel level above the top of active fuel with containment challenged. This condition represents actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity.
Releases can be reasonably expected to exceed EPA PAG exposure levels offsite for more than the immediate site area.
Following an extended loss of core decay heat removal and inventory makeup, decay heat will cause reactor coolant boiling and a further reduction in reactor vessel level. If RCS level cannot be restored, fuel damage is probable.
With CONTAINMENT CLOSURE not established, there is a high potential for a direct and unmonitored release of radioactivity to the environment. If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, then declaration of a General Emergency is not required.
The existence of an explosive mixture means, at a minimum, that the containment atmospheric hydrogen concentration is sufficient to support a hydrogen burn (i.e., at the lower deflagration limit). A hydrogen burn will raise containment pressure and could result in collateral equipment damage leading to a loss of containment integrity. It therefore represents a challenge to Containment integrity.
S[Document No.]                                    Rev. 0                          Page 79 of 240

ATTACHMENT 1 EAL Bases In the early stages of a core uncovery event, it is unlikely that hydrogen buildup due to a core uncovery could result in an explosive gas mixture in containment. If all installed hydrogen gas monitors are out-of-service during an event leading to fuel cladding damage, it may not be possible to obtain a containment hydrogen gas concentration reading as ambient conditions within the containment will preclude personnel access. During periods when installed containment hydrogen gas monitors are out-of-service, operators may use the other listed indications to assess whether or not containment is challenged.
The 30-minute criterion is tied to a readily recognizable event start time (i.e., the total loss of ability to monitor level), and allows sufficient time to monitor, assess and correlate reactor1 and plant conditions to determine if core uncovery has actually occurred (i.e., to account for various accident progression and instrumentation uncertainties). It also allows sufficient time for performance of actions to terminate leakage, recover inventory control/makeup equipment and/or restore level monitoring.       I The inability to monitor RCS level may be caused by instrumentation and/or power failures, or water level dropping below the range of available instrumentation. Ifwater level cannot be monitored, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the RCS.
This EAL addresses concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal; SECY 91-283, Evaluation of Shutdown and Low Power Risk Issues; NUREG-1 449, 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.
ONS Basis Reference(s):
: 1. AP/1 ,2,3/A/1 700/002 Excessive RCS Leakage
: 2. AP/1-2,3/A/1700/030 Auxiliary Building Flood
: 3. UFSAR Table 12-3 Area Radiation Monitors
: 4. UFSAR Section 7.4.1 Nuclear Instrumentation
: 5. OP/1/A/6101/002; OP/2/A/6102/002; OP/3/A/6103/002 Alarm Response Guide 1,2,3SA-02, A-6
: 6. Nuclear Safety Analysis Center (NSAC), 1980, "Analysis of Three Mile Island         - Unit 2 Accident," NSAC-1
: 7. OP/i1,2,3/A/1 502/009 Containment Closure Control
: 8. UFSAR Section 9.3.7 Containment Hydrogen Monitoring System
: 9. UFSAR Section 15.16.3 Evaluation of Hydrogen Concentrations
: 10. NEI 99-01 CG1 S[Document No.]                                 Rev. 0                             Page 80 of 240

ATTACHMENT 1 EAL Bases Category:                  C - Cold Shutdown / Refueling System Malfunction Subcategory:               2 - Loss of Essential AC Power Initiating Condition:      Loss of all but one AC power source to essential buses for 15 minutes or longer EAL:
CU2.1           Unusual Event AC power capability, Table C-3, to essential 4160 V buses MFB-1 and MFB-2 reduced to a single power source for > 15 min. (Note 1)
AND Any additional single power source failure wil result in loss of all AC power to SAFETY SYSTEMS Note 1:   The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
Table C-3 AC Power Sources Offsite:
* Unit Normal Transformer (backcharged)
* Unit Startup Transformer (SWYD)
* Another Unit Startup Transformer (aligned) (SWYD)
* CT5 (Central/energizing Standby Bus)
Emergency:
* Unit Startup Transformer (Keowee)
* Another Unit Startup Transformer (aligned) (Keowee)
* CT4
* CT5 (dedicated line/energizing Standby Bus)
Mode Applicability:
5 - Cold Shutdown, 6 - Refueling, NM           - No Mode Definition(s):
SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2):
Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:
(1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition;*
S[Document No.]                                      Rev. 0                             Page 81 of 240

ATTACHMENT 1 EAL Bases (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.
Basis:
For emergency classification purposes, "capability" means that an AC power source is available to and capable of powering the emergency bus(es) within 15 min, whether or not the buses are currently powered from it.
The condition indicated by this EAL is the degradation of the offsite and onsite power sources such that any additional single failure would result in a loss of all AC power to the emergency b~uses. 4160 V buses MFB-1 and MFB-2 are the essential buses (ref. 1).
Each unit is provided with two physically independent circuits from the switching station. One is the circuit from the 230 kV switching station through the startup transformer, which is designed to be available within a few seconds following a loss Of coolant accident. The second circuit is the path from the switchyard through the main step-up transformer, the main generator bus and the unit auxiliary transformer with the generator disconnected from the main bus. The second circuit is currently used during refueling as an additional power feed for the shutdown unit(s) from the 230 kV switchyard. Whenever there is inadequate power from the generating units, the 230 kV switching station and the hydro units, power is available to the standby power buses either directly from the 100 kV Central Tie Substation or from Lee Steam Station via a Transformer CT5 (ref. 2).
Upon loss of power from the Oconee generating unit and 230 kV switchyard, power is supplied from both Keowee Hydro Station units through two separate and independent routes. One route is an underground feeder to Transformer CT4 which supplies the two redundant Main Feeder Buses (MFB-1 and MFB-2). The other route is an overhead feeder to the 230 KV switching station which supplies each unit startup transformer (ref. 3).
The Standby Shutdown Facility (SSF) consists of standby systems for use in an extreme emergency and is equipped with a manually started diesel generator that can supply power necessary to maintain hot shutdown of the reactors of each unit in the event of loss of power.
However, the SSF cannot supply power to the essential buses and therefore not credited in this EAL (ref. 3).
This cold condition EAL is equivalent to the hot condition EAL SAlI.1.
This IC describes a significant degradation of offsite and onsite AC power sources such that any additional single failure would result in a loss of all AC power to SAFETY SYSTEMS. In this condition, the sole AC power source may be powering one, or more than one, train of safety-related equipment.
When in the cold shutdown, refueling, or no mode, this condition is not classified as an Alert because of the increased time available to restore another power source to service. Additional time is available due to the reduced core decay heat load, and the lower temperatures and pressures in various plant systems. Thus, when in these modes, this condition is considered to be a potential degradation of the level of safety of the plant.
An "AC power source" is a source recognized in APs and EOPs, and capable of supplying required power to an essential bus. Some examples of this condition are presented below.
* A loss of all offsite power with a concurrent failure of all but one essential power source S[Docume~nt No.]                                 Rev. 0                              Page 82 of 2401

ATTACHMENT 1 EAL Bases (e.g., CT4, CT5, CT1 (Keowee).
      *A loss of essential power sources (e.g., CT4, CT5, CT1, 2, 3 (Keowee)) with a single train of essential buses being back-fed from an offsite power source.
Fifteen minutes was selected as a threshold to exclude transient or momentary losses of power.
The subsequent loss of the remaining single power source would escalate the event to an Alert in accordance with IC CA2.
ONS Bas~is Reference(s):I
: 1. UFSAR Figure 8.1 Single Line Diagram
: 2. UFSAR Section 8.20Offsite Power System
: 3. UFSAR Section 8.30Onsite Power Systems
: 4. NEI 99-01 CU2 S[Document No.]                               Rev. 0                         Page 83 of 240

ATTACHMENT 1 EAL Bases Category:                     C - Cold Shutdown / Refueling System Malfunction Subcategory:                  2 - Loss of Essential AC Power Initiating Condition:         Loss of all offsite and all emergency AC power to essential buses for 15 minutes or longer EAL:
CA2,1           Alert Loss of all offsiie and all emergency AC power capability to essential 4160 V buses MFB-1 and MFB-2 for > 15 min. (Note 1)
Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
Mode Applicability:
5 - Cold Shutdown, 6 - Refueling, NM          -  No Mode Basis:
For emergency classification purposes, "capability" means that an AC power source is available to and capable of powering the emergency bus(es) within 15 mai, whether or not the buses are currently powered from it.
The condition indicated by this EAL is the degradation of the offsite and onsite power sources resulting in a loss of all AC power to the emergency buses. 4160 V buses MFB-1 and MFB-2 are the essential buses (ref. 1).
Each unit is provided with two physically independent circuits from the switching station. One is the circuit from the 230 kV switching station through the startup transformer, which is designed to be available within a few seconds following a loss of coolant accident. The second circuit is the path from the switchyard through the main step-up transformer, the main generator bus and the unit auxiliary transformer with the generator disconnected from the main bus. The second circuit is currently used during refueling as an additional power feed for the shutdown unit(s) from the 230 kV switchyard. Whenever there is inadequate power from the generating units, the 230 kV switching station and the hydro units, power is available to the standby power buses either directly from the 100 kV Central Tie Substation or from Lee Steam Station via a Transformer CT5. (ref. 2)
Upon loss of power from the Oconee generating unit and 230 kV switchyard, power is supplied from both Keowee Hydro Station units through two separate and independent routes. One route is an underground feeder to Transformer CT4 which supplies the two redundant Main Feeder Buses (MFB-1 and MFB-2). The other route is an overhead feeder to the 230 KV switching station which supplies each unit startup transformer. (ref. 3)
The Standby Shutdown Facility (SSF) consists of standby systems for use in an extreme emergency and is equipped with a manually started diesel generator that can supply power necessary to maintain hot shutdown of the reactors of each unit in the event of loss of power.
However, the SSF cannot supply power to the essential buses and therefore not credited in this EAL (ref. 3).
I[Document No.]                                      Rev. 0         IPage                         84 of 240

ATTACHMENT 1 EAL Bases This cold condition EAL is equivalent to the hot condition loss of all offsite AC power EAL SS1.1.
This IC addresses a total loss of AC power that compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink.
When in the cold shutdown, refueling, or no mode, this condition is not classified as a Site Area Emergency because of the increased time available to restore an emergency bus to service. Additional time is available due to the reduced core decay heat load, and the lower temperatures and press*Jres in various plant systems. Thus, when in these mfodes, this condition represents an actual or potential substantial degradation of the level of safety of the plant.
Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.
Escalation of the emergency classification level would be via IC CS1 or RS1.
ONS Basis Reference(s):
: 1. UFSAR Figure 8.1 Single Line Diagram
: 2. UFSAR Section 8.20Offsite Power System
: 3. UFSAR Section 8.30Onsite Power Systems
: 4. NEI 99-01 CA2
[Document No.]                                Rev. 0                              Page 85 of 240

ATTACHMENT 1 EAL Bases Category:                      C - Cold Shutdown / Refueling System Malfunction Subcategory:                   3 - RCS Temperature Initiating Condition:          UNPLANNED increase in RCS temperature EAL:
Note 10: In the absence of reliable RCS temperature indication caused by the loss of decay heat removal capability, classification should be based on time to boil data.
Mode Applicability:
5 - Cold Shutdown, 6 - Refueling Definition(s):
UNPLANNED - A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.
Basis:
Several instruments are capable of providing indication of RCS temperature with respect to the Technical Specification cold shutdown temperature limit (200&deg;F, ref. 1). These include cold leg (Tc) temperature indications, hot leg (Th) temperature indications with RCPs running, CETCs and LPI cooler outlet temperature indications (ref. 2).
However, if Low Pressure Injection (LPI) flow is lost, the normal temperature elements used to monitor RCS temperature are not accurate indicators of RCS temperature. The CETCs are the design instruments for these conditions. For some periods of time the CETCs may not be available. The current practices concerning determining time to boil can be used in the evaluation of these EALs. Without CETC indication and with a loss of LPI flow the following guidance should be used (ref. 2):
* Use the predetermined "time to boil" data for evaluating these EALs. This approach reflects the relatively small numerical difference between the typical Technical Specification cold shutdown temperature limit of 2000 F and the boiling temperature of RCS water with the plant in Mode 5 or 6.
* Alternately, the Control Room staff may use a procedure or user aid to determine when RCS temperature will likely exceed 200&deg;F given the actual plant conditions (e.g., using a heat-up curve).
This IC addresses an UNPLANNED increase in RCS temperature above the Technical Specification cold shutdown temperature limit and represents a potential degradation of the level of safety of the plant. Ifthe RCS is not intact and CONTAINMENT CLOSURE is not established during this event, the Emergency Coordinator should also refer to IC CA3.
A momentary UNPLANNED excursion above the Technical Specification cold shutdown temperature limit when the heat removal function is available does not warrant a classification.
r[Document No.]                                        Rev.O                                Page 86 of 240

ATTACHMENT 1 EAL Bases This EAL involves a loss of decay heat removal capability, or an addition of heat to the RCS in excess of that which can currently be removed, such that reactor coolant temperature cannot be maintained below the cold shutdown temperature limit specified in Technical Specifications.
During this condition, there is no immediate threat of fuel damage because the core decay heat load has been reduced since the cessation of power operation.
During an outage, the level in the reactor vessel will normally be maintained at or above the reactor vessel flange. Refueling evolutions that lower water level below the reactor vessel flange are carefully planned and controlled. A loss of forced decay heat removal at reduced inventory may result in a rapid increase iij reactor coolant temperature depending on the timel after shutdown.
Escalation to Alert would be via IC CA1 based on an inventory loss or IC CA3 based on exceeding plant configuration-specific time criteria.
ONS Basis Reference(s):
: 1. ONS Technical Specifications Table 1.1-1
: 2. AP/l ,2,3/N1 700/026 Loss of Decay Heat Removal
: 3. NEI 99-01 CU3 I [Document No.]                                 Rev. 0                          Page   87 of 240 I
[DcmetNo]Re.0                                                                    ae,7of20       I

ATTACHMENT 1 EAL Bases Category:                 C - Cold Shutdown / Refueling System Malfunction Subcategory:              3 - RCS Temperature Initiating Condition:     UNPLANNED increase in RCS temperature EAL:
CU3.2          Unusual Event Loss of all RCS temperature and RCS level indication for ->15 min. (Note 1)
Note 1:  The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
Mode Applicability:
5 - Cold Shutdown, 6- Refueling Definition(s):
None Basis:
Several instruments are capable of providing indication of RCS temperature with respect to the Technical Specification cold shutdown temperature limit (200&deg;F, ref. 1). These include cold leg (Ta) temperature indications, hot leg (Th) temperature indications with RCPs running, CETCs and LPI cooler outlet temperature indications (ref. 2).
Several instruments are capable of providing indication of RCS level including pressurizer level, RVLIS, LT-5 and local monitor (ref. 3).
This EAL addresses the inability to determine RCS temperature and level, and represents a potential degradation of the level of safety of the plant. Ifthe RCS is not intact and CONTAINMENT CLOSURE is not established during this event, the Emergency Coordinator should also refer to IC CA3.
This EAL reflects. a condition where there has been a significant loss of instrumentation capability necessary to monitor RCS conditions and operators would be unable to monitor key parameters necessary to assure core decay heat removal. During this condition, there is no immediate threat of fuel damage because the core decay heat load has been reduced since the cessation of power operation.
Fifteen minutes was selected as a threshold to exclude transient or momentary losses of indication.
Escalation to Alert would be via IC CA1 based on an inventory loss or IC CA3 based on exceeding plant configuration-specific time criteria.
I[Document No.]                                      Rev. 0                              Page 88 of 240

ATTACHMENT 1 EAL Bases ONS Basis Reference(s):
: 1. ONS Technical Specifications Table 1.1-1
: 2. AP/1 ,2,3/A/1700/026 Loss of Decay Heat Removal
: 3. UFSAR Section 7.5.2.2 Inadequate Core Cooling Instruments
: 4. NEI 99-01 CU3 S[Document NO.]                             Rev. 0           Page 89 of 240

ATTACHMENT 1 EAL Bases Category:                      C  -  Cold Shutdown / Refueling System Malfunction Subcategory:                   3  -  RCS Temperature Initiating Condition:           Inability to maintain plant in cold shutdown EAL:
CA3.1 Alert UNPLANNED increase in RCS temperature to                  >  200&deg;F for > Table C-4 duration (Noteds 1, 10)
OR UNPLANNED RCS pressure increase > 10 psig due to a loss of RCS cooling (this EAL does* not apply during water-solid plant conditions)                     I Note 1:      The Emergency Coordinator should declare the event prom'ptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
Note 10:     In the absence of reliable RCS temperature indication caused by the loss of decay heat removal capability, classification should be based on time to boil data.
Table C-4: RCS Heat-up Duration Thresholds CONTAINMENT RCS Status                     CLOSURE Status                Heat-up Duration Intact (but not REDUCED                                          N/A                         60 main.*
INVENTORY)
Not intact                                   established                     20 min.*
OR REDUCED INVENTORY                          not established                    0 mai.
* If an RCS heat removal system is in operation within this time frame and RCS temperature is being reduced, the EAL is not applicable.
Mode Applicability:
5 - Cold Shutdown, 6 - Refueling Definition(s):
CONTAINMENT CLOSURE - The action to secure containment and its associated structures, systems, and components as a functional barrier to fission product release under all plant conditions up to and including a loss of decay heat removal or fuel handling accident inside containment.
As applied to ONS, Containment Closure is established when the requirements of OP/I1,2,3/A/1502/009, Containment Closure Control, are met.
UNPLANNED -. A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.
S[Document No.]                                         Rev. 0                                Page 90 of 240

ATTACHMENT 1 EAL Bases REDUCED INVENTORY - Condition with fuel in the reactor vessel and the level lower than three feet below the reactor vessel flange (RCS level < 50" on LT-5)
Basis:
Several instruments are capable of providing indication of RCS temperature with respect to the Technical Specification cold shutdown temperature limit (200&deg;F, ref. 1). These include cold leg (Ta) temperature indications, hot leg (Th) temperature indications with RCPs running, CETCs and LPI cooler outlet temperature indications (ref. 2).
However, if L~w Pressure Injection (LPI) flow is lost, the normal termperature elements used to monitor RCS temperature are not accurate indicators of RCS temperature. The CETCs are the design instruments for these conditions. For some periods of time the CETCs may not be available. The current practices concerning determining time to boil can be used in the evaluation of these EALs. Without CETC indication and with a loss of LPI flow the following guidance should be used (ref. 2):
* Use the predetermined "time to boil" data for evaluating these EALs. This approach reflects the relatively small numerical difference between the typical Technical Specification cold shutdown temperature limit of 200&deg;F and the boiling temperature of RCS water with the plant in Mode 5 or 6.
* Alternately, the Control Room staff may use a procedure or user aid to determine when ROS temperature will likely exceed 200&deg;F given the actual plant conditions (e.g., using a heat-up curve).
Numerous RCS pressure instruments are capable of measuring pressure to less than 10 psia including RCS low range cooldown pressure indicators RC-P-0086A/B (ref. 3).
This IC addresses conditions involving a loss of decay heat removal capability or an addition of heat to the RCS in excess of that which can currently be removed. Either condition represents an actual or potential substantial degradation of the level of safety of the plant.
A momentary UNPLANNED excursion above the Technical Specification cold shutdown temperature limit when the heat removal function is available does not warrant a classification.
The RCS Heat-up Duration Thresholds table addresses an increase in RCS temperature when CONTAINMENT CLOSURE is established but the RCS is not intact, or RCS inventory is reduced (e.g., mid-loop operation). The 20-minute criterion was included to allow time for operator action to address the temperature increase.
The RCS Heat-up Duration Thresholds table also addresses an increase in RCS temperature with the RCS intact. The status of CONTAINMENT CLOSURE is not crucial in this condition since the intact RCS is providing a high pressure barrier to a fission product release. The 60-minute time frame should allow sufficient time to address the temperature increase without a substantial degradation in plant safety.
Finally, in the case where there is an increase in RCS temperature, the RCS is not intact or is at reduced inventory, and CONTAINMENT CLOSURE is not established, no heat-up duration is allowed (i.e., 0 minutes). This is because 1) the evaporated reactor coolant may be released directly into the containment atmosphere and subsequently to the environment, and
: 2) there is reduced reactor coolant inventory above the top of irradiated fuel.
[Document No.]                                   Rev. 0                            Page 91 of 240

ATTACHMENT 1 EAL Bases The RCS pressure increase threshold provides a pressure-based indication of RCS heat-up in the absence of RCS temperature monitoring capability.
Escalation of the emergency classification level would be via IC OS1 or RSI.
ONS Basis Reference(s):
: 1. ONS Technical Specifications Table 1.1-1
: 2. AP/I1,2,3/A/1 700/026 Loss of Decay Heat Removal
: 3. IP/1 ,2,3/A/0200/047A Reactor Coolant System LTOP Instrument Calibration
: 4. NEI 99-01 CA3 S[Document No.]                              Rev. 0        1Page                92 of 240

ATTACHMENT 1 EAL Bases Category:                  C  -  Cold Shutdown / Refueling System Malfunction Subcategory:              4 - Loss of Vital DC Power Initiating Condition:      Loss of Vital DC power for 15 minutes or longer EAL:
CU4.1          Unusual Event Indicated voltage is < 10O5VDC on vital DC buses required by Technical Specifications for
  >15 min. (Note 1)I Note 1:   The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
Mode Applicability:
5 - Cold Shutdown, 6 - Refueling Definition(s):
None Basis:
The purpose of this EAL 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. The fifteen minute interval is intended to exclude transient or momentary power losses.
For each unit, two independent and physically separated 125 volt DC batteries and DC buses are provided for the vital instrumentation and control power system. (ref. 1, 2). Minimum DC bus voltage is 110 VDC (ref. 3).
This EAL is the cold condition equivalent of the hot condition loss of DC power EAL SS2.1.
This IC addresses a loss of vital DC power which compromises the ability to monitor and control operable SAFETY SYSTEMS when the plant is in the cold shutdown or refueling mode.
In these modes, the core decay heat load has been significantly reduced, and coolant system temperatures and pressures are lower; these conditions increase the time available to restore a vital DC bus to service. Thus, this condition is considered to be a potential degradation of the level of safety of the plant.
As used in this EAL, "required" means the vital DC buses necessary to support operation of the in-service, or operable, train or trains of SAFETY SYSTEM equipment. For example, if Train A is out-of-service (inoperable) for scheduled outage maintenance work and Train B is in-service (operable), then a loss of Vital DC power affecting Train B would require the declaration of an Unusual Event. A loss of Vital DC power to Train A would not warrant an emergency classification.
Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.
Depending upon the event, escalation of the emergency classification level would be via IC CA1 or CA3, or an IC in Recognition Category R.
I[Document No.]                                       Rev. 0                              Page 93 of 2401

ATTACHMENT 1 EAL Bases ONS Basis Reference(s):
: 1. UFSAR Figure 8.5 Typical DC and AC Vital Power System - Single Line
: 2. UFSAR Section 8.3.2 DC Power Systems
: 3. EP/*/AI1 800/001 Blackout Tab
: 4. Technical Specifications 3.8.4 DC Sources - Shutdown
: 5. NEI 99-01 CU4 S[Document No.]                    [      Rev.. 0       [Page           94 of 2401

ATTACHMENT 1 EAL Bases Category:                  C  -  Coid Shutdown / Refueling System Malfunction Subcategory:                5 -  Loss of Communications Initiating Condition:      Loss of all onsite or offsite communications capabilities EAL:
CU5.1          Unusual Event Loss of all Table C-5 onsite communication methods OR Loss of all Table C-5 offsite communication methods OR Loss of all Table C-5 NRC communication methods Table C-5    Communication Methods System                        Onsite    Offsite     NRC Commercial phone service                               X          X         X ONS site phone system                                  X          X         X EOF phone system                                                  X         X Public Address system                                   X Onsite radio system                                    X DEMNET                                                            X Offsite. radio systemX NRC Emergency Telephone System                                              X Satellite Phone                                                   X Mode Applicability:
5 - Cold Shutdown, 6 - Refueling, NM      -  No Mode Definition(s):
None S[Do~cument No.]                                  Rev.O0                            Page 95 of 240I

ATTACHMENT 1 EAL Bases Basis:
Onsite, offsite and NRC communications include one or more of the systems listed in Table 0-5 (ref. 1).
: 1. Commercial phone service The Commercial phone service does not go through the site telephone system.
: 2. ONS site phone system Trhe site phone system is generator and battery backed with:
* Fiber-Optic to Charlotte GO (65 lines)
* Telephone line to Easley (6 circuits)
I     Anderson (4 lines)
* Six Mile (4 lines)
* Site Telephone System: Inward and outward direct dial available from the Control Room, TSC, and OSC
: 3. EOF phone system The emergency communications systems at the Charlotte EOF are designed to ensure the reliable, timely flow of information between all parties having an emergency response role.
: 4. Public Address (Paging) system The paging system provides paging and party line communications between stations located throughout the plant. Inside and outside type wall and desk-mounted stations are used to communicate between roaming personnel and fixed work locations. Plant-wide instructions are issued using the paging feature.
: 5. Onsite radio system The onsite radio system receives emergency backup power from Keowee Hydro Units supporting communications with: Control Room 1&2, 3, Fire Brigade, Chemistry, Safety, Radiation Protection, Maintenance, Medical Emergency Response Team, and Hazardous Materials Response Team.
: 6. DEMNET DEMNET is the primary means of offsite communication. This circuit allows intercommunication among the EOF, TSC, control room, counties, and states. DEMNET operates as an internet based (VolP) communications system with a satellite back-up. Should the internet transfer rate become slow or unavailable, the DEMNET will automatically transfer to satellite mode.
: 7. Offsite radio system The offsite radio system is battery backed supporting communications with: Control Room Units 1&2, TSC, Field Monitoring Teams, EOF, counties and State of South Carolina.
I[Document No.]                                Rev. 0                           Page 96o&deg;240

ATTACHMENT 1 EAL Bases
: 8. NRC Emergency Telephone System (ETS)
The NRC uses a Duke Energy dedicated telephone line which allows direct telephone communications from the plant to NRC regional and national offices. The Duke Energy communications line provides a link independent of the local public telephone network.
Telephones connected to this network are located in the Oconee Control Rooms, Technical Support Center, and Emergency Operations Facility and can be used to establish NRC Emergency Notification System (ENS) and Health Physics Network (HPN) capability.
: 9. Satellitq Phone Satellite Phones can be used for external communications This EAL is the cold condition equivalent of the hot condition EAL SU7.1.
This IC addresses a significant loss of on-site or offsite communications capabilities. While not a direct challenge to plant or personnel safety, this event warrants prompt notifications to OROs and the NRC.
This IC should be assessed only when extraordinary means are being utilized to make communications possible (e.g., use of non-plant, privately owned equipment, relaying of on-site information via individuals or multiple radio transmission points, individuals being sent to offsite locations, etc.).
The first EAL condition addresses a total loss of the communications methods used in support of routine plant operations.
The second EAL condition addresses a total loss of the communications methods used to notify all OROs of an emergency declaration. The OROs referred to here are the State EOC and FEO, Pickens County LEC and EOC, and Oconee County LEC and EOC.
.The third EAL addresses a total loss of the communications methods used to notify the NRC of an emergency declaration.
ONS Basis Reference(s):
: 1. ONS Emergency Plan, Section 7.2 Communications Systems
: 2. NEI 99-01 CU5 S[Document No.]                                  Rev. 0                             Page 97 of 240

ATTACHMENT 1 EAL Bases Category:                C - Cold Shutdown / Refueling System Malfunction Subcategory:             6 - Hazardous Event Affecting Safety Systems Initiating Condition:    Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode EAL:
CA6.1          Alert The occurrence of ahiy Table C-6 hazardous event AND EITHER:
          *Event damage has caused indications of degraded performance in at least one train of a SAFETY SYSTEM needed for the current operating mfode
          *The event has caused VISIBLE DAMAGE to a SAFETY SYSTEM component or structure needed for the current operating mode Table 0-6    Hazardous Events
* Seismic event (earthquake)
* Internal or external FLOODING event
* High winds or tornado strike
* FIRE
* EXPLOSION
* Other events with similar hazard characteristics as determined by the Shift Manager Mode Applicability:
5 - Cold Shutdown, 6 - Refueling Definition(s):
EXPLOSION - A rapid, violent and catastrophic failure of a piece of equipment due to combustion, chemical reaction or overpressurization. A release of steam (from high energy lines or components) or an electrical component failure (caused by short circuits, grounding, arcing, etc.) should not automatically be considered an explosion. Such events require a post-event inspection to determine if the attributes of an explosion are present.
FIRE - Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred but is NOT required if large quantities of smoke and heat are observed.
FLOODING - A condition where water is entering a room or area faster than installed equipment is capable of removal, resulting in a rise of water level within the room or area.
((Document No.]                                 Rev. 0                            Page 98 of 240

ATTACHMENT 1 EAL Bases SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10OCFR50.2):
Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:
(1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which 6ould result in potential offsite exposures.
VISIBLE DAMAGE - Damage to a component or structure that is readily observable without measurements, testing, or analysis. The visual impact of the damage is sufficient ,to cause concern regarding the operability or reliability of the affected component or structure.
Basis:
* The significance of seismic events are discussed under EAL HU2.1 (ref. 1).
* Internal FLOODING may be caused by events such as component failures, equipment misalignment, or outage activity mishaps.
   *External flooding at ONS is highly unlikely since the probable maximum flood (PMF) would be contained by the Keowee Reservoir. Plant grade elevation is 796.0 ft MSL. The minimum external access elevation for the Auxiliary, Turbine, and Service Buildings is 796.5 ft MSL which provides a 6 inch water sill. (ref. 2)
   *High winds in excess of design (95 mph) or tornado strikes can cause significant structural damage (ref. 3).
   *Areas containing functions and systems required for safe shutdown of the plant are identified by fire area (ref. 4, 5).
   *An explosion that degrades the performance of a SAFETY SYSTEM train or visibly damages a SAFETY SYSTEM component or structure would be classified under this EAL.
This IC addresses a hazardous event that causes damage to a SAFETY SYSTEM, or a structure containing SAFETY SYSTEM components, needed for the current operating mode.
This condition significantly reduces the margin to a loss or potential loss of a fission product barrier, and therefore represents an actual or potential substantial degradation of the level of safety of the plant.
The first conditional addresses damage to a SAFETY SYSTEM train that is in service/operation since indications for it will be readily available. The indications of degraded performance should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train.
The second conditional addresses damage to a SAFETY SYSTEM component that is not in service/operation or readily apparent through indications alone, or to a structure containing SAFETY SYSTEM components. Operators will make this determination based on the totality of available event and damage report information. This is intended to be a brief assessment not requiring lengthy analysis or quantification of the damage.
I[Document NO.]                                  Rev. 0          ]Page99o24

ATTACHMENT 1 EAL Bases Escalation of the emergency classification level would be via IC CS1 or RSI.
ONS Basis Reference(s):
: 1. AP/0/A/1 700/005 Earthquake
: 2. UFSAR Section 3.4.1.1 Flood Protection Measures for Seismic Class I Structures
: 3. UFSAR Section 3.3.1.1 Design Wind Velocity
: 4. OSS-0254.00-00-4008 Design Bases Specification for Fire Protection
: 5. AP/1 ,2,3/A/1 700/050 Challenging Plant Fire
: 6. NEI 99-01 CA6             i S[Document No.]                              Rev. 0                          Page 100 of 240

ATTACHMENT 1 EAL Bases Category H - Hazards and Other Conditions Affecting Plant Safety EAL Group: ANY (EALs in this category are applicable to any plant condition, hot or cold.)
Hazards are non-plant, system-related events that can directly or indirectly affect plant operation, reactor plant safety or personnel safety.
: 1. Security Unauthorized entry attempts into tlie PROTECTED AREA, bomb threats, sabotage attempts, and actual security compromises threatening loss of physical control of the plant.
: 2. Seismic Event Natural events such as earthquakes have potential to cause plant structure or equipment damage of sufficient magnitude to threaten personnel or plant safety.
: 3. Natural or Technology Hazard Other natural and non-naturally occurring events that can cause damage to plant facilities include tornados, FLOODING, hazardous material releases and events restricting site access warranting classification.
: 4. Fire Fires can pose significant hazards to personnel and reactor safety. Appropriate for classification are fires within the Plant PROTECTED AREA or which may affect operability of equipment needed for safe shutdown
: 5. Hazardous Gas Toxic, corrosive, asphyxiant or flammable gas leaks can affect normal plant operations or preclude access to plant areas required to safely shutdown the plant.
: 6. Control Room Evacuation Events that are indicative of loss of Control Room habitability. If the Control Room must be evacuated, additional support for monitoring and controlling plant functions is necessary through the emergency response facilities.
: 7. Emergency Coordinator Judgment The EALs defined in other categories specify the predetermined symptoms or events that are indicative of emergency or potential emergency conditions and thus warrant classification. While these EALs have been developed to address the full spectrum of possible emergency conditions which may warrant classification and subsequent implementation of the Emergency Plan, a provision for classification of emergencies based on operator/management experience and judgment is still necessary. The EALs of this category provide the Emergency Coordinator the latitude to classify emergency conditions consistent with the established classification criteria based upon Emergency Coordinator judgment.
I[Document No.]                        I         Rev. 0        IPage                   101lof 240

ATTACHMENT 1 EAL Bases Category:                 H - Hazards Subcategory:              1 - Security Initiating Condition:     Confirmed SECURITY CONDITION or threat EAL:
HUI.1          Unusual Event A SECURITY CONDITION that does not involve a HOSTILE ACTION as reported by the Security Shift Supervision Mode Applicability:
All Definition(s):
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.
HOSTILE ACTION - An act toward ONS 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 ONS. 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).
Basis:
This EAL is based on the Duke Energy Physical Security Plan for ONS (ref. 1).
This IC addresses events that pose a threat to plant personnel or SAFETY SYSTEM equipment, and thus represent a potential degradation in the level of plant safety. Security events which do not meet one of these EALs are adequately addressed by the requirements of 10 CFR &sect; 73.71 or 10 CFR &sect; 50.72. Security events assessed as HOSTILE ACTIONS are classifiable under ICs HA1, HS1 and HGI.
Timely and accurate communications between Security Shift Supervision and the Control Room is essential for proper classification of a security-related event. Classification of these events will initiate appropriate threat-related notifications to plant personnel and Offsite Response Organizations.
Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan [and Independent Spent Fuel Storage InstallationSecurity Program].
This EAL references the Security Shift Supervision because these are the individuals trained to confirm that a security event is occurring or has occurred. Training on security event confirmation and classification is controlled due to the nature of Safeguards and 10 CER &sect; S[Document No.]                                 Rev. 0                            Page 102 of 240

ATTACHMENT 1 EAL Bases 2.39 information.
Emergency plans and implementing procedures are public documents; therefore, EALs should not incorporate Security-sensitive information. This includes information that may be advantageous to a potential adversary, such as the particulars concerning a specific threat or threat location. Security-sensitive information should be contained in non-public documents such as the Duke Energy Physical Security Plan for ONS (ref. 1).
Escalation of the emergency classification level would be via IC HAl.
ONlS Basis Reference(s):
: 1. Duke Energy Physical Security Plan for ONS
: 2. AP/0/A/1 700/045 Site Security Threats
: 3. NEI 99-01 HUI S[Document No.]                               Rev. 0                          Page 103 of 240

ATTACHMENT 1 EAL Bases Category:                 H - Hazards Subcategory:              1 - Security Initiating Condition:     Confirmed SECURITY CONDITION or threat EAL:
HUI.2           Unusual Event Notification of a credible security threat directed at the site Mode App~licability:
All Definition(s):
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.
Basis:
This EAL is based on the Duke Energy Physical Security Plan for ONS (ref. 1).
This IC addresses events that pose a threat to plant personnel or SAFETY SYSTEM equipment, and thus represent a potential degradation in the level of plant safety. Security events which do not meet one of these EALs are adequately addressed by the requirements of 10 CFR &sect; 73.71 or 10 CFR &sect; 50.72. Security events assessed as HOSTILE ACTIONS are classifiable under l~s HAl, HS1 and HGI.
Timely and accurate communications between Security Shift Supervision and the Control Room is essential for proper classification of a security-related event. Classification of these events will initiate appropriate threat-related notifications to plant personnel and Offsite Response Organizations.
Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan.[and Independent Spent Fuel Storage Installation Security Program].
This EAL addresses the receipt of a credible security threat. The credibility of the threat is assessed in accordance with the Duke Energy Physical Security Plan for ONS.
Emergency plans and implementing procedures are public documents; therefore, EALs should not incorporate Security-sensitive information. This includes information that may be advantageous to a potential adversary, such as the particulars concerning a specific threat or threat location. Security-sensitive information should be contained in non-public documents such as the Duke Energy Physical Security Plan for ONS (ref. 1).
Escalation of the emergency classification level would be via IC HA1.
S[Document No.]                                 Rev. 0                            Page 104 of 240

ATTACHMENT 1 EAL Bases ONS Basis Reference(s):
: 1. Duke Energy Physical Security Plan for ONS
: 2. AP/0/A/1 700/045 Site Security Threats
: 3. NEI 99-01 HU1 S[Document No.]                    I        Rev. 0  Pag~e 105 of 240

ATTACHMENT 1 EAL Bases Category:                 H - Hazards Subcategory:              1 - Security Initiating Condition:     Confirmed SECURITY CONDITION or threat EAL:
HUI.3           Unusual Event A validated notification from the NRC providing information of an aircraft threat Mode Applicability:
All Definition(s):
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.
Basis:
This EAL is based on the Duke Energy Physical Security Plan for ONS (ref. 1).
This IC addresses events that pose a threat to plant personnel or SAFETY SYSTEM equipment, and thus represent a potential degradation in the level of plant safety. Security events which do not meet one of these EALs are adequately addressed by the requirements of 10 CFR &sect; 73.71 or 10 CFR &sect; 50.72. Security events assessed as HOSTILE ACTIONS are classifiable under ICs HAl, HS1 and HG1.
Timely and accurate communications between Security Shift Supervision and the Control Room is essential for proper classification of a security-related event. Classification of these events will initiate appropriate threat-related notifications to plant personnel and Offsite Response Organizations.
Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification.Plan,Safeguards Contingency Plan [and Independent Spent Fuel Storage InstallationSecurity Program]..
This EAL addresses the threat from the impact of an aircraft on the plant. The NRC Headquarters Operations Officer (HOO) will communicate to the licensee if the threat involves an aircraft. The status and size of the plane may also be provided by NORAD through the NRC. Validation of the threat is performed in accordance with AP/OIAI17001045 Site Security Threats (ref. 2).
Emergency plans and implementing procedures are public documents; therefore, EALs should not incorporate Security-sensitive information. This includes information that may be advantageous to a potential adversary, such as the particulars concerning a specific threat or threat location. Security-sensitive information should be contained in non-public documents such as the Duke Energy Physical Security Plan for ONS (ref. 1).
Escalation of the emergency classification level would be via IC HAl.
I[Document No.]                         .        Rev. 0                            Page 106 of.240

ATTACHMENT 1 EAL Bases ONS Basis Reference(s):
: 1. Duke Energy Physical Security Plan for ONS
: 2. AP/0/A/1700/045 Site Security Threats
: 3. NEI 99-01 HU1 S[Document No.]                             Rev. 0 Page 107 of 240I

ATTACHMENT 1 EAL Bases Category:                 H - Hazards Subcategory:              1 - Security Initiating Condition:     HOSTILE ACTION within the OWNER CONTROLLED AREA or airborne attack threat within 30 minutes EAL:
HAl.1          Alert A HOSTILE ACTION is occurring or has occdurred within the OWNER CONTROLLED AREA as reported by the Security Shift Supervision Mode Applicability:
All Definition(s):
HOSTILE ACTION - An act toward ONS 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 ONS. 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).
OWNER CONTROLLED AREA - Area outside the PROTECTED AREA fence that immediately surrounds the plant. Access to this area is generally restricted to those entering on official business.
Basis:
This IC addresses the occurrence of a HOSTILE ACTION within the OWNER CONTROLLED AREA or notification of an aircraft attack threat. This event will require rapid response and assistance due to the possibility of the attack progressing to the PROTECTED AREA, or the need to prepare the plant and staff for a potential aircraft impact.
Timely and accurate communications between Security Shift Supervision and the Control Room is essential for proper classification of a security-related event.
Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan [and In dependent Spent Fuel Storage Installation Security Program].
As time and conditions allow, these events require a heightened state of readiness by the plant staff and implementation of onsite protective measures (e.g., evacuation, dispersal or sheltering). The Alert declaration will also heighten the awareness of Offsite Response Organizations (OROs), allowing them to be better prepared should it be necessary to consider further actions.
This IC does not apply to incidents that are accidental events, acts of civil disobedience, or otherwise are not a HOSTILE ACTION perpetrated by a HOSTILE FORCE. Examples include I[Document No.]                        I        Rev.O0                            Page 108 of 2401

ATTACHMENT 1 EAL Bases the crash of a small aircraft, shots from hunters, physical disputes between employees, etc.
Reporting of these types of events is adequately addressed by other EALs, or the requirements of 10 CFR &sect; 73.71 or 10 CFR &sect; 50.72.
This EAL is applicable for any HOSTILE ACTION occurring, or that has occurred, in the OWNER CONTROLLED AREA. This includes any action directed against an ISFSI that is located outside the plant PROTECTED AREA.
Emergency plans and implementing procedures are public documents; therefore, EALs should not incorporate Security-sensitive information. This includes information that may be tadvantageous to a potential adversary, such as the Ij*articulars concerning a specific threat or threat location. Security-sensitive information should be contained in non-public documents such as the Duke Energy Physical Security Plan for ONS (ref. 1).
SEscalation of the emergency classification level would be via IC HSI.
ONS Basis Reference(s):
: 1. Duke Energy Physical Security Plan for ONS
: 2. AP/O/NI1700/045 Site Security Threats
: 3. NEI 99-01 HAl I[Document No.]                                Rev. 0          IPage                  109 of 240I

ATTACHMENT 1 EAL Bases Category:                 H - Hazards Subcategory:              1 - Security Initiating Condition:     HOSTILE ACTION within the OWNER CONTROLLED AREA or airborne attack threat within 30 minutes EAL:
HA1.2          Alert iAvalidated notification from NRC of an aircraft attaclk threat within 30 min. of the site Mode Applicability:
*All Definition(s):
HOSTILE ACTION - An act toward ONS 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 ONS. 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).
Basis:
This IC addresses the occurrence of a HOSTILE ACTION within the OWNER CONTROLLED AREA or notification of an aircraft attack threat. This event will require rapid response and assistance due to the possibility of the attack progressing to the PROTECTED AREA, or the need to prepare the plant and staff for a potential aircraft impact.
Timely and accurate communications between Security Shift Supervision and the Control Room is essential for proper classification of a security-related event.
Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and QualificationPlan, Safeguards Contingency Plan [and Independent Spent Fuel Storage Installation Security Program].
As time and conditions allow, these events require a heightened state of readiness by the plant staff and implementation of onsite protective measures (e.g., evacuation, dispersal or sheltering). The Alert declaration will also heighten the awareness of Offsite Response Organizations (OROs), allowing them to be better prepared should it be necessary to consider further actions.
This IC does not apply to incidents that are accidental events, acts of civil disobedience, or otherwise are not a HOSTILE ACTION perpetrated by a HOSTILE FORCE. Examples include the crash of a small aircraft, shots from hunters, physical disputes between employees, etc.
Reporting of these types of events is adequately addressed by other EALs, or the requirements of 10 CFR &sect; 73.71 or 10 CFR &sect; 50.72.
I[Document No.1                                 Rev. 0                            Page 110 of 240I

ATTACHMENT 1 EAL Bases This EAL addresses the threat from the impact of an aircraft on the plant, and the anticipated arrival time is within 30 minutes. The intent of this EAL is to ensure that threat-related notifications are made in a timely manner so that plant personnel and OROs are in a heightened state of readiness. This EAL is met when the threat-related information has been validated in accordance with AP/0/AII1700/045 Site Security Threats (ref. 2).
The NRC Headquarters Operations Officer (HOO) will communicate to the licensee if the threat involves an aircraft. The status and size of the plane may be provided by NORAD through the NRC.
In some Icases, it may not be readily apparent if an aircraft imlbact within the OWNER CONTROLLED AREA was intentional (i.e., a HOSTILE ACTION). It is expected, although not certain, that notification by an appropriate Federal agency to the site would clarify this point. In this case, the appropriate federal agency is intended to be NORAD, FBI, FAA or NRC. The emergency declaration, including one based on other ICs/EALs, should not be unduly delayed while awaiting notification by) a Federal agency.
Emergency plans and implementing procedures are public documents; therefore, EALs should not incorporate Security-sensitive information. This includes information that may be advantageous to a potential adversary, such as the particulars concerning a specific threat or threat location. Security-sensitive information should be contained in non-public documents such as the Duke Energy Physical Security Plan for ONS (ref. I).
Escalation of the emergency classification level would be via IC HSl.
ONS Basis Reference(s):
: 1. Duke Energy Physical Security Plan for ONS
: 2. AP/0/A/1700/045 Site Security Threats
: 3. NEI 99-01 HA1 S[Document No.] .Rev.                               0                            Page 111 of 240

ATTACHMENT 1 EAL Bases Category:                 H - Hazards Subcategory:              1 - Security Initiating Condition:     HOSTILE ACTION within the PROTECTED AREA EAL:
HSI.1          Site Area Emergency A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by the Security Shift Supervision Mode Applicability:
All Definition(s):
HOSTILE ACTION - An act toward ONS 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 ONS. 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).
PROTECTED AREA - That part of the nuclear station property consisting of the Reactor, Auxiliary, Turbine, and Service Building and grounds, contained within the owner controlled security fence.
Basis:
The Security Shift Supervision are the designated on-site personnel 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 Duke Energy Physical Security Contingency Plan for ONS (Safeguards) information. (ref. 1)
This IC addresses the occurrence of a HOSTILE ACTION within the PROTECTED AREA.
This event will require rapid response and assistance due to the possibility for damage to plant equipment.
Timely and accurate communications between Security Shift Supervision and the Control Room is essential for proper classification of a security-related event (ref. 1, 2).
Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan [and Independent Spent Fuel Storage Installation Security Program].
As time and conditions allow, these events require a heightened state of readiness by the plant staff and implementation of onsite protective measures (e.g., evacuation, dispersaror sheltering). The Site Area Emergency declaration will mobilize Offsite Response Organization (ORO) resources and have them available to develop and implement public protective actions in the unlikely event that the attack is successful in impairing multiple safety functions.
I[Document No.]                                  Rev. 0                            Page 112 of 240I

ATTACHMENT 1 EAL Bases This IC does not apply to a HOSTILE ACTION directed at an ISFSI PROTECTED AREA located outside the PROTECTED AREA; such an attack should be assessed using IC HAl. It also does not apply to incidents that are accidental events, acts of civil disobedience, or otherwise are not a HOSTILE ACTION perpetrated by a HOSTILE FORCE. Examples include the crash of a small aircraft, shots from hunters, physical disputes between employees, etc.
Reporting of these types of events is adequately addressed by other EALs, or the requirements of 10 CFR &sect; 73.71 or 10 CFR &sect; 50.72.
Emergency plans and implementing procedures are public documents; therefore, EALs should not incorporate Security-sensitive information. This includes information that npay be advantageous to a potential adversary, such as the particulars concerning a specific threat or threat location. Security-sensitive information should be contained in non-public documents such as the Duke Energy Physical Security Plan for ONS (ref. 1).
Escalation of the emergency classification level would be via IC HG1.
ONS Basis Reference(s):
: 1. Duke Energy Physical Security Plan for ONS
: 2. AP/0/A/1700/045 Site Security Threats
: 3. NEI 99-01 HS1
[Document No.]                                Rev. 0                            Page 113 of 240

ATTACHMENT 1 EAL Bases Category:                   H - Hazards Subcategory:                 1 - Security Initiating Condition:       HOSTILE ACTION resulting in loss of physical control of the facility EAL:
HGI.1           General Emergency A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by the Security Shift Supprvision AND EITHER of the following has occurred:
Any of the following safety functions cannot be controlled or maintained
* Reactivity
* Core cooling
* RCS heat removal OR Damage to spent fuel has occurred or is IMMINENT Mode Applicability:
All Definition(s):
HOSTILE ACTION - An act toward ONS 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 ONS. 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).
IMMINENT - The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions PROTECTED AREA - That part of the nuclear station property consisting of the Reactor, Auxiliary, Turbine, and Service Building and grounds, contained within the owner controlled security fence.
Basis:
Indications of damaged spent fuel are provided in AP/1 ,2,3/A/1 700/009 Spent Fuel Damage (ref. 4).
This IC addresses an event       in which a HOSTILE FoRCE has taken physical control of the facility to the extent that the  plant staff can no longer operate equipment necessary to maintain key safety functions. It also    addresses a HOSTILE ACTION leading to a loss of physical control that results in actual   or IMMINENT damage to spent fuel due to 1) damage to a spent I[Document No.]                           I        Rev. 0                         Page 114 of 240I

ATTACHMENT 1 EAL Bases fuel pool cooling system (e.g., pumps, heat exchangers, controls, etc.) or, 2) loss of spent fuel pool integrity such that sufficient water level cannot be maintained.
Timely and accurate communications between the Security Shift Supervision and the Control Room is essential for proper classification of a security-related event (ref. 2, 3).
Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Trainingand Qualification Plan, Safeguards Contingency Plan [and Independent Spent Fuel Storage Installation Security Program].
Emergency plans and implementing procediures are public documents; therefore, EALs should not incorporate Security-sensitive information. This includes information that may be advantageous to a potential adversary, such as the particulars concerning a specific threat or threat location. Security-sensitive information should be contained in non-public documents such as the Duke Energy Physical Security Plan for ONS (ref.1).
ONS Basis Reference(s):
: 1. Duke Energy Physical Security Plan for ONS
: 2. AP/0/A/1 700/045 Site Security Threats
: 3. APIOIAI17001046 Extensive Damage Mitigation
: 4. AP/1 ,2,3/A/1700/009 Spent Fuel Damage
: 5. NEI 99-01 HG1
* [Document No.]                                  Rev. 0                           Page115 of 240

ATTACHMENT 1 EAL Bases Category:                H - Hazards and Other Conditions Affecting Plant Safety Subcategory:              2 - Seismic Event Initiating Condition:    Seismic event greater than OBE levels EAL:
HU2.1          Unusual Event Seismic event > OBE as indicated by EITHER of the following:
* ISA-9/E-1 (SEISMIC TRIGGER) alarm
* 35A-9/E-1 (SEISMIC TRIGGER) alarm Mode Applicability:
All Definition(s):
None Basis:
This EAL is based on a VALID receipt of either of the specified seismic trigger alarms.
For ONS, the Operating Basis Earthquake (OBE) is equivalent to the Design Basis Earthquake (DBE). The design basis earthquake ground acceleration at the site is 0.05g. The maximum hypothetical earthquake ground acceleration is 0.10g and 0.15g for Class 1 structures founded on bedrock and overburden respectively. For ONS, the Operating Basis Earthquake (OBE) is equivalent to the Design Basis Earthquake (DBE). (ref. 1)
If an earthquake of > 0.05 g has occurred on site, all units are required to be shut down to Mode 5 once a plant damage assessment is complete along with the completion of any needed repairs to support the units ability to achieve safe shutdown. (ref. 2)
Earthquake instrumentation is the SMA-3 system consisting of a central recording system, control panel, one TS-3 triaxial seismic trigger package, and two force-balance triaxial accelerometer packages. The seismic trigger and one accelerometer of the SMA-3 system are located in the Unit 1 Tendon Gallery. Also, a second accelerometer is located directly above at elevation 797' +6" in the Oconee 1 Reactor Building. The recorder for the system is located in the Unit 1 Cable Room. Also, a seismic trigger/switch is located in the Unit 1 tendon gallery.
The TS-3 has a preset acceleration threshold of 0.05g which activates the statalarm in Units 1 and 3 control rooms, when design conditions occur. (ref. 3)
To avoid inappropriate emergency classification resulting from spurious actuation of the seismic instrumentation or felt motion not attributable to seismic activity, an offsite agency (USGS, National Earthquake Information Center (NEIC)) can confirm that an earthquake has occurred in the area of the plant. Such confirmation should not, however, preclude a timely emergency declaration based on receipt of the OBE alarm. The NEIC can be contacted by calling (303) 273-8500 (ref. 2). Select option #1 and inform the analyst you wish to confirm recent seismic activity in the vicinity of ONS. If requested, provide the analyst with the S[Document No.]                                 Rev. 0         .Page                    116 of 240,

ATTACHMENT 1 EAL Bases following ONS coordinates:     340 47' 38.2" north latitude, 820 53' 55.4" west longitude (ref. 4).
Alternatively, near real-time seismic activity can be accessed via the NEIC website:
http://earthqaeusgs.gov/eqcenterI This IC addresses a seismic event that results in accelerations at the plant site greater than those specified for an Operating Basis Earthquake (OBE). An earthquake greater than an OBE but less than a Safe Shutdown Earthquake (SSE) should have no significant impact on safety-related systems, structures and components; however, some time may be required for the plant staff to ascertain the actual post-event condition of the plant (e.g., performs walk-downsI and post-event inspections). Given the time necessary to perform walk-downs and inspections, and fully understand any impacts, this event represents a potential degradation of the level of safety of the plant.
Event qverification with external sources should not be necessary during or following an OBE.
Earthquakes of this magnitude should be readily felt by on-site personnel and recognized as a seismic event (e.g., lateral accelerations in excess of 0.05g). The Shift Manager or Emergency Coordinator may seek external verification if deemed appropriate (e.g., a call to the USGS, check internet news sources, etc.); however, the verification action must not preclude a timely emergency declaration.
Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via IC CA6 or SA9.
ONS Basis Reference(s):
: 1. UFSAR Section 3.2.1.3 Seismic Loading Conditions
: 2. APIO/AI17001005 Earthquake
: 3. UFSAR Section 3.7.4 Seismic Instrumentation Program
: 4. UFSAR Section 2.1.1.1 Specification of Location
: 5. NEI 99-01 HU2 l[Document No.]                                 Rev. 0                             Page117 of 240

ATTACHMENT 1 EAL Bases Category:                  H - Hazards and Other Conditions Affecting Plant Safety Subcategory:               3 - Natural or Technological Hazard Initiating Condition:      Hazardous event EAL:
HU3.1          Unusual Event A tornado strike within the PROTECTED AREA Mode Applicability:
All Definition(s):   I PROTECTEDAREA - That part of the nuclear station property consisting of the Reactor, Auxiliary, Turbine, and Service Building and grounds, contained within the owner controlled security fence.
Basis:
Response actions associated with a tornado onsite is provided in AP/0/A/1700/006, Natural Disaster (ref. 1).
If damage is confirmed visually or by other in-plant indications, the event may be escalated to an Alert under EAL CA6.1 or SA9.1.
A tornado striking (touching down) within the PROTECTED AREA warrants declaration of an Unusual Event regardless of the measured wind speed at the meteorological tower. A tornado is defined as a violently rotating column of air in contact with the ground and extending from the base ofa thunderstorm.
This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant.
EAL HU3.1 addresses a tornado striking (touching down) within the PROTECTED AREA.
Escalation of the emergency classification level would be based on ICs in Recognition Categories R, F, S or C.
ONS Basis Reference(s):
: 1. AP/O/A/1 700/006 Natural Disaster
: 2. NEI 99-01 HU3 I[Document No.]                                 Rev. 0                           page 118 of 240

ATTACHMENT 1 EAL Bases Category:                  H - Hazards and Other Conditions Affecting Plant Safety Subcategory:               3 - Natural or Technological Hazard Initiating Condition:       Hazardous event EAL:
HU3.2           Unusual Event Internal room or area FLOODING of a magnitude sufficient to require manual or automatic electrical isolation of a SAFETY SYSTEM component needed for the current1 operating mode Mode Applicability:
All Definition(s):
FLOODING - A condition where water is entering a room or area faster than installed equipment is capable of removal, resulting in a rise of water level within the room or area.
SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2):
Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:
(1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.
Basis:
Areas susceptible to internal flooding are the Turbine Building and Auxiliary Building (ref.1, 2).
Refer to EAL CA6.1 or SA9.1 for internal flooding affecting one or more SAFETY SYSTEM trains.
This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant.
This EAL addresses FLOODING of a building room or area that results in operators isolating power to a SAFETY SYSTEM component due to water level or other wetting concerns.
Classification is also required if the water level or related wetting causes an automatic isolation of a SAFETY SYSTEM component from its power source (e.g., a breaker or relay trip). To warrant classification, operability of the affected component must be required by Technical Specifications for the current operating mode.
Escalation of the emergency classification level would be based on ICs in Recognition Categories R, F, SorC.
ONS Basis Reference(s):
I[Document No.]                                   Rev.O0                            Page 119 of 240

ATTACHMENT 1 EAL Bases
: 1. AP/1,2,3/A/1700/010 Turbine Building Flood
: 2. AP/1-2,3/A/1700/030 Auxiliary Building Flood
: 3. NEI 99-01 HU3 S[Document No.]                            Rev. 0  Page 120 of 240

ATTACHMENT 1 EAL Bases Category: *H                 - Hazards and Other Conditions Affecting Plant Safety Subcategory:               3-  Natural or Technological Hazard Initiating Condition:      Hazardous event EAL:
HU3.3           Unusual Event Movement of personnel within the PROTECTED AREA is IMPEDED due to an offsite event involving hazardous mater~ials (e.g., an offsite chemical spill or toxic gas releas$)
Mode Applicability:
All Definition(s):
IMPEDE(D) - Personnel access to a room or area is hindered to an extent that extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., requiring use of protective equipment, such as SCBAs, that is not routinely employed).
PROTECTED AREA - That part of the nuclear station property consisting of the Reactor, Auxiliary, Turbine, and Service Building and grounds, contained within the owner controlled security fence.
Basis:
As used here, the term "offsite" is meant to be areas external to the ONS PROTECTED AREA.
This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant.
This EAL addresses a hazardous materials event originating at an offsite location and of sufficient magnitude to impede the movement of personnel within the PROTECTED AREA.
Escalation of the emergency classification level would be based on ICs in Recognition Categories R, F, S or C.
ONS Basis Reference(s):
: 1. NEI 99-01 HU3 S[Document No.]                                Rev. 0                            Page 121   of 240I

ATTACHMENT 1 EAL Bases Category:                  H - Hazards and Other Conditions Affecting Plant Safety Subcategory:               3 - Natural or Technological Hazard Initiating Condition:      Hazardous event EAL:
Note 7:  This EAL does not apply to routine traffic impediments such as fog, snow, ice, or vehicle breakdowns or accidents.
Mode Applicability:I All Definition(s):
None Basis:
This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant.
This EAL addresses a hazardous event that causes an on-site impediment to vehicle movement and significant enough to prohibit the plant staff from accessing the site using personal vehicles. Examples of such an event include site FLOODING caused by a hurricane, heavy rains, up-river water releases, dam failure, etc., or an on-site train derailment blocking the access road.
This EAL is not intended apply to routine impediments such as fog, snow, ice, or vehicle breakdowns or accidents, but rather to more significant conditions such as the Hurricane Andrew strike on Turkey Point in 1992, the flooding around the Cooper Station during the Midwest floods of 1993, or the flooding around Ft. Calhoun Station in 2011.
Escalation of the emergency classification level would be based on ICs in Recognition Categories R, F, S or C.
ONS Basis Reference(s):
: 1. NEI 99-01 HU3 I[Document No.]                                       Rev. 0                                Page 122 of 240I

ATTACHMENT I EAL Bases Category:                   H - Hazards and Other Conditions Affecting Plant Safety Subcategory:                4- Fire Initiating Condition:       FIRE potentially degrading the level of safety of the plant EAL:
HU4.I           Unusual Event A FIRE is not extinguished within 15 min. of any of the following FIRE detection indications (Note 1):
* Report from the field (i.e., visual observation)
* Receipt of multiple (more than 1) fire alarms or indications
* Field verification of a single fire alarm n AND The FIRE is located within any Table H-I area Note 1:   The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
Table H-I Fire Areas
* Reactor Building
* Auxiliary Building
* Turbine Building
* Standby Shutdown Facility
* Intake Structure
* Electrical Blockhouse
* Keowee Hydro & associated transformers
* Transformer Yard
* Protected Service Water Building
* Essential Siphon Vacuum Building Mode Applicability:
All Definition(s):
FIRE - Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred but is NOT required if large quantities of smoke and heat are observed.
Basis:
The 15 minute requirement begins with a credible notification that a fire is occurring, or receipt of multiple valid fire detection system alarms or field validation of a single fire alarm. The alarm is to be validated using available Control Room indications or alarms to prove that it is not spurious, or by reports from the field.
I[Document No.]                                       Rev. 0                            Page 123 of 240I

ATTACHMENT I EAL Bases Table H-I Fire Areas are those areas that contain equipment necessary for safe operation and shutdown of the plant (ref. 1, 2).
This IC addresses the magnitude and extent of FIRES that may be indicative of a potential degradation of the level of safety of the plant.
For EAL HU4.1 the intent of the 15-minute duration is to size the FIRE and to discriminate against small FIRES that are readily extinguished (e.g., smoldering waste paper basket). In addition to alarms, other indications of a FIRE could be a drop in fire main pressure, automatic activation of a suppression system, etc.
Upon receipt, operators will take prompt actions to confirm the validity of an initial fire alarm, indication, or report. For EAL assessment purposes, the emergency declaration clock starts at the time that the initial alarm, indication, or report was received, and not the time that a subsequent verification action was performed. Similarly, the fire duration clock also starts at the time of receipt of the initial alarm, indication or report.
Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via IC CA6 or SA9.
ONS Basis Reference(s):
: 1. OSS-0254.00-00-4008 Design Bases Specification for Fire Protection
: 2. AP/1,2,3/A/1700/050 Challenging Plant Fire
: 3. NEI 99-01 HU4 S[Document No.]                                   Rev. 0        [Page                   124 of 240I

ATTACHMENT 1 EAL Bases Category:                  H - Hazards and Other Conditions Affecting Plant Safety Subcategory:               4 - Fire Initiating Condition:      FIRE potentially degrading the level of safety of the plant EAL:
HU4.2          Unusual Event Receipt of a single fire alarm (i.e., no other indications of a FIRE)
AND        I The fire alarm is indicating a FIRE within any Table H-i area AND The existence of a FIRE is not verified within 30 mm. of alarm receipt (Note 1)
Note 1:    The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
Table H-I Fire Areas
* Reactor Building
* Auxiliary Building
* Turbine Building
* Standby Shutdown Facility
* Intake Structure
* Electrical Blockhouse
* Keowee Hydro & associated transformers
* Transformer Yard
* Protected Service Water Building
* Essential Siphon Vacuum Building Mode Applicability:
All Definition(s):
FIRE - Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred but is NOT required if large quantities of smoke and heat are observed.
Basis:
The 30 minute requirement begins upon receipt of a single valid fire detection system alarm.
The alarm is to be validated using available Control Room indications or alarms to prove that it is not spurious, or by reports from the field. Actual field reports must be made within the 30 minute time limit or a classification must be made. If a fire is verified to be occurring by field report, classification shall be made based on EAL HU4.1.
S[Document No.]                                       Re.0Page                                  125 of 240I

ATTACHMENT 1 EAL Bases Control Room indications that may be used to validate a single fire alarm include (ref. 3):
  -     Remote camera system
  -     CRD service structure air temperature
  -     PZR tailpipe temperature
  -     RB dome temperature
    -RBCU        inlet and outlet temperatures
    -RCP       parameters
    -Status      lights of components located inside RB Table F Fire Areas *re those areas that contain equipment necessary fo&safe operation and shutdown of the plant (ref. 1, 2).
The ONS Fire Protection Program is based on 10 CFR 50.48 (a) and (c) requiring compliance with NFPA 805. The NFPA 805 based Fire Protection Program requirements provide are consistent with the NEI 99-01 basis stated below (ref. 1, 4).
This IC addresses the magnitude and extent of FIRES that may be indicative of a potential degradation of the level of safety of the plant.
This EAL addresses receipt of a single fire alarm, and the existence of a FIRE is not verified (i.e., proved or disproved) within 30-minutes of the alarm. Upon receipt, operators will take prompt actions to confirm the validity of a single fire alarm. For EAL assessment purposes, the 30-minute clock starts at the time that the initial alarm was received, and not the time that a subsequent verification action was performed.
A single fire alarm, absent other indication(s) of a FIRE, may be indicative of equipment failure or a spurious activation, and not an actual FIRE. For this reason, additional time is allowed to verify the validity of the alarm. The 30-minute period is a reasonable amount of time to determine if an actual FIRE exists; however, after that time, and absent information to the contrary, it is assumed that an actual FIRE is in progress.
If an actual FIRE is verified by a report from the field, then HU4.1 is immediately applicable, and the emergency must be declared if the FIRE is not extinguished within 15-minutes of the report. If the alarm is verified to be due to an equipment failure or a spurious activation, and this verification occurs within 30-minutes of the receipt of the alarm, then this EAL is not applicable and no emergency declaration is warranted.
Basis-Related Requirements from Appendix R Appendix R to 10 CFR 50, states in part:
Criterion 3 of Appendix A to this part specifies that "Structures, systems, and components important to safety shall be designed and located to minimize, consistent with other safety requirements, the probability and effect of fires and explosions."
When considering the effects of fire, those systems associated with achieving and maintaining safe shutdown conditions assume major importance to safety because damage to them can lead to core damage resulting from loss of coolant through boil-off.
Because fire may affect safe shutdown systems and because the loss of function of systems used to mitigate the consequences of design basis accidents under post-fire conditions does not per se impact public safety, the need to limit fire damage to systems required to achieve and maintain safe shutdown conditions is greater than the need to I[Document No.]                                    Rev. 0                            Page 1.26 of 240I

ATTACHMENT 1 EAL Bases limit fire damage to those systems required to mitigate the consequences of design basis accidents.
In addition, Appendix R to 10 CFR 50, requires, among other considerations, the use of 1-hour fire barriers for the enclosure of cable and equipment and associated non-safety circuits of one redundant train (G.2.c). As used in HU4.2, the 30-minutes to verify a single alarm is well within this worst-case 1-hour time period.
Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via IC CA6 or SA9.
ONS Basis Reference(s):
: 1. OSS-0254.00-00-4008 Design Bases Specification for Fire Protection
: 2. AP/1 ,2,3/A/1700/050 Challenging Plant Fire
: 3. OP/i1,2,31AI61011003          I
: 4. NRC Letter to T. Preston Gillespie (Duke); ONS Units 1, 2, and 3, Issuance of Amendments Regarding Transition to a Risk-Informed, Performance-Based Fire Protection Program in Accordance With 10 CFR 50.48(c); dated December 29, 2010
: 5. NEI 99-01 HU4 S[Document No.]                               Rev. 0                          Page 127 of 240

ATTACHMENT 1 EAL Bases Category:                  H - Hazards and Other Conditions Affecting Plant Safety Subcategory:               4- Fire Initiating Condition:      FIRE potentially degrading the level of safety of the plant EAL:
HU4.3          Unusual Event A FIRE within the PROTECTED AREA not extinguished within 60 min. of the initial report, alarm or indication (Note 1)
Note 1:   The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
Mode Applicability:
All Definition(s):
FIRE - Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred but is NOT required if large quantities of smoke and heat are observed.
PROTECTEDAREA - That part of the nuclear station property consisting of the Reactor, Auxiliary, Turbine, and Service Building and grounds, contained within the owner controlled security fen ce.
Basis:
This IC addresses the magnitude and extent of FIRES that may be indicative of a potential degradation of the level of safety of the plant.
In addition to a FIRE addressed by EAL HU4.1 or HU4.2, a FIRE within the plant PROTECTED AREA not extinguished within 60-minutes may also potentially degrade the level of plant safety.
Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via IC CA6 or SA9.
ONS Basis Reference(s):
: 1. NEI 99-01 HU4 r[Document No.]                                       Rev. 0                            Page 128 of 240

ATTACHMENT 1 EAL Bases Category:                 H - Hazards and Other Conditions Affecting Plant Safety Subcategory:              4 - Fire Initiating Condition:      FIRE potentially degrading the level of safety of the plant EAL:
HU4.4           Unusual Event A FIRE within the PROTECTED AREA that requires firefighting support by an offsite fire response agency to extinguish Mode Applicability:
All Definition(s):
FIRE - Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred but is NOT required if large quantities of smoke and heat are observed.
PROTECTED AREA - That part of the nuclear station property consisting of the Reactor, Auxiliary, Turbine, and Service Building and grounds, contained within the owner controlled security fence.
Basis:
This IC addresses the magnitude and extent of FIRES that may be indicative of a potential degradation of the level of safety of the plant.
If a FIRE within the PLANT PROTECTED AREA is of s~ufficient size to require a response by an offsite firefighting agency (e.g., a local town Fire Department), then the level of plant safety is potentially degraded. The dispatch of an offsite firefighting agency to the site requires an emergency declaration only if it is needed to actively support firefighting efforts because the fire is beyond the capability of the Fire Brigade to extinguish. Declaration is not necessary if the agency resources are placed on stand-by, or supporting post-extinguishment recovery or investigation actions.
Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via IC CA6 or SA9.
ONS Basis Reference(s):
: 1. NEI 99-01 HU4 I[D~cument No.]1                                  Rev. 0                           Page 129 of 240

ATTACHMENT 1 EAL Bases Category:                       H - Hazards and Other Conditions Affecting Plant Safety Subcategory:                    5 - Hazardous Gases Initiating Condition:           Gaseous release IMPEDING access to equipment necessary for normal plant operations, cooldown or shutdown EAL:
HA5.1              Alert Release of a toxic, corrosive, asphyxiant or flanmmable gas into any Table H-2 rooms or areas AND Entry into the room or area is prohibited or IMPEDED (Note 5)*
Note 5:    If the equipment in the listed room or area was already inoperable or out-of-service before the event occurred, then no emergency classification is warranted.
Table H-2           Safe Operation & Shutdown RoomslAreas Room/Area                                        Mode Applicability Turbine Building                                                                         1, 2, 3 Equipment and Cable Rooms                                                                1, 2, 3 Auxiliary Building                                                                    1, 2, 3,4, 5 Reactor Buildings                                                                        3, 4, 5 Mode Applicability:
All Definition(s):
IMPEDE(D) - Personnel access to a room or area is hindered to an extent that extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., requiring use of protective equipment, such as SCBAs, that is not routinely employed).
Basis:
If the equipment in the listed room or area was already inoperable, or out-of-service, before the event, occurred, then no emergency should be declared since the event will have no adverse impact beyond that already allowed by Technical Specifications at the time of the event.
The .list of plant rooms or areas with entry-related mode applicability identified specify those rooms or areas that contain equipment which require a manual/local action as specified in operating procedures used for normal plant operation, cooldown and shutdown. Rooms or areas in which actions of a contingent or emergency nature would be performed (e.g., an action to address an off-normal or emergency condition such as emergency repairs, corrective measures or emergency operations) are not included. In addition, the list specifies the plant mode(s) during which entry would be required for each room or area (ref. 1).
I[Document NO.]                                             Rev. 0                                      Page 130 of 2401

ATTACHMENT 1 EAL Bases This IC addresses an event involving a release of a hazardous gas that precludes or impedes access to equipment necessary to maintain normal plant operation, or required for a normal plant cooldown and shutdown. This condition represents an actual or potential substantial degradation of the level of safety of the plant.
An Alert declaration is warranted if entry into the affected room/area is, or may be, procedurally required during the plant operating mode in effect at the time of the gaseous release. The emergency classification is not contingent upon whether entry is actually necessary at the time of the release.
Evaluation of the IC and EAL do not require atmospheri* sampling; it only requires the Emergency Coordinator's judgment that the gas concentration in the affected room/area is sufficient to preclude or significantly impede procedurally required access. This judgment may be based on a variety of factors including an existing job hazard analysis, report of ill effects on personnel, advice from a subject matter expert or operating experience with the same or similar hazards. Access should be considered as impeded if extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., requiring use of protective equipment, such as SCBAs, that is not routinely employed).
An emergency declaration is not warranted if any of the following conditions apply:
* The plant is in an operating mode different than the mode specified for the affected room/area (i.e., entry is not required during the operating mode in effect at the time of the gaseous release). For example, the plant is in Mode 1 when the gaseous release occurs, and the procedures used for normal operation, cooldown and shutdown do not require entry into the affected room until Mode 4.
   *The gas release is a planned activity that includes compensatory measures which address the temporary inaccessibility of a room or area (e.g., fire suppression system testing).
   *The action for which room/area entry is required is of an administrative or record keeping nature (e.g., normal rounds or routine inspections).
   *The access control measures are of a conservative or precautionary nature, and would not actually prevent or impede a required action.
   *If the equipment in the listed room or area was already inoperable, or out-of-service, before the event occurred, then no emergency should be declared since the event will have no adverse impact beyond that already allowed by Technical Specifications at the time of the event.
An asphyxiant is a gas capable of reducing the level of oxygen in the body to dangerous levels. Most commonly, asphyxiants work by merely displacing air in an enclosed environment.
This reduces the concentration of oxygen below the normal level of around 19%, which can lead to breathing difficulties, unconsciousness or even death.
This EAL does not apply to firefighting activities that automatically or manually activate a fire suppression system in an area..
Escalation of the emergency classification level would be via Recognition Category R, C or F I~s.
I[Document No.]                                  Rev. 0                           Page 131 of 240I

ATTACHMENT 1 EAL Bases ONS Basis Reference(s):
: 1. Attachment 3 Safe Operation & Shutdown Rooms/Areas Tables R-3 & H-2 Bases
: 2. NE! 99-01 HA5 S[Document No.]                          Rev. 0                       Page 132 of 240J

ATTACHMENT 1 EAL Bases Category:                 H - Hazards and Other Conditions Affecting Plant Safety Subcategory:              6 - Control Room Evacuation' Initiating Condition:     Control Room evacuation resulting in transfer of plant control to alternate locations EAL:
HA6.1            Alert An event has resulted in plant control being transferred from the Control Room to the Auxiliary Shutdown Panel or Standby Shutdown Facility Mode Applicability:
All Definition(s):
None Basis:
The Control Room Supervisor (CRS) determines if the Control Room is uninhabitable and requires evacuation. Control Room inhabitability may be caused by fire, dense smoke, noxious fumes, bomb threat in or adjacent to the Control Room, or other life threatening conditions.
AP/1 ,2,3/A/1700/008, Loss of Control Room, provides the instructions for tripping the unit, and maintaining RCS inventory and Hot Shutdown conditions from outside the Control Room due to reasons other than fire (ref. 1).
AP/1 ,2,3/A/1700/050, Challenging Plant Fire, provides the instructions for tripping the unit, and maintaining RCS inventory and Hot Shutdown conditions from outside the Control Room due to a fire (ref. 2).
If normal post-trip conditions cannot be maintained with the Auxiliary Shutdown Panel or there is a challenging fire in an SSF risk area, plant shutdown may be directed from the Standby Shutdown Facility (ref. 3, 4).
Inability to establish plant control from outside the Control Room escalates this event to a Site Area Emergency per EAL HS6.1.
This IC addresses an evacuation of the Control Room that results in transfer of plant control to alternate locations outside the Control Room. The loss of the ability to control the plant from the Control Room is considered to be a potential substantial degradation in the level of plant safety.
Following a Control Room evacuation, control of the plant will be transferred to alternate shutdown locations. The necessity to control a plant shutdown from outside the Control Room, in addition to responding to the event that required the evacuation of the Control Room, will present challenges to plant operators and other on-shift personnel. Activation of the ERO and emergency response facilities will assist in responding to these challenges.
Escalation of the emergency classification level would be via IC HS6.
I[Document No.]                                  Rev. 0                         Page.133 of 240

ATTACHMENT 1 EAL Bases ONS Basis Reference(s):
: 1. AP/1 ,2,3/A/1 700/008 Loss of Control Room
: 2. AP/1 ,2,3/A/1 700/050 Challenging Plant Fire
: 3. AP/0/A/1700/025 Standby Shutdown Facility Emergency Operating Procedure
: 4. AP/0/A/1700/043 Fire Brigade Response Procedure
: 5. NEl 99-01 HA6
[Document No.]                              Rev. 0                    Page 134 of 240I

ATTACHMENT 1 EAL Bases Category:                   H - Hazards and Other Conditions Affecting Plant Safety Subcategory:                6 - Control Room Evacuation
                                                                                                        /
Initiating Condition:       Inability to control a key safety function from outside the Control Room EAL:
HS6.1            Site Area Emergency An event has resulted in plant control being transferred from the Control Room to the Auxiliary Shutdown Panel or Standby Shutdown Facility AND Control of any of the following key safety functions is not re-established within 15 min.
(Note 1):                                                                                                I
* Reactivity (Modes 1, 2 and 3 only)
* Core cooling
* RCS heat removal Note 1:    The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
Mode Applicability:
1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown, 5 - Cold Shutdown, 6 -
Refuel Definition(s):
None Basis:
The Control Room Supervisor (CRS) determines if the Control Room is uninhabitable and requires evacuation. Control Room inhabitability may be caused by fire, dense smoke, noxious fumes, bomb threat in or adjacent to the Control Room, or other life threatening conditions.
AP/1,2,3/A/1700/008, Loss of Control Room, provides the instructions for tripping the unit, and maintaining RCS inventory and Hot Shutdown conditions from outside the Control Room due to reasons other than fire (ref. 1).
AP/1 ,2,3/A/1700/050, Challenging Plant Fire, provides the instructions for tripping the unit, and maintaining ROS inventory and Hot Shutdown conditions from outside the Control Room due to a fire (ref. 2).
If normal post-trip conditions cannot be maintained with the Auxiliary Shutdown Panel or there is a challenging fire in an SSF risk area, plant shutdown may be directed from the Standby Shutdown Facility (ref. 3, 4).
The intent of this EAL is to capture events in which control of the plant cannot be reestablished in a timely manner. The fifteen minute time for transfer starts when the Control Room is evacuated (when CRS leaves the Control Room, not when AP/1 ,2,3/A/1 700/008 or AP/1 ,2,3/A/1700/050 is entered). The time interval is based on how quickly control must be S[Document No.]                                        Rev. 0                            Page 135 of 240I

ATTACHMENT 1 EAL Bases reestablished without core uncovery and/or core damage. The determination of whether or not control is established from outside the Control Room is based on Emergency Coordinator judgment. The Emergency Coordinator is expected to make a reasonable, informed judgment that control of the plant from outside the Control Room cannot be established within the fifteen minute interval.
Once the Control Room is evacuated, the objective is to establish control of important plant equipment and maintain knowledge of important plant parameters in a timely manner. Primary emphasis should be placed on components and instruments that supply protection for and information about safety functions. Typically, these safety functions are reactivity control (ability to shutdown the reactor and maintain it shutdown), RCS inventory (ability to cool the core), and secondary heat removal (ability to maintain a heat sink).
This IC addresses an evacuation of the Control Room that results in transfer of plant control to alternate locations, and the control of a key safety function cannot be reestablished in a timely manner. The failure to gain control of a key safety function following a transfer of plant control to alternate locations is a precursor to a challenge to one or more fission product barriers within a relatively short period of time.
The determination of whether or not "control" is established at the remote safe shutdown location(s) is based on Emergency Coordinator judgment. The Emergency Coordinator is expected to make a reasonable, informed judgment within 15 minutes whether or not the operating staff has control of key safety functions from the remote safe shutdown location(s).
Escalation of the emergency classification level would be via IC FG1 or CG1 ONS Basis Reference(s):
: 1. AP/1 ,2,3/A/11700/008 Loss of Control Room
: 2. AP/1,2,3/A/1700/050 Challenging Plant Fire
: 3. AP/0/A/1700/025 Standby Shutdown Facility Emergency Operating Procedure
: 4. AP/0/A/1700/043 Fire Brigade Response Procedure
: 5. NEI 99-01 HS6
[Document No.]                                 Rev. 0                            Page 136 of 240J

ATTACHMENT 1 EAL Bases Category:                    H - Hazards and Other conditions Affecting Plant Safety Subcategory:                 7 - Emergency Coordinator Judgment Initiating Condition:        Other conditions existing that' in the judgment of the Emergency Coordinator warrant declaration of a UE EAL:
HU7.1                   Unusual Event Other conditions exist which in the judgment of the Errergency Coordinator 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 uless further degradation of SAFETY SYSTEMS occurs.
Mode Applicability:
All Definition(s):
SAFETY SYSTEM -A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2):
Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:
(1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures..
Basis:
The Emergency Coordinator is the designated onsite individual having the responsibility and authority for implementing the ONS Emergency Plan (ref. 1). The Operations Shift Manager (SM) initially acts in the capacity of the Emergency Coordinator and takes actions as outlined in the Emergency Plan implementing procedures (ref. 2). If required by the emergency classification or if deemed appropriate by the Emergency Coordinator, emergency response personnel are notified and instructed to report to their emergency response locations. In this manner, the individual usually in charge of activities in the Control Room is responsible for initiating the necessary emergency response, but plant management is expected to manage the emergency response as soon as available to do so in anticipation of the possible wide-ranging responsibilities associated with managing a major emergency.
This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Coordinator to fall under the emergency classification level description for an Unusual Event.
S[Document No.]                                  Rev. 0                             Page 137 of 240

ATTACHMENT 1 EAL Bases ONS Basis Reference(s):
: 1. ONS Emergency Plan Section A Assignment of Responsibility
: 2. RP/0/A/1 000/001 Emergency Classification
: 3. NEI 99-01 HU7 Rev.'0 I
[Document No.]                            Rev.' 0                 138 of 240 I '
Page 138

ATTACHMENT 1 EAL Bases Category:                  H - Hazards and Other Conditions Affecting Plant Safety Subcategory:               7 - Emergency Coordinator Judgment Initiating Condition:      Other conditions exist that in the judgment of the Emergency Coordinator warrant declaration of an Alert EAL:
HA7.1           Alert Other conditions dxist which, in the judgment of the Emergency Coordiihator, indicate that events are in progress or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant or a security event that involves probable life threatening risk to site personnel or damage to site equipment because of HOSTILE ACTION. Any releases are expected to be limited to small fractions of the EPA Protective Action Guideline exposure levels.
Mode Applicability:
All Definition(s):
HOSTILE ACTION - An act toward ONS 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 ONS. Non-terrorism-based EALs should be used to address such activities (ie.e, this may include violent acts between individuals in the owner controlled area).
Basis:
The Emergency Coordinator is the designated onsite individual having the responsibility and authority for implementing the ONS Emergency Plan (ref. 1). The Operations Shift Manager (SM) initially acts in the capacity of the Emergency Coordinator and takes actions as outlined in the Emergency Plan implementing procedures (ref. 2). If required by the emergency classification or if deemed appropriate by the Emergency Coordinator, emergency response personnel are notified and instructed to report to their emergency response locations. In this manner, the individual usually in charge of activities in the Control Room is responsible for initiating the necessary emergency response, but Plant Management is expected to manage the emergency response as soon as available to do so in anticipation of the possible wide-ranging responsibilities associated with managing a major emergency.
This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Coordinator to fall under the emergency classification level description for an Alert.
S[Document No.]                                Rev. 0                            Page 139 of 240I

ATTACHMENT 1 EAL Bases ONS Basis Reference(s):
: 1. ONS Emergency Plan Section A Assignment of Responsibility
: 2. RP/0/A/1 000/001 Emergency Classification
: 3. NEI 99-01 HA7 S[Document No.]                           Rev. 0              Page 140 of 240

ATTACHMENT 1 EAL Bases Category:                    H - Hazards and Other Conditions Affecting Plant Safety S ubcategory:               7 - Emergency Coordinator Judgment Initiating Condition:      Other conditions existing that in the judgment of the Emergency Coordinator warrant declaration of a Site Area Emergency EAL:
HS7.1          Site Area Emergency Other conditions exist which in the judgment of the Emergency Coordinator indicate that events are in progress or have occurred which involve actual or likely major failures of plant functions needed for protection of the public or HOSTILE ACTION that results in intentional damage or malicious acts, (1) toward site personnel or equipment that could lead to the likely failure of or, (2) that prevent effective access to equipment needed for the protection of the public. Any releases are not expected to result in exposure levels which exceed EPA Protective Action Guideline exposure levels beyond the SITE BOUNDARY Mode Applicability:
All Definition(s):
HOSTILE ACTION - An act toward ONS 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 ONS. 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).
SITE BOUNDARY- That area, including the PROTECTED AREA, in which Duke Energy has the authority to control all activities including exclusion or removal of personnel and property (1 mile radius) from the center of Unit 2.
Basis:
The Emergency Coordinator is the designated onsite individual having the responsibility and authority for implementing the ONS Emergency Plan (ref. 1). The Operations Shift Manager (SM) initially acts in the capacity of the Emergency Coordinator and takes actions as outlined in the Emergency Plan implementing procedures (ref. 2). If required by the emergency classification or if deemed appropriate by the Emergency Coordinator, emergency response personnel are notified and instructed to report to their emergency response locations. In this manner, the individual usually in charge of activities in the Control Room is responsible for initiating the necessary emergency response, but Plant Management is expected to manage the emergency response as soon as available to do so in anticipation of the possible wide-ranging responsibilities associated with managing a major emergency.
S[Document No.]                           .        Rev. 0                         Page 141 of 240,

ATTACHMENT 1 EAL Bases This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Coordinator to fall under the emergency classification level description for a Site Area Emergency.
ONS Basis Reference(s):
: 1. ONS Emergency Plan Section A Assignment of Responsibility
: 2. RP/0/N/1000/001 Emergency Classification
: 3. NEI 99-01 HS7 S[Document No.]                              Rev.' 0                            Page 142 of 240

ATTACHMENT 1 EAL Bases Category:                 H - Hazards and Other Conditions Affecting Plant Safety Subcategory:              7 - Emergency Coordinator Judgment Initiating Condition:     Other conditions exist which in the judgment of the Emergency Coordinator warrant declaration of a General Emergency EAL:
HG7.1                  General Emergency Other conditions exist which in the judgment* of the Emergency Coordinator indicate that events are in progress or have occurred which involve actual Or IMMINENT substantial core degradation or melting with potential for loss of containment integrity or HOSTILE ACTION that results in an actual loss of physical control of the facility. Releases can be reasonably expected to exceed EPA Protective Action Guideline exposure levels offsite for more than the immediate site area.
Mode Applicability:
All Definition(s):
HOSTILE ACTION - An act toward ONS 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 ONS. 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).
IMMINENT - The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions.
Basis:
The Emergency Coordinator is the designated onsite individual having the responsibility and authority for implementing the ONS Emergency Plan (ref. 1). The Operations Shift Manager (SM) initially acts in the capacity of the Emergency Coordinator and takes actions as outlined in the Emergency Plan implementing procedures (ref. 2). If required by the emergency classification or if deemed appropriate by the Emergency Coordinator, emergency response personnel are notified and instructed to report to their emergency response locations. In this manner, the individual usually in charge of activities in the Control Room is responsible for initiating the necessary emergency response, but Plant Management is expected to manage the emergency response as soon as available to do so in anticipation of the possible wide-ranging responsibilities associated with managing a major emergency.
Releases can reasonably be expected to exceed EPA PAG plume exposure levels outside the Site Boundary.
S[Document No.]                                Rev.O0                            Page143 of 240

ATTACHMENT 1 EAL Bases This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Coordinator to fall under the emergency classification level description for a General Emergency.
ONS Basis Reference(s):
: 1. ONS Emergency Plan Section A Assignment of Responsibility
: 2. RP/0/A/1 000/001 Emergency Classification
: 13. NEI 99-01 HG7I I[Document No.]                               Rev. 0                             Page 144 of 240

ATTACHMENT 1 EAL Bases Category S     -  System Malfunction EAL Group: Hot Conditions (RCS temperature       > 210&deg;F); EALs in this category are applicable only in one or more hot operating modes.
Numerous system-related equipment failure events that warrant emergency classification have been identified in this category. They may pose actual or potential threats to plant safety.
The events of this category pertain to the following subcategories:
: 1. Ldss of Essential AC PowerI Loss of essential electrical power can compromise plant safety system operability including decay heat removal and emergency core cooling systems which may be necessary to ensure fission product barrier integrity. This category includes loss of onsite and offsite sources for 4160V AC essential buses.
: 2. Loss of Vital DC Power Loss of emergency electrical power can compromise plant safety system operability including decay heat removal and emergency core cooling systems which may be necessary to ensure fission product barrier integrity. This category includes loss of vital plant 125V DC power sources.
: 3. Loss of Control Room Indications Certain events that degrade plant operator ability to effectively assess plant conditions within the plant warrant emergency classification. Losses of indicators are in this subcategory.
: 4. RCS Activity During normal operation, reactor coolant fission product activity is very low. Small concentrations of fission products in the coolant are primarily from the fission of tramp uranium in the fuel clad or minor perforations in the clad itself. Any significant increase from these base-line levels (2% - 5% clad failures) is indicative of fuel failures and is covered under the Fission Product Barrier Degradation category. However, lesser amounts of clad damage may result in coolant activity exceeding Technical Specification limits.. These fission products will be circulated with the reactor coolant and can be detected by coolant sampling.
: 5. RCS Leakage The reactor vessel provides a volume for the coolant that covers the reactor core. The reactor pressure vessel and associated pressure piping (reactor coolant system) together provide a barrier to limit the release of radioactive material should the reactor fuel clad integrity fail. Excessive RCS leakage greater than Technical Specification limits indicates potential pipe cracks that may propagate to an extent threatening fuel clad, RCS and containment integrity.
: 6. RPS Failure This subcategory includes events related to failure of the Reactor Protective System (RPS) to initiate and complete reactor trips. In the plant licensing basis, postulated failures of the RPS to complete a reactor trip comprise a specific set of analyzed events referred to as I[Document No.]
* Rev. 0                     .      Page 145 of 240I

ATTACHMENT 1 EAL Bases Anticipated Transient Without Scram (ATWS) events. For EAL classification, however, ATWS is intended to mean any trip failure event that does not achieve reactor shutdown. If RPS actuation fails to assure reactor shutdown, positive control of reactivity is at risk and could cause a threat to fuel clad, RCS and containment integrity.
: 7. Loss of Communications Certain events that degrade plant operator ability to effectively communicate with essential personnel within or external to the plant warrant emergency classification.
: 8. Containment Failure Failure of containment isolation capability (under conditions in which the containment is not currently challenged) warrants emergency classification. Failure of containment pressure control capability also warrants emergency classification.
: 9. Hazardous Event Affectina Safety Systems Various natural and technological events that result in degraded plant safety system performance or significant visible damage warrant emergency classification under this subcategory.
J[Document No.]                      J         Rev. 0                          Page 146 of 240

ATTACHMENT 1 EAL Bases Category:                  S   -   System Malfunction Subcategory:                1 - Loss of Essential AC Power Initiating Condition:     Loss of all offsite AC power capability to essential buses for 15 minutes or longer EAL:
SUI.1           Unusual Event Loss of all offsite AC povw'er capability, Table S-i, to essential 4160 V buses Mv~FB-1 and MFB-2 for > 15 mai. (Note 1)
Note 1:   The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or sWill likely be exceeded.
Table S-I     AC Power Sources Offsite:
* Unit Normal Transformer (backcharged)
* Unit Startup Transformer (SWYD)
* Another Unit Startup Transformer (aligned) (SWYD)
* CT5 (Central/energizing Standby Bus)
Emergency:
* Unit Startup Transformer (Keowee)
* Another Unit Startup Transformer (aligned) (Keowee)
* CT4
* CT5 (dedicated line/energizing Standby Bus)
Mode Applicability:
1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):
None Basis:
The 4160 V AC System provides the power requirements for operation and safe shutdown of the plant. 4160 V buses MFB-1 and MFB-2 are the essential buses (ref. 1).
The condition indicated by this EAL is the degradation of all offsite AC power sources such that only onsite AC power capability exists for 1,5 minutes or longer.
Each unit is provided with two physically independent circuits from the switching station. One is the circuit from the 230 kV switching station through the startup transformer, which is designed to be available within a few seconds following a loss of coolant accident. The second circuit is the path from the switchyard through the main step-up transformer, the main generator bus and the unit auxiliary transformer with the generator disconnected from the main bus. The second circuit is currently used during refueling as an additional power feed for the shutdown II[Document    No.]                                     Rev. 0                           Page 147 .of 240I

ATTACHMENT 1 EAL Bases unit(s) from the 230 kV switchyard. Whenever there is inadequate power from the generating units, the 230 kV switching station and the hydro units, power is available to the standby power buses either directly from the 100 kV Central Tie Substation or from Lee Steam Station via Transformer CT5. (ref. 2)
Upon loss of power from the Oconee generating unit and 230 kV switchyard, power is supplied from both Keowee Hydro Station units through two separate and independent routes. One route is an underground feeder to Transformer CT4 which supplies the two redundant Main Feeder Buses (MFB-1 and MFB-2). The other route is an overhead feeder to the 230 KV switching station which supplies each unit startup transformer. (ref. 3)
The Standby Shutdown Facility (SSF) consists of standby systems for use in an extreme emergency and is equipped with a manually started diesel generator that can supply power necessary to maintain hot shutdown of the reactors of each unit in the event of loss of ,power.
However, the SSF cannot supply'power to the essential buses and therefore not credited in this EAL (ref. 3).
The 15-minute interval was selected as a threshold to exclude transient or momentary power losses.
This IC addresses a prolonged loss of offsite power. The loss of offsite power sources renders the plant more vulnerable to a complete loss of power to AC essential buses. This condition represents a potential reduction in the level of safety of the plant.
For emergency classification purposes, "capability" means that an offsite AC power source(s) is available to the essential buses, whether or not the buses are powered from it.
Escalation of the emergency classification level would be via IC SAI.
ONS Basis Reference(s):
: 1. UFSAR Figure 8.1 Single Line Diagram
: 2. UFSAR Section 8.20Offsite Power System
: 3. UFSAR Section 8.30Onsite Power Systems
: 4. NEI 99-01 SU1I'
[Document No.]                                Rev. 0                           Page 148 of 240

ATTACHMENT 1 EAL Bases Category:                   S  -  System Malfunction Subcategory:                1 - Loss of Essential AC Power Initiating Condition:       Loss of all but one AC power source to essential buses for 15 minutes or longer EAL:'
SAI.1           Alert AC power capability, Table S-1, to essential 4160 V buses MFB-1 and MFB-2 reduced to a single power source for &#x17d;- 15 min. (Note 1)
AND Any additional single power source failure will result in loss of all AC power to SAFETY SYSTEMS Note 1:    The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
Table S-I     AC Power Sources Offsite:
* Unit Normal Transformer (backcharged)
* Unit Startup Transformer (SWYD)
* Another Unit Startup Transformer (aligned) (SWYD)
* CT5 (Central/energizing Standby Bus)
Emergency:
* Unit Startup Transformer (Keowee)
* Another Unit Startup Transformer (aligned) (Keowee)
* CT4
* CT5 (dedicated line/energizing Standby Bus)
Mode Applicability:
1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4         - Hot Shutdown Definition(s):
SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10OCFR50.2):
Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:
(1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; Rv                                Pg       49o        *4
[DcmntN.
*I[Document No.]                                        Rev. 0 Page 149 of 240 I

ATTACHMENT 1 EAL Bases (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.
Basis:
For emergency classification purposes, "capability" means that an AC power source is available to and capable of powering the emergency bus(es) within 15 min, whether or not the buses are currently powered from it.
The 4160 V AC System provides the power requirements for operation and safe shutdown of the plant. 4160 V buses MFB-1 and MFB-* are the essential buses (ref. 1).
The condition indicated by this EAL is the degradation of the offsite and onsite power sources such that any additional single failure would result in a loss of all AC power to the essential buses.
Each unit is provided with two physically independent circuits from the switching station. One is the circuit from the 230 kV switching station through the startup transformer, which is designed to be available within a few seconds following a loss of coolant accident. The second circuit is the path from the switchyard through the main step-up transformer, the main generator bus and the unit auxiliary transformer with the generator disconnected from the main bus. The second circuit is currently used during refueling as an additional power feed for the shutdown unit(s) from the 230 kV switchyard. Whenever there is inadequate power from the generating units, the 230 kV switching station and the hydro units, power is available to the standby power buses either directly from the 100 kV Central Tie Substation or from Lee Steam Station via a Transformer CT5. (ref. 2)
Upon loss of power from the Oconee generating unit and 230 kV switchyard, power is supplied from both Keowee Hydro Station units through two separate and independent routes. One route is an underground feeder to Transformer CT4 which supplies the two redundant Main Feeder Buses (MFB-1 and MFB-2). The other route is an overhead feeder to the 230 KV switching station which supplies each unit startup transformer. (ref. 3)
The Standby Shutdown Facility (SSF) consists of standby systems for use in an extreme emergency and is equipped with a manually started diesel generator that can supply power necessary to maintain hot shutdown of the reactors of each unit in the event of loss of power.
However, the SSF cannot supply power to the essential buses and therefore not credited in this EAL (ref. 3).
The 15-minute interval was selected as a threshold to exclude transient or momentary power losses. If the capability of a second source of emergency bus power is not restored within 15 minutes, an Alert is declared under this EAL.
This IC describes a significant degradation of offsite and onsite AC power sources such that any additional single failure would result in a loss of all AC power to SAFETY SYSTEMS. In this condition, the sole AC power source may be powering one, or more than one, train of safety-related equipment. This IC provides an escalation path from IC SUI.
An "AC power source" is a source recognized in APs and EOPs, and capable of supplying required power to an essential bus. Some examples of this condition are presented below.
* A loss of all offsite power with a concurrent failure of all but one essential power source (e.g., CT4, CT5, CT1, 2 ,3 (Keowee)).
((Document No.]                          [        Rev. 0         [Page                    150 of 240I

ATTACHMENT 1 EAL Bases
       *A loss of essential power sources (e.g., CT4, CT5, CT1, CT2, CT3 (Keowee)) with a single train of essential buses being back-fed from an offsite power source.
Escalation of the emergency classification level would be via IC SSI.
ONS Basis Reference(s):
: 1. UFSAR Figure 8.1 Single Line Diagram
: 2. UFSAR Section 8.2 Offsite Power System
: 3. UFSAR Section 8.30Onsite Power Systems I4. NEI 99-01 SA1 I[Document No.]                       IRev.         0                          Page 151 of 240

ATTACHMENT 1 EAL Bases Category:                  S - System Malfunction Subcategory:               1 - Loss of Essential AC Power Initiating Condition:      Loss of all offsite power and all emergency AC power to essential buses for 15 minutes or longer EAL:
SS1.1           Site Area Emergency Loss o*f all offsite and all emergency AC power capability td essential 4160 V buses MFB-1 and MFB-2 for > 15 min. (Note 1)
Note 1:   The Emergency Coordinator should declare the event promptly upon determining that time limit has I been exceeded, or will likely be exceeded.
Mode Applicability:
1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):
None Basis:
For emergency classification purposes, "capability" means that an AC power source is available to and capable of powering the emergency bus(es) within 15 minutes, whether or not the buses are currently powered from it.
The condition indicated by this EAL is the degradation of the offsite and emergency power sources resulting in a loss of all AC power to the emergency buses. 4160 V buses MFB-1 and MFB-2 are the essential buses (ref. 1).
Each unit is provided with two physically independent circuits from the switching station. One is the circuit from the 230 kV switching station through the startup transformer, which is designed to be available within a few seconds following a loss of coolant accident. The second circuit is the path from the switchyard through the main step-up transformer, the main generator bus and the unit auxiliary transformer with the generator disconnected from the main bus. The second circuit is currently used during refueling as an additional power feed for the shutdown unit(s) from the 230 kV switchyard. Whenever there is inadequate power from the generating units, the 230 kV switching station and the hydro units, power is available to the standby power buses either directly from the 100 kV Central Tie Substation or from Lee Steam Station via a Transformer CT5. (ref. 2)
Upon loss of power from the Oconee generating unit and 230 kV switchyard, power is supplied from both Keowee Hydro Station units through two separate and independent routes. One route is an underground feeder to Transformer CT4 which supplies the two redundant Main Feeder Buses (MFB-1 and MFB-2). The other route is an overhead feeder to the 230 KV switching station which supplies each unit startup transformer. (ref. 3)
The Standby Shutdown Facility (SSF) consists of standby systems for use in an extreme emergency and is equipped with a manually started diesel generator that can supply power necessary to maintain hot shutdown of the reactors of each unit in the event of loss of power.
mI[Document   No.]                                   Rev. 0                            Page 152 of 2401

ATTACHMENT 1 EAL Bases However, the SSF cannot supply power to the essential buses and therefore not credited in this EAL. (ref. 3).
The 15-minute interval was selected as a threshold to exclude transient or momentary power losses. The interval begins when both offsite and onsite AC power capability are lost.
This IC addresses a total loss of AC power that compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink.
In addition, fission product barrier monitoring capabilities may be degraded under these conditions. Thfs IC represents a condition that involves actual or"likely major failures of plant functions needed for the protection of the public.
Escalation of the emergency classification level would be via ICs RG1, FG1 or SGI.
ONS Basis Reference(s):
: 1. UFSAR Figure 8.1 Single Line Diagram
: 2. UFSAR Section 8.20Offsite Power System
: 3. UFSAR Section 8.30Onsite Power Systems
: 4. NEI 99-01 SS1 S[Document No.]                                Rev. 0                          page 153 of 240I

ATTACHMENT 1 EAL Bases Category:                  S -System Malfunction S ubcategory:               1 - Loss of Essential AC Power Initiating Condition:       Prolonged loss of all offsite and all emergency AC power to essential buses EAL:
SGI.1           General Emergency Loss of all offsite and all emergency AC power capability to essential 4160 \* buses MEB-1 and MFB-2 AND Failure to power 5SF equipment and PSW unavailable AND EITHER:
* Restoration of at least one essential bus in < 4 hour is not likely (Note 1)
* CETC reading > 1200&deg;F Note 1:   The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
Mode Applicability:
1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):
None Basis:
For emergency classification purposes, "capability" means that an AC power source is available to and capable of powering the essential bus(es), whether or not the buses are currently powered from it.~4160 V buses MFB-1 and MFB-2 are the essential buses (ref. 1).
Each unit is provided with two physically independent circuits from the switching station. One is the circuit from the 230 kV switching station through the startup transformer, which is designed to be available within a few seconds following a loss of coolant accident. The second circuit is the path from the switchyard through the main step-up transformer, the main generator bus and the unit auxiliary transformer with the generator disconnected from the main bus. The second circuit is currently used during refueling as an additional power feed for the shutdown unit(s) from the 230 kV switchyard. Whenever there is inadequate power from the generating units, the 230 kV switching station and the hydro units, power is available to the standby power buses either directly from the 100 kV Central Tie Substation or from Lee Steam Station via Transformer CT5. (ref. 2)
Upon loss of power from the Oconee generating unit and 230 kV switchyard, power is supplied from both Keowee Hydro Station units through two separate and independent routes. One route is an underground feeder to Transformer CT4 which supplies the two redundant Main S[Document No.]                                       Rev. 0                            Page 154 of 240

ATTACHMENT 1 EAL Bases Feeder Buses (MFB-1 and MFB-2). The other route is an overhead feeder to the 230 KV switching station which supplies each unit startup transformer. (ref. 3)
The Standby Shutdown Facility (SSF) consists of standby systems for use in an extreme emergency and is equipped with a manually started diesel generator and Protected Service Water (PSW) power supply that can supply power necessary to maintain hot shutdown of the reactors of each unit in the event of loss of power. Although the SSF requires manual initiation, it is considered in this EAL because it may be capable of powering the SSF load center (ref. 3, 4).
The station blackout coping peribd is four hours (ref. 5).
Core Exit Thermocouple readings of 1200&deg;F are indicative of superheat conditions and inability to adequately remove heat from the core (ref. 6).-
This IC addresses a prolonged toss of all power sources to AC essential buses. A Io'ss of all AC power compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink. A prolonged loss of these buses will lead to a loss of one or more fission product barriers. In addition, fission product barrier monitoring capabilities may be degraded under these conditions.
The EAL should require declaration of a General Emergency prior to meeting the thresholds for IC FGI. This will allow additional time for implementation of offsite protective actions.
Escalation of the emergency classification from Site Area Emergency will occur if it is projected that power cannot be restored to at least one AC essential bus by the end of the analyzed station blackout coping period. Beyond this time, plant responses and event trajectory are subject to greater uncertainty, and there is an increased likelihood of challenges to multiple fission product barriers.
The estimate for restoring at least one essential bus should be based on a realistic appraisal of the situation. Mitigation actions with a low probability of success should not be used as a basis for delaying a classification upgrade. The goal is to maximize the time available to prepare for, and implement, protective actions for the public.
The EAL will also require a General Emergency declaration if the loss of AC power results in parameters that indicate an inability to adequately remove decay heat from the core.
ONS Basis Reference(s):
: 1. UFSAR Figure 8.1 Single Line Diagram
: 2. UFSAR Section 8.20Offsite Power System
: 3. UFSAR Section 8.3 Onsite Power Systems
: 4. UFSAR Section 9.6 Standby Shutdown Facility
: 5. UFSAR Section 8.3.2.2.4 Station Blackout Analysis
: 6. RP/O/A/1000/18 Core Damage Assessment
: 7. NEI 99-01 SGI I[Document No.]                                 Rev. 0                             Page 155 of 240

ATTACHMENT 1 EAL Bases Category:                  S -System Malfunction Subcategory:               1 - Loss of Essential AC Power Initiating Condition:      Loss of all essential AC and vital DC power sources for 15 minutes or longer EAL:
SG1.2           General Emergency Loss of all offsite and all emergency ACipower capability to essential 4160 V buses MEB-1 and MFB-2 for > 15 min.
AND Failure to power SSF equipment and PSW unavailable AND Loss of 125 VDC power based on battery bus voltage indications      <  105 VDC on both vital DC Distribution Centers DCA and DCB for > 15 min. (Note 1) been exceeded, or will likely be exceeded.
Mode Applicability:
1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):
None Basis:
This EAL is indicated by the loss of all offsite and emergency AC power capability to 4160 V essential buses MFB-1 and MFB-2 for greater than 15 minutes in combination with degraded vital DC power voltage. This EAL addresses operating experience from the March 2011 accident at Fukushima Daiichi.
For emergency classification purposes, "capability" means that an AC power source is available to and capable of powering the emergency bus(es) within 15 minutes, whether or not the buses are currently powered from it.
Each unit is provided with two physically independent circuits from the switching station. One is the circuit from the 230 kV switching station through the startup transformer, which is designed to be available within a few seconds following a loss of coolant accident. The second circuit is the path from the switchyard through the main step-up transformer, the main generator bus and the unit auxiliary transformer with the generator disconnected from the main bus. The second circuit is currently used during refueling as an additional power feed for the shutdown unit(s) from the 230 kV switchyard. Whenever there is inadequate power from the generating units, the 230 kV switching station and the hydro units, power is available to the standby power buses either directly from the 100 kV Central Tie Substation or from Lee Steam Station via Transformer CT5. (ref. 2) e..                        ae   5    f20
[DcmnIo]
I [Document No.]                                       Rev. 0 Page 156 of 240 I

ATTACHMENT 1 EAL Bases Upon loss of power from the Oconee generating unit and 230 kV switchyard, power is supplied from both Keowee Hydro Station units through two separate and independent routes. One route is an underground feeder to Transformer CT4 which supplies the two redundant Main Feeder Buses (MFB-1 and MFB-2). The other route is an overhead feeder to the 230 KV switching station which supplies each unit startup transformer. (ref. 3)
The Standby Shutdown Facility (SSF) consists of standby systems for use in an extreme emergency and is equipped with a manually started diesel generator and Protected Service Water (PSW) power supply that can supply power necessary to maintain hot shutdown of the reactors of each unit in the event of loss of powe~r. Although the SSF requires manual initiation, it is considered in this EAL because it may be capable of powering the SSF load center (ref. 3).
For each unit, two independent and physically separated 125 volt DC batteries and DC buses are provided for the vital instrumentation and control power system. (ref. 4, 5). Minimum DC bus voltage is 105 VDC (ref. 6).
This IC addresses a concurrent and prolonged loss of both emergency AC and Vital DC power. A loss of all emergency AC power compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink.
A loss of vital DC power compromises the ability to monitor and control SAFETY SYSTEMS.
A sustained loss of both emergency AC and vital DC power will lead to multiple challenges to fission product barriers.
Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.
The 15-minute emergency declaration clock begins at the point when both EAL thresholds are met.
ONS Basis Reference(s):
: 1. UFSAR Figure 8.1 Single Line Diagram
: 2. UFSAR Section 8.20Offsite Power System
: 3. UFSAR Section 8.30Onsite Power Systems
: 4. UFSAR Figure 8.5 Typical DC and AC Vital Power System - Single Line
: 5. UFSAR Section 8.3.2 DC Power Systems
: 6. EP/*/A/1800/001 Blackout Tab
: 7. NEI 99-01 SG8 S[Document No.]                                Rev. 0                          Page 157 of 240

ATTACHMENT 1 EAL Bases Category:                   S - System Malfunction Subcategory:                2 - Loss of Vital DC Power Initiating Condition:      Loss of all vital DC power for 15 minutes or longer EAL:
Note 1:  The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
Mode Applicability:I 1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):
None Basis:
For each unit, two independent and physically separated 125 volt DC batteries and DC buses are provided for the vital instrumentation and control power system. (ref. 1, 2). Minimum DC bus voltage is 105 VDC (ref. 3).
This IC addresses a loss of vi~tal DC power which compromises the ability to monitor and control SAFETY SYSTEMS. In modes above Cold Shutdown, this condition involves a major failure of plant functions needed for the protection of the public.
Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.
Escalation of the emergency classification level would be via ICs RG1, FG1 or SGI.
ONS Basis Reference(s):
: 1. UFSAR Figure 8.5 Typical DC and AC Vital Power System - Single Line
: 2. UFSAR Section 8.3.2.DC Power Systems
: 3. PI*/IAI1 800/00 1 Blackout Tab
: 4. Technical Specifications 3.8.3 DC Sources             - Operating
: 5. NEI 99-01 SS8 S[Document No.]                                       Rev. 0                           Page 158 of 240

ATTACHMENT 1 EAL Bases Category:                   S - System Malfunction Subcategory:                3 - Loss of Control Room Indications Initiating Condition:        UNPLANNED loss of Control Room indications for 15 minutes or longer EAL:
SU3.1            Unusual Event An UNPLANNED event results in the inability to monitor one or mbre Table S-2 parameters from within the Control Room for --15 min. (Note 1)
Note 1:    The Emergency Coordinator should declare the event promptly upon determining that time limit has been! exceeded, or will likely be exceeded.
Table S-2       Safety System Parameters
* Reactor power
* RCS level
* RCS pressure
* CETC temperature
* Level in at least one S/G
* EFW flow to at least one SIG Mode Applicability:
1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):
UNPLANNED - A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.
Basis:
SAFETY SYSTEM parameters listed in Table S-2 are monitored in the Control Room through a combination of hard control panel indicators as well as computer based information systems.
The SPDS serves as a redundant compensatory indicator which may be utilized in lieu of normal Control Room indicators (ref. 1).
This IC addresses the difficulty associated with monitoring normal plant conditions without the ability to obtain SAFETY SYSTEM parameters from within the Control Room. This condition is a precursor to a more significant event and represents a potential degradation in the level of safety of the plant.
As used in this EAL, an "inability to monitor" means that values for one or more of the listed parameters cannot be determined from within the Control Room. This situation would require a loss of all of the Control Room sources for the given parameter(s). For example, the reactor power level cannot be determined from any analog, digital and recorder source within the Control Room.
S[Document No.]                                         Rev. 0                           Page 159 of 240

ATTACHMENT 1 EAL Bases An event involving a loss of plant indications, annunciators and/or display systems is evaluated in accordance with 10 CER 50.72 (and associated guidance in NUREG-1022) to determine if an NRC event report is required. The event would be reported if it significantly impaired the capability to perform emergency assessments. In particular, emergency assessments necessary to implement abnormal operating procedures, emergency operating procedures, and emergency plan implementing procedures addressing emergency classification, accident assessment, or protective action decision-making.
This EAL is focused on a selected subset of plant parameters associated with the key safety functions of reactivity control, core cooling and RCS heat removal.i The loss of the ability to determine one or more of these parameters from within the Control Room is considered to be more significant than simply a reportable condition. In addition, if all indication sources for one or more of the listed parameters are lost, then the ability to determine the values of other SAFETY SYSTEM parameters may be impacted as well. For example, if the value for reactor vessel level cannot be determined from the indications and recorders on a main control board, the SPDS or the plant computer, the availability of other parameter values may be" compromised as well.
Fifteen minutes was selected as a threshold to exclude transient or momentary losses of indication.
Escalation of the emergency classification level would be via IC SA3.
ONS Basis Reference(s):
: 1. UFSAR Section 7.5 Display Instrumentation
: 2. NEI 99-01 SU2
[Document No.]                                 Rev. 0                             Page 160 of 240

ATTACHMENT 1 EAL Bases Category:                S - System Malfunction Subcategory:              3 - Loss of Control Room Indications Initiating Condition:     UNPLANNED loss of Control Room indications for 15 minutes or longer with a significant transient in progress EAL:
SA3.1          Alert An UNPLANNED eveht results in the inability to monitor one or more Tablb S-2 parameters from within the Control Room for __15 mai. (Note 1)
AND Any significant transient is in progress, Table S-3 Note 1:  The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
Table S-2        Safety System Parameters
* Reactor power
* RCS level
* RCS pressure
* CETC temperature
* Level in at least one S/G
* EFW flow to at least one S/G Table S-3      Significant Transients
* Reactor trip
* Runback > 25% thermal power
* Electrical load rejection   > 25%
electrical load
                                     *ECCS actuation Mode Applicability:
I - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):
UNPLANNED - A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.
S[Document No.]                                      Rev. 0                            Page 1.61 of 240

ATTACHMENT 1 EAL Bases Basis:
SAFETY SYSTEM parameters listed in Table S-2 are monitored in the Control Room through a combination of hard control panel indicators as well as computer based information systems.
The SPDS serves as a redundant compensatory indicator which may be utilized in lieu of normal Control Room indicators (ref. 1).
Significant transients are listed in Table S-3 and include response to automatic or manually initiated functions such as reactor trips, runbacks involving greater than 25% thermal power change, electrical load rejections of greater than 25% full electrical load, reactor power cutbacks or ECCS (SI) injectio'n actuations.
This IC addresses the difficulty associated with monitoring rapidly changing plant conditions during a transient without the ability to obtain SAFETY SYSTEM parameters from within the Control Room. During this condition, the margin to a potential fission product barrier challenge is reduced. It thus represents a potential substantial degradation in the level of safety of the plant.
As used in this EAL, an "inability to monitor" means that values for one or more of the listed parameters cannot be determined from within the Control Room. This situation would require a loss of all of the Control Room sources for the given parameter(s). For example, the reactor power level cannot be determined from any analog, digital and recorder source within the Control Room.
An event involving a loss of plant indications, annunciators and/or display systems is evaluated in accordance with 10 CFR 50.72 (and associated guidance in NUREG-1 022) to determine if an NRC event report is required. The event would be reported if it significantly impaired the capability to perform emergency assessments. In particular, emergency assessments necessary to implement abnormal operating procedures, emergency operating procedures, and emergency plan implementing procedures addressing emergency classification, accident assessment, or protective action decision-making.
This EAL is focused on a selected subset of plant parameters associated with the key safety functions of reactivity control, core cooling and ROS heat removal. The loss of the ability to determine one or more of these parameters from within the Control Room is considered to be more significant than simply a reportable condition. In addition, if all indication sources for one or more of the listed parameters are lost, then the ability to determine the values of other SAFETY SYSTEM parameters may be impacted as well. For example, if the value for reactor vessel level cannot be determined from the indications and recorders on a main control board, the SPDS or the plant computer, the availability of other parameter values may be compromised as well.
Fifteen minutes was selected as a threshold to exclude transient or momentary losses of indication.
Escalation of the emergency classification level would be via ICs FS1 or IC RS1 ONS Basis Reference(s):
: 1. UFSAR Section 7.5 Display Instrumentation
: 2. NEI 99-01 SA2 S[Document No.]                                  Rev. 0                            Page 162 of 24.0

ATTACHMENT 1 EAL Bases Category:                S - System Malfunction Subcategory:             4 - RCS Activity Initiating Condition:    Reactor coolant activity greater than Technical Specification allowable limits EAL:
SU4.1         Unusual Event ROS activity > 50 pCi/gm Dose Equivalent 1-131 for > 48 hr continuous period OR RCS activity > 280 pCi/gm Dose Equivalent Xe-I133 for > 48 hr continuous period Mode Applicability:
I - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):
None Basis:
The specific iodine activity is limited to -<50 pCi/gm Dose Equivalent 1-131 for > 48 hr continuous period. The specific Xe-I133 activity is limited to -<280 pCi/gm Dose Equivalent Xe-133 for > 48 hr continuous period. Entry into Condition C of LCO 3.4.11 meets the intent of this EAL (ref 1).
This IC addresses a reactor coolant activity value that exceeds an allowable limit specified in Technical Specifications. This condition is a precursor to a more significant event and represents a potential degradation of the level of safety of the plant.
Escalation of the emergency classification level would be via ICs FA1 or the Recognition Category R ICs.
ONS Basis Reference(s):
: 1. ONS Technical Specifications LCO 3.4.11 RCS Specific. Activity
: 2. NEI 99-01 SU3 r[Document No.]                                  Rev. 0                         Page  163 of 2401

ATTACHMENT 1 EAL Bases Category:                  S - System Malfunction Suboategory:               5 - RCS Leakage Initiating Condition:      RCS leakage for 15 minutes or longer EAL:
SU5.1          Unusual Event RCS unidentified or pressure boundary leakage > 10 gpm for >- 15 mai.
OR RCS identified leakage >25 gpm for > 15 mai.
OR Leakage from the RCS to a location outside containment > 25 gpm for > 15 min.
(Note 1)
Note 1:   The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
Mode Applicability:
1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):
None Basis:
Manual or computer-based methods of performing an RCS inventory balance are normally used to determine RCS leakage (ref. 1).
Identified leakage includes (ref. 2):
* Leakage such as that from pump seals or valve packing (except reactor coolant pump (RCP) seal water injection or leakoff), that is captured and conducted to collection systems or a sump or collecting tank, or
* Leakage into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be pressure boundary leakage, or
* RCS leakage through a steam generator to the secondary system.
Unidentified leakage is all leakage (except RCP seal water injection or leakoff) that is not identified leakage (ref. 2).
Pressure Boundary leakage is leakage (except SG leakage) through a nonisolable fault in an RCS component body, pipe wall, or vessel wall (ref. 2).
Reactor coolant leakage outside of the containment that is not considered identified or unidentified leakage per Technical Specifications includes leakage via interfacing systems.
Escalation of this EAL to the Alert level is via Category F, Fission Product Barrier Degradation, EAL FAI.1.
S[Document No.]                                       Rev. 0                            Page 164 of 2401

ATTACHMENT I EAL Bases This IC addresses RCS leakage which may be a precursor to a more significant event. In this case, RCS leakage has been detected and operators, following applicable procedures, have been unable to promptly isolate the leak. This condition is considered to be a potential degradation of the level of safety of the plant.
The first and second EAL conditions are focused on a loss of mass from the RCS due to "unidentified leakage", "pressure boundary leakage" or "identified leakage" (as these leakage types are defined in the plant Technical Specifications). The third condition addresses an RCS mass loss caused by an UNISOLABLE leak through an interfacing system. These conditions thu~s apply to leakage into the containment, a secondary-side system (e.g., steam generator tube leakage) or a location outside of containment.
The leak rate values for each condition were selected because they are usually observable with normal Control Room indications. Lesser values ty~pically require time-consuming calculations to determine (e.g., a mass balance calculation). The first condition uses a lower value that reflects the greater significance of unidentified or pressure boundary leakage.
The release of mass from the RCS due to the as-designed/expected operation of a relief valve does not warrant an emergency classification. An emergency classification would be required if a mass loss is caused by a relief valve that is not functioning as designed/expected (e.g., a relief valve sticks open and the line flow cannot be isolated).
The 15-minute threshold duration allows sufficient time for prompt operator actions to isolate the leakage, if possible.                '
Escalation of the emergency classification level would be via ICs of Recognition Category R or F.
ONS Basis Reference(s):
: 1. PT/I1,2,3/A/0600/01 0 Reactor Coolant Leakage
: 2. ONS Technical Specifications Section 1.1 Definitions
: 2. NEI 99-01 SU4 S[Document No.]                                 Rev. 0                           Page 165 of 240

ATTACHMENT 1 EAL Bases Category:                     S - System Malfunction Subcategory:                  6 - RPS Failure Initiating Condition:         Automatic or manual trip fails to shut down the reactor EAL:
SU6.1            Unusual Event An automatic trip did not shut down the reactor as indicated by reactor power > 5% after any RPS setpoint is exceeded AND A subsequent automatic trip or the manual trip pushbutton is successful in shutting down the reactor as indicated by reactor power          < 5% (Note 8)
Note 8:   A manual trip action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron injection strategies.
Mode Applicability:
1 - Power Operation Definition(s):
None Basis:
The first condition of this EAL identifies the need to cease critical reactor operations by actuation of the automatic Reactor Protective System (RPS) trip function. A reactor trip is automatically initiated by the RPS when certain continuously monitored parameters exceed predetermined setpoints (ref. 1).
Following a successful reactor trip, rapid insertion of the control rods occurs. Nuclear power promptly drops to a fraction of the original power level and then decays to a level several decades less with a negative startup rate. The reactor power drop continues until reactor power reaches the point at which the influence of source neutrons on reactor power starts to be observable. A predictable post-trip response from an automatic reactor trip signal should therefore consist of a prompt drop in reactor power as sensed by the nuclear instrumentation and a lowering of power into the source range. For the purpose of emergency classification a successful trip has occurred when there is sufficient rod insertion from the trip of RPS to bring the reactor power below the Power Operation Mode threshold of 5% (ref. 2).
5% rated power is the Power Operation mode threshold. Below 5%, plant response will be similar to that observed during a normal shutdown. Nuclear instrumentation can be used to determine if reactor power is greater than or equal to 5 % power (ref. 1, 2).
For the purposes of emergency classification, a successful manual trip action is that which can be quickly performed from the reactor control console (actuation of the manual trip pushbutton). There is a separate set of switch contacts in series with the output of each reactor trip component. All switch contacts are actuated through a mechanical linkage from a single pushbutton. Reactor shutdown achieved by use of other trip actions such as opening supply S[Document No.]                                       Rev. 0                                 Page 166 of 240I

ATTACHMENT 1 EAL Bases breakers, emergency boration, or manually driving control rods) do not constitute a successful manual trip (ref. 3).
Following any automatic RPS trip signal, insertion of redundant manual trip signals are performed to back up the automatic RPS trip function and ensure reactor shutdown is achieved. Even if the first subsequent manual trip signal inserts all control rods to the full-in position immediately after the initial failure of the automatic trip, the lowest level of classification that must be declared is an Unusual Event.
In the event that the operator identifies a reactor trip is imminent and initiates a successful manual reactor trip Ibefore the automatic RPS trip setpoint is reached, noldeclaration is
*required. The successful manual trip of the reactor before it reaches its automatic trip setpoint or reactor trip signals caused by instrumentation channel failures do not lead to a potential fission product barrier loss. However, if subsequent manual reactor trip actions fail to reduce reactor power belov* 5%, the event escalates to the Alert under EAL SA6.1.
If by procedure, operator actions include the initiation of an immediate manual trip following receipt of an automatic trip signal and there are no clear indications that the automatic trip failed (such as a time delay following indications that a trip setpoint was exceeded), it may be difficult to determine if the reactor was shut down because of automatic trip or manual actions.
If a subsequent review of the trip actuation indications reveals that the automatic trip did not cause the reactor to be shut down, consideration should be given to evaluating the fuel for potential damage, and the reporting requirements of 10CFR50.72 should be considered for the transient event.
This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor trip that results in a reactor shutdown, and either a subsequent operator manual action taken at the reactor control consoles or an automatic trip is successful in shutting down the reactor.
This event is a precursor to a more significant condition and thus represents a potential degradation of the level of safety of the plant.
Following the failure on an automatic reactor trip, operators will promptly initiate manual actions at the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor trip). Ifthese manual actions are successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plant's decay heat removal systems.
If an initial manual reactor trip is unsuccessful, operators will promptly take manual action at another location(s) on the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor trip) using a different switch). Depending upon several factors, the initial or subsequent effort to manually trip the reactor, or a concurrent plant condition, may lead to the generation of an automatic reactor trip signal. If a subsequent manual or automatic trip is successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plant's decay heat removal systems.
A manual action at the reactor control Consoles is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor trip). This action does not include manually driving in control rods or implementation of boron injection strategies. Actions taken at back-panels or other locations within the Control Room, or any location outside the Control Room, are not considered to be "at the reactor control consoles".
r[Document No.]                                     Rev. 0                          Page 167 of 240

ATTACHMENT 1 EAL Bases The plant response to the failure of an automatic or manual reactor trip will vary based upon several factors including the reactor power level prior to the event, availability of the condenser, performance of mitigation equipment and actions, other concurrent plant conditions, etc. If subsequent operator manual actions taken at the reactor control consoles are also unsuccessful in shutting down the reactor, then the emergency classification level will escalate to an Alert via IC SA6. Depending upon the plant response, escalation is also possible via IC FAI. Absent the plant conditions needed to meet either IC SA6 or FA1, an Unusual Event declaration is appropriate for this event.
Should a reactor trip signal b~e generated as a result of plant work (e.g., RPS setp~oint testing),
the following classification guidance should be applied.
* If the signal causes a plant transient that should have included an automatic reactor trip and the RPS fails to automatically shutdown the reactor, then this IC and the EALs are applicable, and shoul~d be evaluated.
* If the signal does not cause a plant transient and the trip failure is determined through other means (e.g., assessment of test results), then this IC and the EALs are not applicable and no classification is warranted.
ONS Basis Reference(s):
: 1. ONS Technical Specifications Section 3.3.1 Reactor Protective System (RPS)
Instrumentation   - Operating
: 2. ONS Technical Specifications Table 1.1-1 Modes
: 3. UFSAR Section 7.2.3.7 Manual Trip
: 4. OSC-02820 Emergency Procedure Setpoints, Setpoint No. 7.1.7.5
: 5. NEI 99-01 SU5 r[Document No.]                                  Rev. 0                             Page 168 of 240

ATTACHMENT 1 EAL Bases
*Category:                     S - System Malfunction Subcategory:                  6 - RPS Failure Initiating Condition:         Automatic or manual trip fails to shut down the reactor EAL:
SU6.2            Unusual Event A manual trip did not shut down the reactor as indicated by reactor power > 5% after any manual trip action was initiated            I AND A subsequent automatic trip or the manual trip pushbutton is successful in shutting down the reactor as indicated by reactor power          < 5% (Note 8)
Note 8:   A manual trip action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving-in control rods or implementation of boron injection strategies.
Mode Applicability:
1 - Power Operation Definition(s):
None Basis:
This EAL addresses a failure of a manually initiated trip in the absence of having exceeded an automatic RPS trip setpoint and a subsequent automatic or manual trip is successful in shutting down the reactor (ref. 1).
Following a successful reactor trip, rapid insertion of the control rods occurs. Nuclear power promptly drops to a fraction of the original power level and then decays to a level several decades less with a negative startup rate. The reactor power drop continues until reactor power reaches the point at which the influence of source neutrons on reactor power starts to be observable. A predictable post-trip response from an a~utomatic reactor trip signal should therefore consist of a prompt drop in reactor power as sensed by the nuclear instrumentation and a lowering of power into the source range. For the purpose of emergency classification a successful trip has occurred when there is sufficient rod insertion from the manual trip to bring the reactor power below the Power Operation Mode threshold level of 5% (ref. 2).
5% rated power is the Power Operation mode threshold. Below 5%, plant response will be similar to that observed during a normal shutdown. Nuclear instrumentation can be used to determine if reactor power is greater than or equal to 5% power (ref. 1, 2).
For the purposes of emergency classification, a successful manual trip action is that which can be quickly performed from the reactor control console (actuation of the manual trip pushbutton). There is a separate set of switch contacts in series with the output of each reactor trip component. All switch contacts are actuated through a mechanical linkage from a single pushbutton. Reactor shutdown achieved by use of other trip actions such as opening supply r[Document No.]                                        Rev. 0                                 Page 169 of 240

ATTACHMENT 1 EAL Bases breakers, emergency boration, or manually driving control rods) do not constitute a successful manual trip (ref. 3).
Following any automatic RPS trip signal, insertion of redundant manual trip signals are performed to back up the automatic RPS trip function and ensure reactor shutdown is achieved. Even if a subsequent automatic trip signal or the first subsequent manual trip signal inserts all control rods to the full-in position immediately after the initial failure of the manual trip, the lowest level of classification that must be declared is an Unusual Event.
If both subsequent automatic and subsequent manual reactor trip actions in the Control Room fail to reduce reactor power     < 5% follbwing a failure of an initial manual trip, the event      I escalates to an Alert under EAL SA6.1.
This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor trip that results in a reactor shutdown, and either a subsequent operator manual action taken at the reactor control consoles or an automatic trip is successful in shutting down the reactor.
This event is a precursor to a more significant condition and thus represents a potential degradation of the level of safety of the plant.
Following the failure on 'an automatic reactor trip, operators will promptly initiate manual actions at the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor trip). If these manual actions are successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plant's decay heat removal systems.
If an initial manual reactor trip is unsuccessful, operators will promptly take manual action at another location(s) on the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor trip) using a different switch). Depending upon several factors, the initial or subsequent effort to manually the reactor, or a concurrent plant condition, may lead to the generation of an automatic reactor trip signal. If a subsequent manual or automatic trip is successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plant's decay heat removal systems.
A manual action at the reactor control consoles is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor trip). This action does not include manually driving in control rods or implementation of boron inject~ion strategies. Actions taken at back-panels or other locations within the Control Room, or any location outside the Control Room, are not considered to be "at the reactor control consoles".
The plant response to the failure of an automatic or manual reactor trip will vary based upon several factors including the reactor power level prior to the event, availability of the condenser, performance of mitigation equipment and actions, other conicurrent plant conditions, etc. If subsequent operator manual actions taken at the reactor control consoles are also unsuccessful in shutting down the reactor, then the emergency classification level will escalate to an Alert via IC SA6. Depending upon the plant response, escalation is also possible via IC FAI. Absent the plant conditions needed to meet either IC SA6 or FA1, an Unusual Event declaration is appropriate for this event.
Should a reactor trip signal be generated as a result of plant work (e.g., RPS setpoint testing),
the following classification guidance should be applied.
I[Document No.]                          I         Rev. 0           IPage                    170 of 240

ATTACHMENT 1 EAL Bases
* Ifthe signal causes a plant transient that should have included an automatic reactor trip and the RPS fails to automatically shutdown the reactor, then this IC and the EALs are applicable, and should be evaluated.
* If the signal does not cause a plant transient and the trip failure is determined through other means (e.g., assessment of test results), then this IC and the EALs are not applicable and no classification is warranted.
ONS Basis Reference(s):
: 1. ONS Technical Specifications Section 3.3.11 Reactor Protective System (RPS)
Instrumentation - Operating
: 2. ONS Technical Specifications Table 1.1-1
: 3. UFSAR Section 7.2.3.7 Manual Trip
: 4. OSC-02820 Emergency Procedure Setpoints, Setpoint No. 7.1.7.5
: 5. NEI 99-01 SU5 I[Document No.]                       r        Rev. 0                             Page 171 of 2401

ATTACHMENT 1 EAL Bases Category:                     S - System Malfunction Subcategory:                  6 - RPS Failure Initiating Condition:         Automatic or manual trip fails to shut down the reactor and subsequent manual actions taken at the reactor control consoles are not successful in shutting down the reactor EAL:
SA6.1            Alert An automatic or manual trip fails to shut down the reactor as indicated by reactor power
  ->5%
AND Manual trip pushbutton is not successful in shutting down the reactor as indicated by reactor power --5% (Note 8)
Note 8:   Amanual trip action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving incontrol rods or implementation of boron injection strategies.
Mode Applicability:
1 - Power Operation Definition(s):
None Basis:
This EAL addresses any automatic or manual reactor trip signal that fails to shut down the reactor followed by a subsequent manual trip that fails to shut down the reactor to an extent the reactor is producing significant power (ref. 1).
Following a successful reactor trip, rapid insertion of the control rods occurs. Nuclear power promptly drops to a fraction of the original power level and then decays to a level several decades less with a negative startup rate. The reactor power drop continues until reactor power reaches the point at which the influence of source neutrons on reactor power starts to be observable. A predictable post-trip response from an automatic reactor trip signal should therefore consist of a prompt drop in reactor power as sensed by the nuclear instrumentation and a lowering of power into the source range. For the purpose of emergency classification a successful trip has occurred when there is sufficient rod insertion from the manual trip to bring the reactor power below 5% (ref. 2).
5% rated power is the Power Operation mode threshold. Below 5%, plant response will be similar to that observed during a normal shutdown. Nuclear instrumentation can be used to determine if reactor power is greater than or equal to 5% power (1, 2).
For the purposes of emergency classification, a successful manual trip action is that which can be quickly performed from the reactor control console (actuation of the manual trip pushbutton). There is a separate set of switch contacts in series with the output of each reactor trip component. All switch contacts are actuated through a mechanical linkage from a single I[Document No.]                                       Rev. 0              IPage                      172 of 240I

ATTACHMENT 1 EAL Bases pushbutton. Reactor shutdown achieved by use of other trip actions such as opening supply breakers, emergency boration, or manually driving control rods) do not constitute a successful manual trip (ref. 3).
Escalation of this event to a Site Area Emergency would be under EAL SS6.1 or Emergency Coordinator judgment.
This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor trip that results in a reactor shutdown, and subsequent operator manual actions taken at the reactor control consoles to shutdown the reactor are also unsuccessful. This condition represents an actual or potential substantial degradation of the' level of safety of the plant. An emergency declaration is r~equired even if the reactor is subsequently shutdown by an action taken away from the reactor control consoles since this event entails a significant failure of the RPS.
A manual action at the reactor control console is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor trip). This action does not include manually driving in control rods or implementation of boron injection strategies. If this action(s) is unsuccessful, operators would immediately pursue additional manual actions at locations away from the reactor control console (e.g., locally opening breakers). Actions taken at back panels or other locations within the Control Room, or any location outside the Control Room, are not considered to be "at the reactor control console".
The plant response to the failure of an automatic or manual reactor trip will vary based upon several factors including the reactor power level prior to the event, availability of the condenser-, performance of mitigation equipment and actions, other concurrent plant conditions, etc. If the failure to shut down the reactor is prolonged enough to cause a challenge to the core cooling or RCS heat removal safety functions, the emergency classification level will escalate to a Site Area Emergency via IC SS6. Depending upon plant responses and symptoms, escalation is also possible via IC ESI. Absent the plant conditions needed to meet either IC SS6 or FS1, an Alert declaration is appropriate for this event.
It is recognized that plant responses or symptoms may also require an Alert declaration in accordance with the Recognition Category F ICs; however, this IC and EAL are included to ensure a timely emergency declaration.
A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.
ONS Basis Reference(s):
: 1. ONS Technical Specifications Section 3.3.1 Reactor Protective System (RPS)
Instrumentation   - Operating
: 2. ONS Technical Specifications Table 1.1-1
: 3. UFSAR Section 7.2.3.7 Manual Trip
: 4. OSC-02820 Emergency Procedure Setpoints, Setpoint No. 7.1.7.5
: 5. NEI 99-01 SA5 S[Document No.]                                  Rev. 0                            Page 173 of 240

ATTACHMENT I EAL Bases Category:       S - System Malfunction Subcategory:              6 - RPS Failure Initiating Condition:     Inability to shut down the reactor causing a challenge to core cooling or RCS heat removal EAL:
SS6.1           Site Area Emergency An automatic or manual trip fails to shut down the reactor as indicated by reactor power
-&#x17d;5%
AND All actions to shut down the reactor are not successful as indicated by reactor power
->5%
AND EITHER:
* CETCs >1200&deg;F on ICCM
* RCS subcooling < 0&deg;F Mode Applicability:
1 - Power Operation Definition(s):
None Basis:
This EAL addresses the following:
      *Any automatic reactor trip signal (ref. 1) followed by a manual trip that fails to shut down the reactor to an extent the reactor is producing energy in excess of the heat load for which the safety systems were designed (ref. 5), and
      *Indications that either core cooling is extremely challenged or heat removal is extremely challenged.
The combination of failure of both front line and backup protection systems to function in response to a plant transient, along with the continued production of heat, poses a direct threat to the Fuel Clad and RCS Barriers.
Reactor shutdown achieved by use of other trip actions such as opening supply breakers, emergency boration, or manually driving control rods are also credited as a successful manual trip provided reactor power can be reduced below 5% before indications of an extreme challenge to either core cooling or heat removal exist (ref. 2, 3).
5% rated power is the Power Operation mode threshold. Below 5%, plant response will be similar to that observed during a normal shutdown. Nuclear instrumentation can be used to determine if reactor power is greater than or equal to 5% power.
[Dcmn o]Rev.                                0                          Page 174 of 240

ATTACHMENT 1 EAL Bases Indication of continuing core cooling degradation is manifested by CETCs are reading greater than 1200&deg;F. This setpoint is used as an indication of an extreme ICC condition and entry into the Oconee Severe Accident Guidelines (OSAG) is initiated for further mitigative actions (ref.
4).
Indication of inability to adequately remove heat from the RCS is manifested by subcooling less than 0&deg;F (ref. 6).
This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor trip that results in a reactor shutdown, all subsequent operator actions to manually shutdown the reactor are unsuccessful, and continued power generation is challenging th* capability to adequately remove heat from the core and/or the RCS. This condition will lead to fuel damage if additional mitigation actions are unsuccessful and thus warrants the declaration of a Site Area Emergency.
In some instances, the emergency classification resulting from this IC/EAL may be higher than that resulting from an assessment of the plant responses and symptoms against the Recognition Category F ICs/EALs. This is appropriate in that the Recognition Category F ICs/EALs do not address the additional threat posed by a failure to shut down the reactor. The inclusion of this IC and EAL ensures the timely declaration of a Site Area Emergency in response to prolonged failure to shutdown the reactor.
A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.
Escalation of the emergency classification level would be via IC RG1 or FGI.
ONS Basis Reference(s):
: 1. ONS Technical Specifications Section 3.3.1 Reactor Protective System (RPS)
Instrumentation   - Operating
: 2. ONS Technical Specifications Table 1.1-1
: 3. UFSAR Section 7.2.3.7 Manual Trip
: 4. OSC-02820 Emergency Procedure Setpoints, Setpoint No. 7.1.1.7
: 5. OSC-02820 Emergency Procedure Setpoints, Setpoint No. 7.1.7.5
: 6. EP/1 ,2,3/A/1800/001 Loss of Subcooling Margin
: 7. NEI 99-01 SS5 S[Document No.]                                 Rev. 0                         Page 175 of 240

ATTACHMENT 1 EAL Bases Category:                S - System Malfunction Subcategory:              7 - Loss of Communications Initiating Condition:     Loss of all onsite or offsite communications capabilities EAL:
SU7.1          Unusual Event Loss of all Table S-4 onsite communication methods OR Loss of all Table S-4 offsite communication methods OR Loss of all Table S-4 NRC communication methods Mode Applicability:
1 - Power Operation, 2 - Startup, 3  - Hot Standby, 4 - Hot Shutdown Definition(s):
None S[Document No.]                                 Rev. 0                           Page 176 of 240

ATTACHMENT 1 EAL Bases Basis:
Onsite, offsite and NRC communications include one or more of the systems listed in Table S-4 (ref. 1).
: 1. Commercial phone service The Commercial phone service does not go through the site telephone system.
: 2. O)NS site phone system The site phone system is generator and ibattery backed with:
* Fiber-Optic to Charlotte GO (65 lines)
* Telephone line to Easley (6 circuits)
* Anderson (4 lines)
* Six Mile (4 lines)
* Site Telephone System: Inward and outward direct dial available from the Control Room, TSC, and OSC
: 3. EOF phone system The emergency communications systems at the Charlotte EOF are designed to ensure the reliable, timely flow of information between all parties having an emergency response role.
: 4. Public Address (Paging) system The paging system provides paging and party line communications between stations located throughout the plant. Inside and outside type wall and desk-mounted stations are used to communicate between roaming personnel and fixed work locations. Plant-wide instructions are issued using the paging feature.
: 5. Onsite radio system The onsite radio system receives emergency backup power from Keowee Hydro Units supporting communications with: Control Room 1&2, 3, Fire Brigade, Chemistry, Safety, Radiation Protection, Maintenance, Medical Emergency Response Team, and Hazardous Materials Response Team.
: 6. DEMNET DEMNET is the primary means of offsite communication. This circuit allows intercommunication among the EOF, TSC, control room, counties, and states. DEMNET operates as an internet based (VolP) communications system with a satellite back-up. Should the internet transfer rate become slow or unavailable, the DEMNET will automatically transfer to satellite mode.
: 7. Offsite radio system The offsite radio system is battery backed supporting communications with: Control Room Units 1&2, TSC, Field Monitoring Teams, EOF, counties and State of South Carolina.
.[Document No.]                                Rev. 0          rPage                177 of 240

ATTACHMENT 1 EAL Bases
: 8. NRC Emergency Telephone System (ETS)
The NRC uses a Duke Energy dedicated telephone line which allows direct telephone communications from the plant to NRC regional and national offices. The Duke Energy communications line provides a link independent of the local public telephone network.
Telephones connected to this network are located in the Oconee Control Rooms, TSC, and EOF and can be used to establish NRC Emergency Notification System (ENS) and Health Physics Network (HPN) capability.
: 9. Satellite Phone\
Satellite Phones can be used for external communications This EAL is the hot condition equivalent of the cold condition EAL CU5.1.
This IC addresses a significant loss of on-site or offsite communications capabilities. While not a direct challenge to plant or personnel safety, this event warrants prompt notifications to OROs and the NRC.
This IC should be assessed only when extraordinary means are being utilized to make communications possible (e~g., use of non-plant, privately owned equipment, relaying of on-site information via individuals or multiple radio transmission points, individuals being sent to offsite locations, etc.).
The first EAL condition addresses a total loss of the communications methods used in support of routine plant operations.
The second EAL condition addresses a total loss of the communications methods used to notify all OROs of an emergency declaration. The OROs referred to here are the State EOC and FEO, Pickens County LEC and EOC, and Oconee County LEC and EOC.
The third EAL addresses a total loss of the communications methods used to notify the NRC of an emergency declaration.
ONS Basis Reference(s):
: 1. ONS Emergency Plan, Section 7.2 Communications Systems
: 2. NEI 99-01 SU6 S[Document No.]                                 Rev. 0           *Page                  178 of 240I

ATTACHMENT 1 EAL Bases Category:                  S - System Malfunction Subcategory:               8 - Containment Failure Initiating Condition:      Failure to isolate containment or loss of containment pressure control.
EAL:
SU8.1          Unusual Event Any penetration is not closed within 15 min. of a VALID ES actuation signal OR Containment pressure > 10 psig with < one full train of containment heat removal system (1 RBS with    >  700 gpm spray flow OR 2 RBCUs) operating per design for
 > 15 min. (Note 1)
Note 1:   The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
Mode Applicability:
1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):
VALID - An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment.
Basis:
Reactor Building isolations are initiated by Engineered Safeguards Actuation Channels 5 and 6 in response to a high reactor building pressure signal (3.0 psig) (ref. 1, 2, 4).
Two engineered safeguards Systems, the Reactor Building Spray System and the Reactor Building Cooling System, are provided to remove heat from the containment atmosphere following an accident. Both the Reactor Building Spray System and the Reactor Building Cooling System, with either at full capacity, are individually capable of maintaining the containment pressure below the design limit following a LOCA or MSLB. (ref. 1, 3)
The Reactor Building Spray (RBS) System consists of two separate trains of equal capacity.
Spray flow greater or equal to 700 gpm satisfies the spray flow design requirement. The Reactor Building pressure setpoint (10 psig) is the pressure at which the Reactor Building Spray equipment should actuate and begin performing its function (ref. 1, 2, 3, 5).
Each of three Reactor Building Cooling Units (RBCUs) consists of a fan, cooling coils, and the required distribution duct work. The Reactor Building atmosphere is circulated past cooling coils by fans and returned to the building. Cooling water for the cooling units is supplied by the Low Pressure Service Water System. The Reactor Building Cooling System provides the design heat removal capacity with two of three coolers operating (ref. 1).
I[Document No.].*                                     Rev. 0                    .       Page 179 of 240I

ATTACHMENT 1 EAL Bases This EAL addresses a failure of one or more containment penetrations to automatically isolate (close) when required by an actuation signal. It also addresses an event that results in high containment pressure with a concurrent failure of containment pressure control systems.
Absent challenges to another fission product barrier, either condition represents potential degradation of the level of safety of the plant.
For the first condition, the containment isolation signal must be generated as the result on an off-normal/accident condition (e.g., a safety injection or high containment pressure); a failure resulting from testing or maintenance does not warrant classification. The determination of containment and penetration status - isolated or not isolated i- should be made in accordance with the appropriate criteria contained in the plant APs and EOPs. The 15-minute criterion is included to allow operators time to manually isolate the required penetrations, if possible.
The second condition addresses a condition where containment pressure is greater than the setpoint at which containment energy (heat) removal systems are designed to automatically actuate, and less than one full train of equipment is capable of operating per design. The 15-minute criterion is included to allow operators time to manually start equipment that may not have automatically started, if possible. The inability to start the required equipment indicates that containment heat removal/depressurization systems (e.g., containment sprays) are either lost or performing in a degraded manner.
This event would escalate to a Site Area Emergency in accordance with IC FS1 if there were a concurrent loss or potential loss of either the Fuel Clad or RCS fission product barriers.
ONS Basis Reference(s):
: 1. UFSAR Section 6.2.2 Containment Heat Removal Systems
: 2. UFSAR Table 7-2 Engineered Safeguards Actuation Conditions
: 3. UFSAR Table 6-25 Minimum Acceptable Combinations of Containment Heat Removal Equipment Performance
: 4. OSC-02820 Emergency Procedure Setpoints, Setpoint No. 7.1 .1.1
: 5. OSC-02820 Emergency Procedure Setpoints, Setpoint No. 7.4.1.2
: 6. NEI 99-01 SU7
[Document No.].                                Rev. 0                           Page 180 of 240

ATTACHMENT 1 EAL Bases Category:                 S - System Malfunction Subcategory:              9 - Hazardous Event Affecting Safety Systems Initiating Condition:     Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode EAL:
SA9.1          Alert The occurrence bf any Table S-5 hazardous event AND EITHER:
       *Event damage has caused indications of degraded performance in at least one train of a SAFETY SYSTEM needed for the current operating mode*
* The event has caused VISIBLE DAMAGE to a SAFETY SYSTEM component or structure needed for the current operating mode
[         Table S-5     Hazardous Events
* Seismic event (earthquake)
* Internal or external FLOODING event
* High winds or tornado strike
* FIRE
* EXPLOSION
* Other events with similar hazard characteristics as determined by the Shift Manager Mode Applicability:
1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):
EXPLOSION - A rapid, violent and catastrophic failure of a piece of equipment due to combustion, chemical reaction or overpressurization. A release of steam (from high energy lines or components) or an electrical component failure (caused by short circuits, grounding, arcing, etc.) should not automatically be considered an explosion. Such events require a post-event inspection to determine if the attributes of an explosion are present.
FIRE - Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred but is NOT required if large quantities of smoke and heat are observed.
FLOODING - A condition where water is entering a room or area faster than installed equipment is capable of removal, resulting in a rise of water level within the room or area.
SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems S[Document No.]                                 Rev. 0                          Page 181 of 240

ATTACHMENT 1 EAL Bases classified as safety-related (as defined in 10OCFR50.2):
Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:
(1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.
VISIBLE DAMAGE - Damage to a component or structure that is readily obse rvable without measurements, testing, or analysis. The visual impact of the damage is sufficient to cause concern regarding the operability or reliability of the affected component or structure.
Basis:
* The significance of seismic events are discussed under EAL HU2.1 (ref. 1).
* Internal FLOODING may be caused by events such as component failures, equipment misalignment, or outage activity mishaps.
   *External flooding at ONS is highly unlikely since the probable maximum flood (PMF) would be contained by the Keowee Reservoir. Plant grade elevation is 796.0 ft MSL. The minimum external access elevation for the Auxiliary, Turbine, and Service Buildings is 796.5 ft MSL which provides a 6 inch water sill. (ref. 2)
   *High winds in excess of design (95 mph) or tornado strikes can cause significant structural damage (ref. 3).
   *Areas containing functions and systems required for safe shutdown of the plant are identified by fire area (ref. 4, 5).
   *An explosion that degrades the performance of a SAFETY SYSTEM train or visibly damages a SAFETY SYSTEM component or structure would be classified under this EAL.
This IC addresses a hazardous event that causes damage to a SAFETY SYSTEM, or a structure containing SAFETY SYSTEM components, needed for the current operating mode.
This condition significantly reduces the margin to a loss or potential loss of a fission product barrier, and therefore represents an actual or potential substantial degradation of the level of safety of the plant.
The first condition addresses damage to a SAFETY SYSTEM train that is in service/operation since indications for it will be readily available. The indications of degraded performance should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train.
The second condition addresses damage to a SAFETY SYSTEM component that is not in service/operation or readily apparent through indications alone, or to a structure containing SAFETY SYSTEM components. Operators will make this determination based on the totality of available event and damage report information. This is intended to be a brief assessment not requiring lengthy analysis or quantification of the damage.
Escala'tion of the emergency classification level would be via IC FS1 or RSI.
I[Document No.]                                   Rev. 0                         Page 182 .of 240I

ATTACHMENT I EAL Bases ONS Basis Reference(s):
I. AP/O/A/1 700/005 Earthquake
: 2. UFSAR Section 3.4.1.1 Flood Protection Measures for Seismic Class 1 Structures
: 3. UFSAR Section 3.3.1.1 Design Wind Velocity
: 4. OSS-0254.00-00-4008 Design Bases Specification for Fire Protection
: 5. AP/1,2,3/A/1700/050 Challenging Plant Fire
: 6. NEI 99-01 SA9
[Document No.]                             Rev. 0                            Page 183 of 240

ATTACHMENT I EAL Bases Category F - Fission Product Barrier Degradation EAL Group: Hot Conditions (RCS temperature        > 200&deg;F); EALs in this category are applicable only in one or more hot operating modes.
EALs in this category represent threats to the defense in depth design concept that precludes the release of highly radioactive fission products to the environment. This concept relies on multiple physical barriers any one of which, if maintained intact, precludes the release of significant amounts of radioactive fission products to the environment. The primary fission product barriers are:
A. Fuel Clad (FC): The Fuel Clad Barrier consists of the cladding material that contains the fuel pellets.
B. Reactor Coolant System (RCS): The RCS Barrier includes the RCS primary side and its connections up to and including the pressurizer safety and relief valves, and other connections up to and including the primary isolation valves.
C. Containment (CMT): The Containment (Reactor Building) Barrier includes the Reactor Building and connections up to and including the outermost containment isolation valves. This barrier also includes the main steam, feedwater, and blowdown line extensions outside the Reactor Building up to and including the outermost secondary side isolation valve. Containment Barrier thresholds are used as criteria for escalation of the ECL from Alert to a Site Area Emergency or a General Emergency.
The EALs in this category require evaluation of the loss and potential loss thresholds listed in the fission product barrier matrix of Table F-I (Attachment 2). "Loss" and "Potential Loss" signify the relative damage and threat of damage to the barrier. "Loss" means the barrier no longer assures containment of radioactive materials. "Potential Loss" means integrity of the barrier is threatened and could be lost if conditions continue to degrade. The number of barriers that are lost or potentially lost and the following criteria determine the appropriate emergency classification level:
Alert:
Any loss or any potential loss of either Fuel Clad or RCS Barrier Site Area Emerqency:
Loss or potential loss of any two barriers General Emergqency:
Loss of any two barriersand loss or potential loss of third barrier The logic used for emergency classification based on fission product barrier monitoring should reflect the following considerations:                                                           .
* The Fuel Clad Barrier and the RCS Barrier are weighted more heavily than the Containment Barrier.
* Unusual Event ICs associated with RCS and Fuel Clad Barriers are addressed under System Malfunction ICs.
* For accident conditions involving a radiological release, evaluation of the fission product barrier thresholds will need to be performed in conjunction with dose assessments to l/[Document      No.]                               Rev. 0.                         Page 184 of 240 I

ATTACHMENT 1 EAL Bases ensure correct and timely escalation of the emergency classification. For example, an evaluation of the fission product barrier thresholds may result in a Site Area Emergency classification while a dose assessment may indicate that an EAL for General Emergency IC RG1 has been exceeded.
     *The fission product barrier thresholds specified within a scheme reflect plant-specific ONS design and operating characteristics.
     *As used in this category, the term RCS leakage encompasses not just those types defined in Technical Specifications but also includes the loss of RCS mass to any S location- inside the primary containment, ar* interfacing system, or outside of the primary containment. The release of liquid or steam mass from the RCS due to the as designed/expected operation of a relief valve is not considered to be RCS leakage.
     *At  the Site Area Emergency level, EAL users should maintain cognizance of how far present conditions are from meeting a threshold that would require a General Emergency declaration. For example, if the Fuel Clad and RCS fission product barriers were both lost, then there should be frequent assessments of containment radioactive inventory and integrity. Alternatively, if both the Fuel Clad and RCS fission product barriers were potentially lost, the Emergency Coordinator would have more assurance that there was no immediate need to escalate to a General Emergency.
I[Document No.]                                Rev. 0                           Page 185 of 240I

ATTACHMENT I EAL Bases Category:                  Fission Product Barrier Degradation Su bcategory:              N/A Initiating Condition:     Any loss or any potential loss of either Fuel Clad or RCS barrier EAL:
FA1.1           Alert Any loss or any potential loss of either Fuel Clad or RCS barrier (Table F-I)
Mode Applicability:
1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):
None Basis:
Fuel Clad, RCS and Containment comprise the fission product barriers. Table F-I (Attachment
: 2) lists the fission product barrier thresholds, bases and references.
At the Alert classification level, Fuel Clad and RCS barriers are weighted more heavily than the Containment barrier. Unlike the Containment barrier, loss or potential loss of either the Fuel Clad or RCS barrier may result in the relocation of radioactive materials or degradation of core cooling capability. Note that the loss or potential loss of Containment barrier in combination with loss or potential loss of either Fuel Clad or RCS barrier results in declaration of a Site Area Emergency under EAL FSI.1 ONS Basis Reference(s):
I. NEI 99-01 FA1 I[Document No;]                                  Rev. 0         IPage                   186of2401

ATTACH MENT 1 EAL Bases Category:                  Fission Product Barrier Degradation Subcategory:             .,N/A Initiating Condition:      Loss or potential loss of any two barriers EAL:
FSI.1           Site Area Emergency Loss or potential loss of any two barriers (Table F-i)
Mode Applicability:
1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):
None Basis:
Fuel Clad, RCS and Containment comprise the fission product barriers. Table F-I (Attachment
: 2) lists the fission product barrier thresholds, bases and references.
At the Site Area Emergency classification level, each barrier is weighted equally. A Site Area Emergency is therefore appropriate for any combination of the following conditions:
* One barrier loss and a second barrier loss (i.e., loss - loss)
* One barrier loss and a second barrier potential loss (i.e., loss - potential loss)
* One barrier potential loss and a second barrier potential loss (i.e., potential loss -
potential loss)
At the Site Area Emergency classification level, the ability to dynamically assess the proximity of present conditions with respect to the threshold for a General Emergency is important. For example, the existence of Fuel Clad and RCS Barrier loss thresholds in addition to offsite dose assessments would require continual assessments of radioactive inventory and Containment integrity in anticipation of reaching a General Emergency classification. Alternatively, if both Fuel Clad and RCS potential loss thresholds existed, the Emergency Coordinator would have greater assurance that escalation to a General Emergency is less imminent.
ONS Basis Reference(s):
: 1. NE199-O1 FSI I[Document No.]                                  Rev. 0                          Page 187 of 240I

ATTACHMENT 1 EAL Bases Category:                 Fission Product Barrier Degradation Subcategory:              N/A Initiating Condition:     Loss of any two barriers and loss or potential loss of third barrier EAL:
FGI.1            General Emergency Loss of any two barriers AND Loss or potential loss of third barrier (Table F-I)
Mode Applicability:
I - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):
None Basis:
Fuel Clad, RCS and Containment comprise the fission product barriers. Table F-I (Attachment
: 2) lists the fission product barrier thresholds, bases and references.
At the General Emergency classification level each barrier is weighted equally. A General Emergency is therefore appropriate for any combination of the following conditions:
* Loss of Fuel Clad, RCS and Containment Barriers
* Loss of Fuel Clad and RCS Barriers with potential loss of Containment Barrier
* Loss of RCS and Containment Barriers with potential loss of Fuel Clad Barrier
* Loss of Fuel Clad and Containment Barriers with potential loss of RCS Barrier ONS Basis Reference(s):
I. NEI 99-01 FG1
[Document No.]                                  Rev. 0                          Page 188 of 240

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Introduction Table F-I lists the threshold conditions that define the Loss and Potential Loss of the three fission product barriers (Fuel Clad, Reactor Coolant System, and Containment). The table is structured so that each of the three barriers occupies adjacent columns. Each fission product barrier column is further divided into two columns; one for Loss thresholds and one for Potential Loss thresholds.
The first column of the table (to the left of the Fuel Clad Loss column) lists the categories (types) of fission Product barrier thresholds. The fission product barrier &#xb6;.ategories are:
A. RCS or SG Tube Leakage B. Inadequate Heat removal C. CMT Radiation / RCS Activity D. CMT Integrity or Bypass E. Emergency Coordinator Judgment Each category occupies a row in Table F-I thus forming a matrix defined by the categories.
The intersection of each row with each Loss/Potential Loss column forms a cell in which one or more fission product barrier thresholds appear. If NEI 99-01 does not define a threshold for a barrier Loss/Potential Loss, the word "None" is entered in the cell.
Thresholds are assigned sequential numbers within each Loss and Potential Loss column beginning with number one. In this manner, a threshold can be identified by its category title and number. For example, the first Fuel Clad Barrier Loss in Category A would be assigned "FC Loss A.I ," the third Containment Barrier Potential Loss in Category C would be assigned "CMT P-Loss C.3," etc.
If a cell in Table F-i contains more than one numbered threshold, each of the numbered thresholds, if exceeded, signifies a Loss or Potential Loss of the barrier. It is not necessary to exceed all of the thresholds in a category before declaring a barrier Loss/Potential Loss.
Subdivision of Table F-i by category facilitates association of plant conditions to the applicable fission product barrier Loss and Potential Loss thresholds. This structure promotes a systematic approach to assessing the classification status of the fission product barriers.
When equipped with knowledge of plant conditions related to the fission product barriers, the EAL-user first scans down the category column of Table F-i, locates the likely category and then reads across the fission product barrier Loss and Potential Loss thresholds in that category to determine if a threshold has been exceeded. If a threshold has not been exceeded, the EAL-user proceeds to the next likely category and continues review of the thresholds in the new category If the EAL-user determines that any threshold has been exceeded, by definition, the barrier is lost or potentially lost - even if multiple thresholds in the same barrier column are exceeded, only that one barrier is lost or potentially lost. The EAL-user must examine each of the three fission product barriers to determine if other barrier thresholds in the category are lost or potentially lost. For example, if containment radiation is sufficiently high, a Loss of the Fuel Clad and RCS Barriers and a Potential Loss of the Containment Barrier can occur. Barrier S[Document No.]                                 Rev. 0          IPage                   189 of 240I

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Losses and Potential Losses are then applied to the algorithms given in EALs FG1 .1, ES1 .1, and FA1 .1 to determine the appropriate emergency classification.
In the remainder of this Attachment, the Fuel Clad Barrier threshold bases appear first, followed by the RCS Barrier and finally the Containment Barrier threshold bases. In each barrier, the bases are given according category Loss followed by category Potential Loss beginning with Category A, then B,..., E.
S[Document No.]                             Rev. 0                           Page 190 of 240I

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Table F-I Fission Product Barrier Threshold Matrix Fuel Clad (FC) Barrier                                              Reactor Coolant System (RCS) Barrier                                                        Containment (CMT) Barrier Category                    LOSS                            Potential Loss                                    Loss                              Potential Loss                                  Loss                          Potential Loss 1.RCS leakage > normal makeup capacity due to EITHER:
S1.                                                                    An automatic or manual ES
* UNISOLABLE RCS leakage At                                                                                            actuation required by EITHER:                &deg; SG tube leakage RCS or                     None                      1. RVLS -<O(Note 9)
* UNISOLABLE RCS                                                             1. Aleaking SG ia FAULTED outside                       None of containment SG Tube                                                                                                leakage                         2. RCS cooldown < 400&deg;F Leakage                                                                                             *SG tube RUPTURE                        at > 100&deg;Flhr OR HPI has operated in the injection mode with no RCPs operating
: 1. CETCs > 700&deg;F                                                                 1. RCS heat removal cannot be B                                                                                                                                       established                                                                1. CETCs >1200&deg;F Indqae 1    E~        20F2.                       RCS beat removal cannot be                                                           AND                                                 None                       AND Heat                                                  esalse                                                  oeRSsboln                                         FRestoration                                                         procedurea not Removal                                                    AND                                                                                                                                                             effective within 15 mam.(Note 1)
RCS subcooting    < 0 "F                                                      2. HPI forced cooling initiated C                                                                                         1. containment radiation:
: 1. 1/213RIA 57/58 > Table F-2                                                           &deg; 1,3 RIA57158 >1.0 R/hr                                                                                          1   12315/vTbeF2clm CMT            column "FC Loss"                                    None
* 2 RIA 57 > 1.6 Rlhr                              None                                    None                      C1 T PotentiA 75 Lossa ble         oum Radiation     2.CooanTativtyt30ntiial              lLossA8>.Olh I RCS       2     oln      ciiy>30pim           E                                                          I   8>10Rh Activity     ________________
: 1. Containment isolation is required                           1. Containment pressure      > 59 psig AND EITHER:
* Containment integrity has     2. Containment hydrogen D                                                                                                                                       been lost based on EmrecDoriao concentration   >  4%
CMT                         None                                    None                                      None                                      None                            judgment Integrity                                                                                                                                                                              *UNISOLABLE pathway from        3. Containment presaure >10 psig Continmet tothewith               < one full train of containment or BypacssCnanetoth                                                                                                                                                                                                        heat removal system (1 RBS with >
envionmet euats700               gpm spray flow OR 2 RBCUs) operating per design for > 15 mai.
: 2. Indications of RCS leakage            (Note 1) outside of Containment E         1. Any condition in the opinion of      1.Any condition in the judgment of     1. An1.dto n h ugeto                          Any condition in the judgment of    1. Any condition in the judgment of  1. Any condition in the judgment of the Emergency Coordinator that           the Emergency Coordinator that       . n odto ntejdmn f                        the Emergency Coordinator that          the Emergency Coordinator that        the Emergency Coordinator that EC     indicates loss of the fuel clad          indicates potential loss of the         the Emergency Coordinator that             indicates potential loss of the RCS     indicates loss of the Containment     indicates potential loss of the Judgment          barrier                                  Fuel Clad Banrier                        indicates loss of the RCS Barrder         Barrier                                 Barrier                               Containment Barrier I[DOcument NO.]                                                                                             rRev.                      0                                                                                      Page 191 of 240

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier:              Fuel Clad Category:             A. RCS or SG Tube Leakage Degradation Threat:   Loss Threshold:
None I[Document No.]                  I        Rev. 0          IPage              192 of 240I

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier:                     Fuel Clad Category:                    A. RCS or SG Tube Leakage Degradation Threat:          Potential Loss Threshold:
: 1. RVLS < 0" (Note 9)I Note 9: RVLS is not valid if EITHER of the following exists:
        - One or more RCPs are running OR
        - LPI pump(s) are running AND taking suction from the LPI drop line Definition(s):
None Basis:
RVLS indicated level < 0" with all RCPs not running and both LPI pumps taking suction from the drop line not running represents reactor vessel level below the bottom of the RCS hotleg (without instrument uncertainty considered). This is the lowest measurable reactor vessel level and is used in lieu of actual reactor vessel level indication of level at or below top of active fuel.
This reading indicates a reduction in reactor vessel water level sufficient to allow the onset of heat-induced cladding damage.
ONS Basis Reference(s):
: 1. OSC-2820 Emergency Procedure Setpoints, Setpoint No. 7.1.6.5
: 2. NEI 99-01 RCS or SG Tube Leakage Potential Loss 1I.A
[Document No.]                                      Rev. 0                       Page 193 of 240

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier:                Fuel Clad Category:               B. Inadequate Heat Removal Degradation Threat:     Loss Threshold:
1l. CETCs > 1200&deg;F Definition(s):
iNone Basis:
CETCs > 1200&deg;F indicates extreme ICC conditions that may result in at least 516&deg;F of superheat.
This reading indicates temperatures within the core are sufficient to cause significant superheating of reactor coolant.
ONS Basis Reference(s):
: 1. OSC-2820 Emergency Procedure Setpoints, Setpoint No. 7.1 .1.7
: 2. NEI 99-01 Inadequate Heat Removal Loss 2.A S[Document No.]                             Rev. 0          {Page                    194 of 240

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier:                Fuel Clad Category:               B. Inadequate Heat Removal Degradation Threat:     Potential Loss Threshold:
: 1. CET3s > 700&deg;F Definition(s):
None                                                          i Basis:
CETCs > 7000&deg;F indicates conditions that may result in at least -1 6&deg;F of superheat and that may indicate core uncovery.
This reading indicates a reduction in reactor vessel water level sufficient to allow the onset of heat-induced cladding damage.
ONS Basis Reference(s):
: 1. OSC-2820 Emergency Procedure Setpoints, Setpoint No. 7.1.1.6
: 2. NEI 99-01 Inadequate Heat Removal Potential Loss 2.A I[Document No.] "                            Rev. 0                             Page 195 of 240

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier:                  Fuel Clad Category:                 B. Inadequate Heat Removal Degradation Threat:       Potential Loss Threshold:
: 2. RCS heat removal cannot be established AND RCS subcooling < 00 F Definition(s):
None Basis:
In combination with RCS Potential Loss B.1, meeting this threshold results in a Site Area Emergency.
The combination of these conditions indicates the ultimate heat sink function is under extreme challenge (i.e., superheated). This threshold addresses loss of functions required for hot shutdown with the reactor at pressure and temperature and thus a potential loss of the Fuel Clad Barrier (ref. 1).
This condition indicates an extreme challenge to the ability to remove RCS heat using the steam generators (i.e., loss of an effective secondary-side heat sink). This condition represents a potential loss of the Fuel Clad Barrier. In accordance with EOPs, there may be unusual accident conditions during which operators intentionally reduce the heat removal capability of the steam generators; during these conditions, classification using threshold is not warranted.
ONS Basis Reference(s):
: 1. OSC-2820 Emergency Procedure.Setpoints, Setpoint No. 7.1.3.1
: 2. NEI 99-01 Inadequate Heat Removal Potential Loss 2.B S[Document No.]                                Rev. 0                          Page 196 of 240

                                          *ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier:                   Fuel Clad Category:                  C. CMT Radiation / RCS Activity Degradation Threat:        Loss Threshold:
Table F-2       Containment Radiation - R/hr (1/2/3RIA 57/58,)
TieAfe SDFC             Loss          CMT Potential Loss (Hrs)           RIA 57      RIA 58      RIA 57      RIA 58 0O-<O0.5           300          140        1500        700 0.5 - <2.0           80          40         400         195 2.0O- <8.0           32           15         160         75
                        > 8.0             10            5          50         25 Definition(s):
None Basis:
The specified containment radiation monitor readings (ref. 1) indicate the release of reactor coolant, with elevated activity indicative of fuel damage, into the Containment. The readings are derived assuming the instantaneous release and dispersal of the reactor coolant noble gas and iodine inventory associated with approximately 4% fuel cladding failure 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. This value is higher than that specified for RCS barrier Loss #3.
Monitors used for this fission product barrier loss threshold are the Containment High Range Radiation Monitors RIA 57 and RIA 58 (ref. 1).
The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the containment, assuming that reactor coolant activity equals 300 pCi/gm dose equivalent I-1 31. Reactor coolant activity above this level is greater than that expected for iodine spikes and corresponds to an approximate range of 2% to 5% fuel clad damage. Since this condition indicates that a significant amount of fuel clad damage has occurred, it represents a loss of the Fuel Clad Barrier.
I[Document No.]                                 Rev. 0                           Page 197 of 240

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases The radiation monitor reading in this threshold is higher than that specified for RCS Barrier Loss threshold C.1 since it indicates a loss of both the Fuel Clad Barrier and the RCS Barrier.
Note that a combination of the two monitor readings appropriately escalates the ECL to a Site Area Emergency.
ONS Basis Reference(s):
: 1. OSC-5283 oNs Core Damage Assessment Guidelines, Rev. 2, 2/27/12
: 2. NEI 99-01 CMT Radiation / RCS Activity FC Loss 3.A I [Document    No.]                           *Rev.00                            Page 198 off 240 II4
[DI   metNo]Pae19

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier:                  Fuel Clad Category:                 C. CMT Radiation / RCS Activity Degradation Threat:       Loss Threshold:
S2. Coolant atvy> 300 p~/IDEl Definition(s):
None Basis:
Elevated reactor coolant activity represents a potential degradation in the level of safety of the plant and a potential precursor of more serious problems. The threshold Dose Equivalent 1-131 (DEl) concentration is well above that expected for iodine spikes and corresponds to about 2%
to 5% fuel clad damage. When reactor coolant activity reaches this level the Fuel Clad Barrier is considered lost (ref. 1).
This threshold indicates that RCS radioactivity concentration is greater than 300 pCi/gm dose equivalent I-131. Reactor coolant activity above this level is greater than that expected for iodine spikes and corresponds to an approximate range of 2% to 5% fuel clad damage. Since this condition indicates that a significant amount of fuel clad damage has occurred, it represents a loss of the Fuel Clad Barrier.
It is recognized that sample collection and analysis of reactor coolant with highly elevated activity levels could require several hours to complete. Nonetheless, a sample-related threshold is included as a backup to other indications.
There is no Potential Loss threshold associated with RCS Activity / Containment Radiation.
ONS Basis Reference(s):
: 1. NEI 99-01 CMT Radiation / RCS Activity Fuel Clad Loss 3.B l[Document No.]                      I         Rev. 0
* IPage                  199 of 240

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier:              Fuel Clad Category:             C. CMT Radiation / RCS Activity Degradation Threat:   Potential Loss Threshold:
SNoneI
*[Document No.]                           Rev. 0 *Page                       200 of 240

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier:              Fuel Clad Category:             D. CMT Integrity or Bypass Degradation Threat:   Loss Threshold:
None S[Document No.]                           Rev. 0                          Page 201 of 240

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier:              Fuel Clad Category:             D. CMT Integrity or Bypass Degradation Threat:   Potential Loss Threshold:
None S[Document No.]                  [Rev.         0                          Page 202 of 240

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier:                   Fuel Clad Category:                  E. Emergency Coordinator Judgment Degradation Threat:        Loss Threshold:
: 1. Any condition in the judgment of the Emergency Coordinator that indicates loss of the Fuel Clad Barrier Definition(s):   I None Basis:
The Emergency Coordinator judgment threshold addresses any other factors relevant to determining if the Fuel Clad Barrier is lost. Such a determination should include imminent barrier degradation, barrier monitoring capability and dominant accident sequences.
* Imminent barrier degradation exists if the degradation will likely occur within relatively short period of time based on a projection of current safety system performance. The term "imminent" refers to recognition of the inability to reach safety function acceptance criteria before completion of all checks.
* Barrier monitoring capability is decreased if there is a loss or lack of reliable indicators.
This assessment should include instrumentation operability concerns, readings from portable instrumentation and consideration of offsite monitoring results.
* Dominant ac~cident sequences lead to degradation of all fission product barriers and likely entry to the EOPs. The Emergency Coordinator should be mindful of the Loss of AC power (Station Blackout) and ATWS EALs to assure timely emergency classification declarations.
This threshold addresses any other factors that are to be used by the Emergency Coordinator in determining whether the Fuel Clad barrier is lost ONS Basis Reference(s):
: 1. NEI 99-01 Emergency Director Judgment Fuel Clad Loss 6.A
[Document No.]                       I        Rev. 0                             Page 203 of 240

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier:                   Fuel Clad Category:                  E. Emergency Coordinator Judgment Degradation Threat:        Potential Loss Threshold:
: 1. Any condition in the1 judgment of the Emergency Coordinator that indicates potential loss of the Fuel Clad Barrier Basis:
The Emergency Coordinator judgment threshold addresses any other factors relevant to determining if the Fuel Clad Barrier is potentially lost. Such a determination should include imminent barrier degradation, barrier monitoring capability and dominant accident sequences.
* Imminent barrier deqradation exists if the degradation will likely occur within relatively short period of time based on a projection of current safety system performance. The term "imminent", refers to recognition of the inability to reach safety function acceptance criteria before completion of all checks.
* Barrier monitorinq capability is decreased if there is a loss or lack of reliable indicators.
This assessment should include instrumentation operability concerns, readings from portable instrumentation and consideration of offsite monitoring results.
* Dominant accident sequences lead to degradation of all fission product barriers and likely entry to the EOPs. The Emergency Coordinator should be mindful of the Loss of AC power (Station Blackout) and ATWS EALs to assure timely emergency classification declarations.
This threshold addresses any other factors that are to be used by the Emergency Coordinator in determining whether the Fuel Clad barrier is potentially lost. The Emergency Coordinator should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored.
ONS Basis Reference(s):
: 1. NEI 99-01 Emergency Director Judgment Potential Fuel Clad Loss 6.A I[Document No.]                                Rev. 0             IPage                  204* of 240r

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier:                   Reactor Coolant System Category:                  A. RCS or SG Tube Leakage Degradation Threat:       Loss Threshold:
: 1. An automatic or manual ES actuation required by EITHER:
* UNISOLABLE RCS leakage
* SG tube RUPTURE Definition(s):
UNISOLABLE - An open or breached system line that cannot be isolated, remotely or locally.
RUPTURE - The condition of a steam generator in which primary-to-secondary leakage is of sufficient magnitude to require a safety injection.
Basis:
A RCS leak should be considered UNISOLABLE if the leak cannot be isolated within 15 minutes This threshold is based on an UNISOLABLE ROS leak of sufficient size to require an automatic or manual actuation of the Emergency Core Cooling System (ECCS). This condition clearly represents a loss of the RCS Barrier.
This threshold is applicable to unidentified and pressure boundary leakage, as well as identified leakage. It is also applicable to UNISOLABLE RCS leakage through an interfacing system. The mass loss may be into any location      - inside containment, to the secondary-side (i.e., steam generator tube leakage) or outside of containment.
A steam generator with primary-to-secondary leakage of sufficient magnitude to require a
.safety injection is considered to be RUPTURED. If a. RUPTURED steam generator is also FAULTED outside of containment, the declaration escalates to a Site Area Emergency since the Containment Barrier Loss threshold I .A will also be met.
ONS Basis Reference(s):
: 1. UFSAR Section 7.3 Engineered Safeguards Protective System
: 2. NEI 99-0 1 RCS or SG Tube Leakage Reactor Coolant System Loss 1.A S[Document No.]                               Rev. 0                           Page 205 of 240

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier:                  Reactor Coolant System Category:                 A. RCS or SG Tube Leakage Degradation Threat:       Potential Loss Threshold:
: 1. RCS leakage > normal makeup capacity due to EITHER:
* UNISOLABLE RCS leakage/
* SG tube leakage Definition(s):
UNISOLABLE - An open or breached system line that cannot be isolated, remotely or locally.
Basis:
A RCS leak should be considered UNISOLABLE if the leak cannot be isolated within 15 min.
This threshold is based on the inability to maintain liquid inventory within the RCS by normal operation of the High Pressure Injection System (HPI). The HPI includes three pumps. (ref. 1)
Any one HPI pump runout flow rate is 475 gpm (ref. 2).
This threshold is based on an UNISOLABLE RCS leak that results in the inability to maintain pressurizer level within specified limits by operation of a normally used charging (makeup) pump, but an ES actuation has not occurred. The threshold is met when an operating procedure, or operating crew supervision, directs that a HPI (makeup) pump be placed in service to restore and maintain pressurizer level.
This threshold is applicable to unidentified and pressure boundary leakage, as well as identified leakage. It is also applicable to UNISOLABLE RCS leakage through an interfacing system. The mass loss may be into any location       - inside containment, to the secondary-side (i.e., steam generator tube leakage) or outside of containment.
If a leaking steam generator is also FAULTED outside of containment, the declaration escalates to a Site Area Emergency since the Containment Barrier Loss threshold 1 .A will also be met.
ONS Basis Reference(s):
: 1. UFSAR Section 9.3.2 High Pressure Injection System
: 2. OSC-2820 Emergency Procedure Setpoints, Setpoint No. 7.3.1.2
: 3. NEI 99-01 RCS or SG Tube Leakage Reactor Coolant System Potential Loss 1.A Rev. 0                           Page 206 of 240
[Document
[Document No.]                                Rev. 0 Page 206 of 240 I

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier:                  Reactor Coolant System Category:                A. RCS or SG Tube Leakage Degradation Threat:       Potential Loss Threshold:
: 2. RCS cooldown to    < 400&deg; F at > 1000&deg;F/hr OR HPI has operated in the injection mode with no RCPs operating Definition(s):
None Basis:
4000 F is the temperature below which a cooldown greater than 100&deg;F/hr requires implementation of Pressurized Thermal Shock (PTS) guidance (rule 8) (ref. 1, 2).
HPI operating in the injection mode with no RCPs operating also invokes Rule 8 (ref. 3).
This condition indicates an extreme challenge to the integrity of the RCS pressure boundary due to pressurized thermal shock - a transient that causes rapid RCS cooldown while the ROS is in Mode 3 orjhigher (i.e., hot and pressurized).
ONS Basis Reference(s):
: 1. OSC-2820 Emergency Procedure Setpoints, Setpoint No. 7.1.2.7
: 2. OSC-2820 Emergency Procedure Setpoints, Setpoint No. 7.1.8.7
: 3. EP/*/AI1800/001 Rule 8 Pressurized Thermal Shock (PTS)
: 4. NEI 99-0 1 RCS or SG Tube Leakage Reactor Coolant System Potential Loss I .B S[Document No.]                                Rev. 0                        Page 207 of 240

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier:               Reactor Coolant System Category:              B. Inadequate Heat Removal Degradation Threat:    Loss Threshold:
None S[Document No.]                          'Rev. 0                          Page 208 of 240I

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier:                  Reactor Coolant System Category:                 B. Inadequate Heat Removal Degradation Threat:       Potential Loss Threshold:
: 1. ROS heat removal cannot be established
          *AND RCS subcooling < 0&deg;F Definition(s):
None Basis:
In combination with FC Potential Loss B.1, meeting this threshold results in a Site Area Emergency.
The combination of these conditions indicates the ultimate heat sink function is under extreme challenge (i.e., superheated). This threshold addresses loss of functions required for hot shutdown with the reactor at pressure and temperature and thus a potential loss of the RCS Barrier.
This condition indicates an extreme challenge to the ability to remove RCS heat using the steam generators (i.e., loss of an effective secondary-side heat sink). This condition represents a potential loss of the RCS Barrier. In accordance with EOPs, there may be unusual accident conditions during which operators intentionally reduce the heat removal capability of the steam generators; during these conditions, classification using threshold is not warranted.
Meeting this threshold results in a Site Area Emergency because this threshold is identical to Fuel Clad Barrier Potential Loss threshold B.2; both will be met. This condition warrants a Site Area Emergency declaration because inadequate RCS heat removal may r~esult in fuel heat-up
 .sufficient to damage the cladding and increase RCS pressure to the point where mass will be lost from the system.
ONS Basis Reference(s):
: 1. OSC-2820 Emergency Procedure Setpoints, Setpoint No. 7.1.3.6
: 2. NEI 99-01 Inadequate Heat Removal RCS Loss 2.B I[Document No.]                      1         Rev. 0         IPage                  209 of 240i

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier:                Reactor Coolant System Category:               B. Inadequate Heat Removal Degradation Threat:     Potential Loss Threshold:
J2. PIforc~ed coolingintae Definition(s):
None Basis:
HPI Forced Cooling (Rule 4) is used when the SGs are not capable of heat removal and RCS pressure is greater than 2300 psig. A Pressurizer PORV is opened to relieve pressure until HPI cools the reactor (feed and bleed). (ref. 1)
ONS Basis Reference(s):
: 1. OSC-2820 Emergency Procedure Setpoints, Setpoint No. 7.1.4.16
: 2. NEI 99-01 Other Indications Potential Loss 5.A S[Document No.1                               Rev. 0                       .Page 210 of 240

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier:                  Reactor Coolant System Category:                 C. CMT Radiation!/RCS Activity Degradation Threat:       Loss Threshold:
: 1. Containment radiation:
* 1,3 RIA57/58 >1.0OR/hr
* 2R1A57>1.6R/hr
* 2RIA58>1.0R/hr Definition(s):
N/A Basis:
Containment radiation monitor readings greater than the specified values (ref. 1) indicate the release of reactor coolant to the Containment. The readings assume the instantaneous release and dispersal of the reactor coolant noble gas and iodine inventory associated with normal operating concentrations (i.e., within Technical Specifications) into the Containment atmosphere. Because of the very high fuel clad integrity, only small amounts of noble gases would be dissolved in the primary coolant.
Monitors used for this fission product barrier loss threshold are the Containment High Range Radiation Monitors RIA-57 and RIA-58. The difference in the threshold values is due to the relative strength of the detector check source which affects the background readings for the detector (the source for 2RIA-57 is stronger than that for the other detectors). (ref. 1)
The radiation monitor reading corresponds, to an instantaneous release of all reactor coolant mass into the containment, assuming that reactor coolant activity equals Technical Specification allowable limits. This value is lower than that specified for Fuel Clad Barrier Loss threshold C.1 since it indicates a loss of the RCS Barrier only.
There is no Potential Loss threshold associated with RCS Activity / Containment Radiation.
ONS Basis Reference(s):
: 1. OSC-4244 ONS High Range Containment Monitor Correlation Factors for RIA-57 and RIA-58
: 2. NEI 99-01 CMT Radiation / RCS Activity RCS Loss 3.A S[Document No.]                               Rev. 0                           Page 211 of 240I

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier:              Reactor Coolant System Category:             B. CMT Radiation!/RCS Activity Degradation Threat:   Potential Loss Threshold:
None
[Document No.]                           Rev. 0                          Page 212 of 240

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier:              Reactor Coolant System Category:             D. CMT Integrity or Bypass Degradation Threat:   Loss Threshold:
None I[Document No.]                 I         Rev.. 0         IPage              213 of 240

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier:              Reactor Coolant System Category:             D. CMT Integrity or Bypass Degradation Threat: Potential Loss Threshold:
NoneI S[Document No.]                           Rev. 0                          Page 214 of 240

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier:                  Reactor Coolant System Category:                 E. Emergency Coordinator Judgment Degradation Threat:       Loss Threshold:
Definition(s):
None Basis:
The Emergency Coordinator judgment threshold addresses any other factors relevant to determining if the RCS Barrier is lost. Sucha determination should include imminent barrier degradation, barrier monitoring capability and dominant accident sequences.
* Imminent barrier de qradation exists if the degradation will likely occur within relatively short period of time based on a projection of current safety system performance. The term "imminent" refers to recognition of the inability to reach safety function acceptance criteria before completion of all checks.
* Barrier monitorinq capability is decreased if there is a loss or lack of reliable indicators.
This assessment should include instrumentation operability concerns, readings from portable instrumentation and consideration of offsite monitoring results.
* Dominant accident sequences lead to degradation of all fission product barriers and likely entry to the EOPs. The Emergency Coordinator should be mindful of the Loss of AC power (Station Blackout) and ATWS EALs to assure timely emergency classification declarations.
This threshold addresses any other factors that may be used by the Emergency Coordinator in determining whether the RCS Barrier is lost.
ONS Basis Reference(s):
: 1. NEI 99-01 Emergency Director Judgment RCS Loss 6.A IDocument No.]                               Rev. 0                             Page 215 of 240

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier:                  Reactor Coolant System Category:                 E. Emergency Coordinator Judgment Degradation Threat:       Potential Loss Threshold:
: 1. Any corjdition in the judgment of the Emergency Coordinator that indicates potential loss of the RCS Barrier Definition(s):
None Basis:
The Emergency Coordinator judgment threshold addresses any other factors relevant to determining if the RCS Barrier is potentially lost. Such a determination should include imminent barrier degradation, barrier monitoring capability and dominant accident sequences.
* Imminent barrier degradation exists if the degradation will likely occur within relatively short period of time based on a projection of current safety system performance. The term "imminent" refers to recognition of the inability to reach safety function acceptance criteria before completion of all checks.
* Barrier monitorinq capability is decreased if there is a loss or lack of reliable indicators.
This assessment should include instrumentation operability concerns, readings from portable instrumentation and consideration of offsite monitoring results.
* Dominant accident sequences lead to degradation of all fission product barriers and likely entry to the EOPs. The Emergency Coordinator should be mindful of the Loss of AC power (Station Blackout) and ATWS EALs to assure timely emergency classification declarations.
This threshold addresses any other factors that may be used by the Emergency Coordinator in determining whether the RCS Barrier is potentially lost. The Emergency Director should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored.
ONS Basis Reference(s):
: 1. NEI 99-01 Emergency Director Judgment RCS Potential Loss 6,A
[Document No.]                      I        Rev. 0            IPage                  216 of 240

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier:                   Containment Category:                  A. RCS or SG Tube Leakage Degradation Threat:        Loss Threshold:
: 1. A leaking SG is FAULTED outside of containment Definition(s):
FAULTED - The term applied to a steam generator that has a steam leak on the secondary side of sufficient size to cause an uncontrolled drop in steam generator pressure or the steam generator to become completely depressurized.
Basis:
This threshold addresses a leaking Steam Generator (SG) that is also FAULTED outside of containment. The condition of the SG leakage, is determined in accordance with the thresholds for RCS Barrier Potential Loss A.1 and Loss A.1, respectively. This condition represents a bypass of the containment barrier.
FAULTED is a defined term within the NEI 99-01 methodology; this determination is not necessarily dependent upon entry into, or diagnostic steps within, an EOP. For example, if the pressure in a steam generator is decreasing uncontrollably (part of the FAULTED definition) and the FAULTED steam generator isolation procedure is not entered because EOP user rules are dictating implementation of another procedure to address a higher priority condition, the steam generator is still considered FAULTED for emergency classification purposes.
The FAULTED criterion establishes an appropriate lower bound on the size of a steam release that may require an emergency classification. Steam releases of this size are readily observable with normal Control Room indications. The lower bound for this aspect of the containment barrier is analogous to the lower bound criteria specified in IC SU4 for the fuel clad barrier (i.e., RCS activity values) and IC SU5 for the RCS barrier (i.e., RCS leak rate values).
This threshold also applies to prolonged steam releases necessitated by operational considerations such as the forced steaming of a leaking steam generator directly to atmosphere to cooldown the plant. These type of condition will result in a significant and sustained release of radioactive steam to the environment (and are thus similar to a FAULTED condition). The inability to isolate the steam flow without an adverse effect on plant cooldown meets the intent of a loss of containment.
Steam releases associated with the expected operation of a SG Atmospheric Dump Valve(s) do not meet the intent of this threshold. Such releases may occur intermittently for a short period of time following a reactor trip as operators process through emergency operating procedures to bring the plant to a stable condition and prepare to initiate a plant cooldown.
Steam releases associated with the unexpected operation of a valve (e.g., a stuck-open safety valve) do meet this threshold.
I[Document No.]                                Rev. 0                           Page 217 of 240I

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Following an SG tube leak, there may be minor radiological releases through a secondary-side system component (e.g., air ejectors, gland seal exhausters, valve packing, steam traps, terry turbine exhaust, etc.). These types of releases do not constitute a loss or potential loss of containment but should be evaluated using the Recognition Category R ICs.
The ECLs resulting from primary-to-secondary leakage, with or without a steam release from the FAULTED SG, are summarized below.
Affected SG is FAULTED Outside of Containment?-
P-to-S Leak Rate                        Yes                       No Less than or equal to 25 gpm             No classification       No classification Greater than 25 gpm                 Unusual Event per SU5.1  Unusual Event per SU5.1 Greater than normal makeup pump capacity (RCS Barrier          Site Area Emergency per PotentialLoss)                                FSI .1             Alert per FAI.1 Requires an automatic or manual Site Area Emergency per ECCS (ES) actuation (RCS                                         'Alert per FAI.1 FSI1.1 BarrierLoss)
There is no Potential Loss threshold associated with RCS or SG Tube Leakage.
ONS Basis Reference(s):
: 1. NEI 99-01 RCS or SG Tube Leakage Containment Loss I .A S[Document No.]                        IRev.            0                            Page 218 of 240

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier:              Containment Category:             A. RCS or SG Tube Leakage Degradation Threat:   Potential Loss Threshold:
Nonei I[Document No.]                           Rev. 0                          Page 219 of 240[

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier:              Containment Category:             B. Inadequate Heat Removal Degradation Threat:   Loss Threshold:
None I[Document No.]                           Rev. 0          IPage 220 of 240I

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier:                   Containment Category:                  B. Inadequate Heat Removal Degradation Threat:        Potential Loss Threshold:
: 1. CETCs >12000 F AND Restoration procedures not effective within 15 min. (Note 1)
Note 1:   The Emergency Coordinator should declare qthe event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
Definition(s):
None Basis:
Core Exit Thermocouples (CETCs) are a component of Inadequate Core Cooling Instrumentation and provide an indirect indication of fuel clad temperature by measuring the temperature of the reactor coolant that leaves the core region. Although clad rupture due to high temperature is not expected for CETC readings less than the threshold, temperatures of this magnitude signal significant superheating of the reactor coolant and core uncover (ref. 1).
The restoration procedures are those emergency operating procedures that address the recovery of the RCS and core heat removal acceptance criteria. The procedure is considered effective if the temperature is decreasing or if the vessel water level is increasing (ref. 1 ).The 15 minute threshold starts when operator action begins taking procedurally directed functional recovery actions.
If CETC readings are greater than I1,200&deg;F, Fuel Clad barrier is also lost.
This threshold addresses any other factors that may be used by the Emergency Coordinator in determining whether the Containment Barrier is potentially lost. The Emergency Director should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored.
ONS Basis Reference(s):
: 1. EP/1 ,2,3/A/1 800/00 1 Inadequate Core Cooling
: 2. NEI 99-01 Inadequate Heat Removal Containment Potential Loss 2.A EDocument No.]                                      Rev. 0          IPage                      221 of 240I

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier:              Containment Category:             C. CMT Radiation/RCS Activity Degradation Threat:   Loss Threshold:
None I[Document No.]                           Rev. 0          IPage              222 of 240I

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier:                     Containment Category:                    C. CMT Radiation/RCS Activity Degradation Threat:          Potential Loss Threshold:
I1. 1/2/3RIA 57/58    >  Table F-2 column "CTPtnilLoss"i Table F-2       Containment Radiation    - R/hr (112/3RIA 57/58)
TieAfe SDFC            Loss         CMT Potential Loss (Hrs)            RIA 57     RIA 58     RIA 57       RIA 58 0- <O.5             300         140       1500          700 0.5 - <2.0             80         40         400          195 2.0O- <8.0             32          15        160           75
                         > 8.0               10          5         50           25 Definition(s):
None Basis:
Containment radiation monitor readings greater than the values shown (ref. 1) indicate significant fuel damage well in excess of that required for loss of the RCS Barrier and the Fuel Clad Barrier.
The specified containment radiation monitor readings (ref. 1) indicate the release of reactor coolant, with significant fuel damage well in excess of that required for loss of the RCS Barrier and the Fuel Clad Barrier, into the Containment. The readings are derived assuming the instantaneous release and dispersal of the reactor coolant noble gas and iodine inventory associated with 20% clad failure into the Containment atmosphere.
Containment radiation readings at or above the Containment Barrier Potential Loss threshold signify a loss of two fission product barriers and Potential Loss of a third, indicating the need to upgrade the emergency classification to a General Emergency.
Monitors used for this fission product barrier loss threshold are the Containment High Range Radiation Monitors RIA-57 and RIA-58 (ref. 1).
I[Document No.]                                  Rev. 0                            Page 223 of 240

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the containment, assuming that 20% of the fuel cladding has failed. This level of fuel clad failure is well above that used to determine the analogous Fuel Clad Barrier Loss and RCS Barrier Loss thresholds.
NUREG-1 22*8, Source Estimations During Incident Response to Severe Nuclear Power Plant Accidents, indicates the fuel clad failure must be greater than approximately 20% in order for there to be a major release of radioactivity requiring offsite protective actions. For this condition to exist, there must already have been a loss of the ROS Barrier and the Fuel Clad Barrier. It is therefore prudent to treat this condition as a potential loss pf containment which would then escalate the ECL to a General Emergency.
ONS Basis Reference(s):
: 1. OSC-5283 ONS Core Damage Assessment Guidelines, Rev. 2, 2/27/12
: 2. NEI 99-01 CMT Radiation / RCS Activity Containment Potential Loss 3.A S[Document No.]                                  Rev. 0                            Page 224 of 240

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier:                   Containment Category:                  D. CMT Integrity or Bypass Degradation Threat:        Loss Threshold:
: 1. Containment isolation is required AND EITHER:
* Containment integrity has been lost based on Emergency Coordinator' judgment
* UNISOLABLE pathway from Containment to the environment exists Definition(s):
UNISOLABLE - An open or breached system line that cannot be isolated, remotely or locally.
Basis:
The pathway should be considered UNISOLABLE if the Containment cannot be isolated within 15 min.
Reactor Building Essential and Non-essential Isolation occurs on an Engineered Safeguards signal of 3 psig (ref. 1).
These thresholds address a situation where containment isolation is required and one of two conditions exists as discussed below. Users are reminded that there may be accident and release conditions that simultaneously meet both bulleted thresholds.
First Threshold   -  Containment integrity has been lost, i.e., the actual containment atmospheric leak rate likely exceeds that associated with allowable leakage (or sometimes referred to as design leakage). Following the release of RCS mass into containment, containment pressure will fluctuate based on a variety of factors; a loss of containment integrity condition may (or may not) be accompanied by a noticeable drop .in containment pressure. Recognizing the inherent difficulties in determining a containment leak rate during accident conditions, it is expected that the Emergency Coordinator will assess this threshold using judgment, and with due consideration given to current plant conditions, and available operational and radiological data (e.g., containment pressure, readings on radiation monitors outside containment, operating status of containment pressure control equipment, etc.).
Refer to the middle piping run of Figure 1. Two simplified examples are provided. One is leakage from a penetration and the other is leakage from an in-service system valve.
Depending upon radiation monitor locations and sensitivities, the leakage could be detected by any of the four monitors depicted in the figure.
Another example would be a loss or potential loss of the RCS barrier, and the simultaneous occurrence of two FAULTED locations on a steam generator where one fault is located inside containment (e.g., on a steam or feedwater line) and the other outside of containment. In this case, the associated steam line provides a pathway for the containment atmosphere to escape to an area outside the containment.
S[Document No.]                                 Rev. 0                            Page 225 of 240

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Following the leakage of RCS mass into containment and a rise in containment pressure, there may be minor radiological releases associated with allowable (design) containment leakage through various penetrations or system components. These releases do not constitute a loss or potential loss of containment but should be evaluated using the Recognition Category R l~s.
Second Threshold     -  Conditions are such that there is an UNISOLABLE pathway for the migration of radioactive material from the containment atmosphere to the environment. As used here, the term "environment" includes the atmosphere of a room or area, outside the containment, that may, in turn, com~municate with the outside-the-plant atmosphere (e.g.I, through discharge of a ventilation system or atmospheric leakage). Depending upon a variety of factors, this condition may or may not be accompanied by a noticeable drop in containment pressure.
Refer to the top piping run of Figure 1. In this simplified example, the inboard and outbdard isolation valves remained open after a containment isolation was required (i.e., containment isolation was not successful). There is now an UNISOLABLE pathway from the containment to the environment.
The existence of a filter is not considered in the threshold assessment. Filters do not remove fission product noble gases. In addition, a filter could become ineffective due to iodine and/or particulate loading beyond design limits (i.e., retention ability has been exceeded) or water saturation from steam/high humidity in the release stream.
Leakage between two interfacing liquid systems, by itself, does not meet this threshold.
*Refer to the bottom piping run of Figure 1. In this simplified example, leakage in an RCP seal cooler is allowing radioactive material to enter the Auxiliary Building. The radioactivity would be detected by the Process Monitor. If there is no leakage from the closed water cooling system to the Auxiliary Building, then no threshold has been met. If the pump developed a leak that allowed steam/water to enter the Auxiliary Building, then second threshold would be met. Depending upon radiation monitor locations and sensitivities, this leakage could be detected by any of the four monitors depicted in the figure and cause the first threshold to be met as well.
Fol~lowing the leakage of RCS mass into containment and a rise in containment pressure, there may be minor radiological releases associated with allowable containment leakage through various penetrations or system components. Minor releases may also occur if a containment isolation valve(s) fails to close but the containment atmosphere escapes to an enclosed system. These releases do not constitute a loss or potential loss of containment but should be evaluated using the Recognition Category R ICs.
The status of the containment barrier during an event involving steam generator tube leakage is assessed using Loss Threshold A.1.
ONS Basis Reference(s):
: 1. UFSAR Section 6.2.3 Containment Isolation System
: 2. NEI 99-01 CMT Integrity or Bypass Containment Loss 4.A
[Document No.]                                 Rev. 0                          Page 226 of 240

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier:                  Containment Category:                 D. CMT Integrity or Bypass Degradation Threat:       Loss Threshold:
: 2. Indications of RCS leakage outside of q3ontainment                                     1 Definition(s):
None Basis:
Containment sump, temperature, pressure and/or radiation levels will increase if reactor coolant mass is leaking into the containment. If these parameters have not increased, then the reactor coolant mass may be leaking outside of containment (i.e., a containment bypass sequence). Increases in sump, temperature, pressure, flow and/or radiation level readings outside of the containment may indicate that the RCS mass is being lost outside of containment.
Unexpected elevated readings and alarms on radiation monitors with detectors outside containment should be corroborated with other available indications to confirm that the source is a loss of RCS mass outside of containment. If the fuel clad barrier has not been lost, radiation monitor readings outside of containment may not increase significantly; however, other unexpected changes in sump levels, area temperatures or pressures, flow rates, etc.
should be sufficient to determine if RCS mass is being lost outside of the containment.
Refer to the middle piping run of Figure 1. In this simplified example, a leak has occurred at a reducer on a pipe carrying reactor coolant in the Auxiliary Building. Depending upon radiation monitor locations and sensitivities, the leakage could be detected by any of the four monitors depicted in the figure and cause threshold D.1 to be met as well.
To ensure proper escalation of the emergency classification, the RCS leakage outside of containment must be related to the mass loss that is causing the RCS Loss and/or Potential Loss threshold A.1 to be met.
ONS Basis Reference(s):
: 1. NEI 99-01 CMT Integrity or Bypass Containment Loss
*.[Document No.]                              Rev. 0                           Page 227 of 240

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Figure 1: Containment Integrity or Bypass Examples Seal Cooling
{[Document No.]                             Rev. 0                           Page 228 of 240I

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier:                 Containment Category:                D. CMT Integrity or Bypass Degradation Threat:      Potential Loss Threshold:
: 1. Containment pressure > 59 psig Definition(s):
None Basis:
The Reactor Building is designed for an internal pressure of 59 psig (ref. 1).
If containment pressure exceeds the design pressure, there exists a potential to lose the Containment Barrier. To reach this level, there must be an inadequate core cooling condition for an extended period of time; therefore, the RCS and Fuel Clad barriers would already be lost. Thus, this threshold is a discriminator between a Site Area Emergency and General Emergency since there is now a potential to lose the third barrier.
ONS Basis Reference(s):
: 1. UFSAR Section 6.2.1 Containment Functional Design
: 2. NEI 99-01 CMT Integrity or Bypass Containment Potential Loss 4.A S[Document No.]                               Rev. 0                           Page 229 of 240I

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier:                   Containment Category:                  D. CMT Integrity or Bypass Degradation Threat:        Potential Loss Threshold:
: 2. Containment hydrogen concentration > 4%
Definition(s):
None Basis:
Following a design basis accident, hydrogen gas may be generated inside the containment by reactions such as zirconium metal with water, corrosion of materials of construction and radiolysis of aqueous solution in the core and sump.
The 4% hydrogen concentration threshold is generally considered the lower limit for hydrogen deflagrations. ONS is equipped with a Containment Hydrogen Monitoring System (CHMS) that provides continuous indication of hydrogen concentration in the containment atmosphere. The measurement capability is provided over the range of 0% to 10%. A continuous indication of the hydrogen concentration is not required in the control room at all times during normal operation. If continuous indication of the hydrogen concentration is not available at all times, continuous indication and recording shall be functioning within 90 minutes of the initiation of the safety injection. (ref. 1, 2)
The existence of an explosive mixture means, at a minimum, that the containment atmospheric hydrogen concentration is sufficient to support a hydrogen burn (i.e., at the lower deflagration limit). A hydrogen burn will raise containment pressure and could result in collateral equipment damage leading to a-loss of containment integrity. It therefore represents a potential loss of the Containment Barrier.
ONS Basis Reference(s):
: 1. UFSAR Section 9.3.7 Containment Hydrogen Monitoring System
: 2. UFSAR Section 15.16.3 Evaluation of Hydrogen Concentrations
: 3. NEI 99-01 CMT Integrity or Bypass Containment Potential Loss 4.B I[Document No.]                        I        Rev. 0        IPage                  230 of 240

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier:                    Containment Category:                   D. CMT Integrity or Bypass Degradation Threat:         Potential Loss Threshold:
: 3. Containment pressure > 10 psig with < one full train of containment heat removal Isystem (1 RBS with > 700 gpm spray flow OR 2 R'BCUs) operating per design for >-
15 min. (Note 1)
Note 1: The Emergency Coordinator should declare the event promptly upon 'determining that time limit has S been exceeded, or will likely be exceeded.
Definition(s):
None Basis:
Two engineered safeguards systems, the Reactor Building Spray System and the Reactor Building Cooling System, are provided to remove heat from the containment atmosphere following an accident. Both the Reactor Building Spray System and the Reactor Building Cooling System, with either at full capacity, are individually capable of maintaining the containment pressure below the design limit following a LOCA or MS LB. (ref. 1, 3)
        *The Reactor Building Spray (RBS) System consists of two separate trains of equal capacity. Spray flow greater or equal to 700 gpm satisfies the spray flow design requirement. The Reactor Building pressure setpoint (10 psig) is the pressure at which the Reactor Building Spray equipment should actuate and begin performing its function (ref. 1,2, 3, 4).
     ,,   Each of three Reactor Building Cooling Units (RBCUs) consists of a fan, cooling coils, and the required distribution duct work. The Reactor Building atmosphere is circulated past cooling coils by fans and returned to the building. Cooling water-for the cooling units is supplied by the Low Pressure Service Water System. The Reactor Building Cooling System provides the design heat removal capacity with two of three coolers operating (ref. 1).
This threshold describes a condition where containment pressure is greater than the setpoint at which containment energy (heat) removal systems are designed to automatically actuate, and less than one full train of equipment is capable of operating per design. The 15-minute criterion is included to allow operators time to manually start equipment that may not have automatically started, if possible. This threshold represents a potential loss of containment in that containment heat removal/depressurization systems (e.g., containment sprays, ice condenser fans, etc., but not including containment venting strategies) are either lost or performing in a degraded manner.
S[Document. No.]                                     Rev. 0                          Page 231 of 240

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases ONS Basis Reference(s):
: 1. UFSAR Section 6.2.2 Containment Heat Removal Systems
: 2. UFSAR Table 7-2 Engineered Safeguards Actuation Conditions
: 3. UFSAR Table 6-25 Minimum Acceptable Combinations of Containment Heat Removal Equipment Performance
: 4. OSC-02820 Emergency Procedure Setpoints, Setpoint No. 7.4.1.2
: 5. NEI 99-01 CMT Integrity or Bypass Containment Potential Loss 4.C S[Document No.]                            Rev. 0                         Page 232 of 240

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier:                   Containment Category:                  E. Emergency Coordinator Judgment Degradation Threat:        Loss Threshold:
: 1. Any condition in the judgment of the Emergency Coordinator that indicates loss of the Containment B~rrier Definition(s):
None Basis:
The Emergency Coordinator judgment threshold addresses any other factors relevant to determining if the Containment Bar'rier is lost. Such a determination should include imminent barrier degradation, barrier monitoring capability and dominant accident sequences.
* Imminent barrier degradation exists if the degradation will likely occur within relatively short period of time based on a projection of current safety system performance. The term "imminent" refers to recognition of the inability to reach safety function acceptance criteria before completion of all checks.
* Barrier monitoring capability is decreased if there is a loss or lack of reliable indicators.
This assessment should include instrumentation operability concerns, readings from portable instrumentation and consideration of offsite monitoring results.
* Dominant accident seqiuences lead to degradation of all fission product barriers and likely entry to the EOPs. The Emergency Coordinator should be mindful of the Loss of AC power (Station Blackout) and ATWS EALs to assure timely emergency classification declarations.
This threshold addresses any other factors that may be used by the Emergency Coordinator in determining whether the Containment Barrier is lost.
ONS Basis Reference(s):
: 1. NEI 99-01 Emergency Director Judgment PC Loss 6.A I[Document No.]                                 Rev. 0          IPage                  23of 240I

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier:                  Containment Category:                 E. Emergency Coordinator Judgment Degradation Threat:       Potential Loss Threshold:
1.I nny condition in the judgment of the Emergency Coor~dinator that indicates potential loss of the Containment Barrier Definition(s):
None Basis:
The Emergency Coordinator judgment threshold addresses any other factors relevant to determining if the Containment Barrier is potentially lost. Such a determination should include imminent barrier degradation, barrier monitoring capability and dominant accident sequences.
* Imminent barrier decqradation exists if the degradation will likely occur within relatively short period of time based on a projection of current safety system performance. The term "imminent" refers to recognition of the inability to reach safety function acceptance criteria before completion of all checks.
* Barrier monitorincq capability is decreased if there is a loss or lack of reliable indicators.
This assessment should include instrumentation operability concerns, readings from portable instrumentation and consideration of offsite monitoring results.
* Dominant accident sequences lead to degradation of all fission product barriers and likely entry to the EOPs. The Emergency Coordinator should be mindful Of the Loss of AC power (Station Blackout) and ATWS EALs to assure timely emergency classification declarations.
This threshold addresses any other factors that may be used by the Emergency Coordinator in determining whether the Containment Barrier is lost.
ONS Basis Reference(s):
: 1. NEI 99-01 Emergency Director Judgment PC Potential Loss 6.A S[Document No.]                               Rev. 0                            Page 234 of 240

ATTACHMENT 3 Safe Operation & Shutdown Rooms/Areas Tables R-2 & H-2 Bases

===Background===
NEI 99-01 Revision 6 l~s AA3 and HA5 prescribe declaration of an Alert based on impeded access to rooms or areas (due to either area radiation levels or hazardous gas concentrations) where equipment necessary for normal plant operations, cooldown or shutdown is located.
These areas are intended to be plant operating mode dependent. Specifically the Developers Notes for AA3 and HA5 states:
The "site-qpecificlist of plant rooms or areas with entry-relatedmnode applicabilityidentified" should specify those rooms or areas that contain equipment which require a manual/local action as specified in operatingproceduresused for normal plant operation, cooldown and shutdown. Do not include rooms or areas in which actions of a contingent or emergency nature would be performed (e.g., an action to address an off-normal or emergency condition such as emergency repairs, corrective measures or emergency operations). In addition, the list should specify the plant mode(s) during which entry would be requiredfor each room or area.
The list should not include rooms or areas for which entry is required solely to perform actions of an administrativeor recordkeeping nature (e.g., normal rounds or routine inspections).
Further, as specified in IC HA5:
The list need not include the Contfrol Room if adequate engineered safety/design features are in place to preclude a Control Room evacuation due to the release of a hazardousgas.
Such features may include, but are not limited to, capability to draw air from multiple air intakes at different and separate locations, inner and outer atmospheric boundaries, or the capabilityto acquire and maintain positive pressure within the Control Room envelope.
[Document No.]                               Rev. 0                            Page 235 of 240r

ATTACHMENT 3 Safe Operation & Shutdown Rooms/Areas Tables R-2 & H-2 Bases ONS Table R-2 and H-2 Bases NEI 99-01 Rev 06 addresses elevated radiation levels and hazardous gases in certain plant rooms/areas sufficient to preclude or impede personnel from performing actions necessary to maintain normal plant operation, or to perform a normal plant shutdown and cool down.
Power Operation was reviewed to determine if any actions are "necessary" to maintain power operations. Over reasonable periods (several days), there are some actions outside the Control Room that are required to be performed to maintain normal operations. The following table lists the locations into which an operator may be dispatched in order perform a normal plant operation, shutc(lown and cool down.I The review was completed using the following procedures as the controlling documents:
* OP/*/A/1102/010 (Controlling Procedure for Unit Shutdown)
* OP/*IAj1 106/001' (Turbine Generator)
* OP/*/A/1106/015 (EHC System)
* OP/*/A~/1 103/004A (RCS Boration)
* OP/*/A/1104/027 (Bleed Transfer Pump Recirculation)
* PTI*IA106001001 B (Surveillance to go to Mode 3)
* OPP *Ill1021010 (Unit SD Mode I to Mode 3)
* IPI*/A/0200104,7 (LTOP Calibration)
* OP/*/AI1 103/006 (RCP Operations)
* OP/*IA.I1 104/012 (CCW Pump Operations)
* CPI11AI20021014 (RCS Sampling)
* OPI/*IN1104/049 (LTOP Operation)
* OP/11AI1104/001 (Core Flood Operations)
* OP/OIAI1 104/048 (TBS Operations)
* OPI*NIA 104/004 (Low Pressure Injection System)
* OPI*IAI1 103/008 (RCS Crud Burst)
Travel paths to the locations where the equipment is operated were considered as part of the determination of affected rooms. ONS Reactor and Auxiliary Building design consist of mostly single entry rooms located off of a common hallway, therefore access to the hallway is required to access a given room. Some equipment is located within the hallway itself.
Room                  Mode          Procedure      Enclosure       Steps TB                                     I      OP/1/A/1102/010       4.1    Unit SD TB                                     1      OP/1/A/1106/001       4.2    TG TB                                     1      OP/llAI1106/014       4.3    MSRH TB                                     I      OP/1lA/1106/015       4.2    EHC A-2 LOST Hatch area                  1,2,3    OP/llAI1103/004 A     4.1    RCS Boration A-i hallway 8 S/col 65                1,2,3    OP/1lAI1103/004 A     4.2    RCS Boration A-i hallway 8' S/ col 65              1,2,3    OP/I/AI1103/004 A     4.3    RCS Boration Unit I BTP Rm                        1,2,3    OP/llAI1103/004 A     4.3    RCS Boration S[Document No.]                                     Rev. 0                          Page 236 of 240

ATTACHMENT 3 Safe Operation & Shutdown Rooms/Areas Tables R-2 & H-2 Bases Unit 1 BTP Rm                    1,2,3    OP/IA1A1103/OO4A       4.4    RCS     Boration A-i hallway 8' S/cal 65          1,2,3    OP/1/A/1103/004 A      4.4    RCS     Boration A-2-Unit 2 LDST Hatch area       1,2,3    OP/1/A/1103/004A       4.4    RCS     Boration Unit 2 BTP Rm                    1,2,3    OP/i/Ni1103/004 A      4.4    RCS     Boration A-I-hallway N of Cal 82          1,2,3    OP/i/A/1103/004 A      4.4    RCS     Boration A-I hallway 5' S/col 67          1,2,3    OPIll/N1103/004 A      4.4    RCS     Boration A-I hallway cal 82               1,2,3    OP/1/A/1103/004 A      4.4    RCS     Boratian A-2 Unit 3 LDST Hatch area       1,2,3    OPI/i/Ni103/004 A      4.5    RCS     Boration Unit 3 BTP Rm                    1,2,3    OP/1/A/1103/004 A      4.5    RCS     Boratian I A-I hallway 10' S/cal 96)        1,2,3    OPI/i/N1103/004 A      4.5    RCS     Boratian A-I hallway 5' S/cal 67          1,2,3    OP/il/N1103/004 A      4.5    RCS Baratian LPI Caoler Rm 1'W/North daor     1,2,3    OP/il/Ni103/004 A      4.5    RCS Boratian A-i- BAMT Rm                       1      OP/i/Ni1103/004 A      4.6    RCS Boration A-i Unit 1 & 2 BAMT Rm             1      OP/i/Ni1103/004 A      4.6    RCS Boration Rm 111                             I      OP1/i/Ni103/004 A      4.6    RCS Boration A-2 LDST Hatch area                1      OP/i/Ni1103/004 A !. 4.6    RCS Boratian CTT Rm                             I      OP/I/A/1103/004 A      4.6    RCS     Boration A-2-1i&2 Chem. Add Panel           1      OP/I/A/1103/004 A      4.6    RCS     Boration A-I-Cal Q70                        1      OP/i/Ni1103/004 A      4.7    RCS     Boration A-2-LDST Hatch area                I      OP/i/Ni1103/004 A      4.7    RCS     Boration A-i-Unit I CBAST Rm                1      OP/I/NI1103/004 A      4.7    RCS Boration Unit I BTP Rm                      1      OP/1/AI1104/027       4.19    BTP Recirc Unit I BTP Rm                      1      OP/1/A/1104/027       4.20. BTP Recirc 1      OP/IllA1102/010        4.2    Unit SD 1      PT/1/AN0600/001       13.2    Surv. Mode3 Unit l-BTP Rm                     1,2     OP1/A/N1103/004        4.5    Makeup Unit l-BTP Rm                     1,2     OP/ll/I1103/004        4.6    Makeup 1,2,3    OP/l/A/1102/010        4.3      SD Mode I to 3 3      OP/I/A/1102/010        4.4 RB 779', Cable Roam, 1UB2          3      IP/11/N0200/047               LTOP Calibration RB 779'                            3      IP/11AI0200/047               LTOP Calibration 1UB2                               3      IP/11A!02001047               LTOP Calibration 1AT7                               3      IP/1 /A0200/047               LTOP Calibration 1MTC-4                             3      IPI1/N0200/047                LTOP Calibration 1AT5                               3      IP/l/A/0200/047               LTOP Calibration LPI Coaler Room                    3      OP/I/A/1103/006       4.12 3      OP/1/A/1102/010        4.7   ________
OP/i/NI1104/012 A      4.2         CCW Pump 3  OP/I/A/1102/0I0        4.7 Unit 1 Primary Sample Hood                CP/1/A/2002/014        4.2 AB SAMPLE RM.308                          CP/11AN2002/014        4.2     _______
A-4-402 PZR Heaters                3      OP/I/Ni1102/010        4.7
[Document No.]                                   Rev. 0                           Page 237 of 240

ATTACHMENT 3 Safe Operation & Shutdown  Rooms/Areas Tables R-2 & H-2 Bases Equip Rrm XO/XP                    3    OP/1/A/1102/010          4.7 Equip Rm XO/XP                     3    OP/1/A/1104/049          4.2 A-4-402 PZR Heaters                3    OP/iI/A/i1104/049        4.2 3    OP/1/A/1104/001         4.14 A-4-409                            3    OP/1/A/1104/001          4.4 A-3-308                            3    OP/1/A/1104/001          4.4 A-2 Hallway                        3    OP/I/AI1104/001          4.4 R-1G-W                             3    OP/1/A/1104/001          4.4 R-1 -around "A" CFT                3  IOP/I/A/1104/001           4.4 R-B above Emer Sump                3    OP/i/Ni1104/001          4.4 R-B above RBNS                     3    OP/1/A/1104/001          4.4 _______
R-l-around "B" CFT                 3    OP/1/A/1104/001          4.4 R-B-20' above LD CIr RM            3    OP/i/Ni1104/001          4.4 Equip Rm XO/XP                     3    OP/1/N/1104/001         4.14 A-4-W Pent Rm                      3    OP/I/Ni1104/001         4.14 A-4-E Pent                         3    OP/i/Ni1104/049          4.2 R-3G East Side                     3    OP/i/Ni1104/049          4.2 A-4-402                            3    OP/i/Ni1104/049          4.2 LTOP Alignment A-2-Col. P-63)                     3    OP/i/Ni1104/049          4.2 LTOP Alignment T-3-Equip Rm)                      3    OP/i/Ni1104/049          4.2 LTOP Alignment 3    OPI1/A/1102/010          4.7 3    OP/11N1102/010          4.15 A-2-Unit 1 BAMT, in hallway        3    OP/1/A/1104/002         4.17 Turbine Building                   3    OP/1/A/1106/002 A       4.14 Turbine Building                   3    OP/0/AI1104/048          4.4         Step 3.7 OP/i/Ni1102/010          4.7 ________
Next actions---LPI LPI System Start-up (CR & SSF-CR)                                3    OP/1/A/1104/004          4.2 LPI Fill & S/U Valve lineup for AB 1st Floor                       3    OP/1/A/1104/004          4.5 LPI Breaker line up AB Pent. Rooms                     3    OP/1/AN1102/010          4.1 S/D Secondary Steam TB-3 &CR                           3    OP/I/A/1102/010              SD Align FDW clean-TB All Levels                      3    OP/i/A/1102/010          4.1 up AB-2                             4 &5   OP/I/A/1102/010         4.11 RCS H2 Sampling RB, AB-i, 2 & 3rd                  5    OP/I/A/1i03/008    _____RCS          Crud Burst I[Document No.]                                Rev. 0                            Page 238 of 240

ATTACHMENT 3 Safe Operation & Shutdown Rooms/Areas Tables R-2 & H-2 Bases Unit Shutdown Room List       Mode Turbine Building                      1,2,3 A-I hallway 8' S/ cal 65              1,2,3 A-i-hallway N of CoJ 82               1,2,3 A-i hallway 5' Stool 67               i1,2,3 A-i hallway col 82                    i1,2,3 A-I hallway 10' Stool 96)             1,2,3 A-l- BAMT Rm                             1 A-1iUniti1& 2BAMT Rm                     1 A-i-Col Q70                              i A-2 LOST Hatch area                   1,2,3 A-2-Unit 2 LOST Hatch area            1,2,3 A-2 Unit 3 LDST Hatch area            1,2,3 A-2-1&2 Chem. Add Panel                  1 A-2-CoI. P-63                            3 A-2-Unit i BAMT, in hallway              3 A-2 Hallway                              3 A-3-308                                  3 A-4-402                                  3 A-4-409                                  3 A-4-W Pent Rm                            3 A-4-E Pent                               3 Unit I BTP Rm                         1,2,3 Unit 2 BTP Rm                         1,2,3 Unit 3 BTP Rm                         1,2,3 Ui LPI Cooler Rm                      1,2,3 RB 779', Cable Room, iUB2                3 RB 779'                                  3 R-1G-W                                   3 R-il-around "A" CFT                      3 R-i-around "B"CFT                        3 R-B above Emer. Sum p                    3 R-B above RBNS                           3 R-B-20' above LD Cooler RM               3 R-3G East Side                           3 I1UB2                                    3 I1AT7                                   3 1MTC-4                                  3 I1AT5                                   3 Unit I Primary Sample Hood               3 AB SAMPLE RM.308                         3 RB, AB                                4&5
[Document No.]                              Rev. 0                 Page 239 of 240

ATTACHMENT 3 Safe Operation & Shutdown Rooms/Areas Tables R-2 & H-2 Bases Table R-2 & H-2 Results Table R-2 & H-2    Safe Operation & Shutdown RoomslAreas Room/Area                       Mode Applicability Turbine Building                                               1,2, 3 Equipment and Cable Rooms                                      1,2, 3 Auxiliary Building                                         1, 2, 3, 4, 5 Reactor Buildings                                            '3, 4, 5 S[Document No.]                           Rev. 0                        Page 240 of 240

SCOTT L. BATSON DUI<EVice                                                                     President Oconee Nuclear Station ENERGYS                                                                                   Duke Energy ONOI VP I 7800 Rochester Hwy Seneca, SC 29672 o: 864-873-3274
: f. 864-873-4208 Scott.Batson@duke-energy.com ONS-201 6-008 10 CFR 50.90 10 CFR 50 Appendix E February 4, 2016 Attn: Document Control Desk U. S. Nuclear Regulatory Commission 11555 Rockville Pike Rockville, MD 20852 Duke Energy Carolinas, LLC (Duke Energy)
Oconee Nuclear Station, Units 1, 2, and 3 Docket Numbers 50-269, 50-270, and 50-287 Renewed Operating License Nos. DPR-38, DPR-47, and DPR-55

==Subject:==
Response to Request for Additional Information Regarding the License Amendment Request (LAR) to Change the Oconee Nuclear Station (ONS)
Emergency Plan to Upgrade ONS Emergency Action Levels Based on NEI 99-01, Revision 6 License Amendment Request No. 2015-04 By letter dated June 23, 2015 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML15183A060), Duke Energy requested approval of a proposed change to the Emergency Action Levels (EALs) used at Oconee Nuclear Station (ONS). Duke Energy proposes to revise their current ONS EAL scheme to one based upon Nuclear Energy Institute (NEI) document NEI 99-01, "Development of Emergency Action Levels for Non-Passive Reactors," Revision 6 (ADAMS Accession No. ML12326A805).
The NRC staff reviewed the request and determined that additional information is needed to complete their review. A letter requesting additional information was sent on December 15, 2015 (ADAMS Accession No. ML15345A398). Enclosure 1Ito this letter provides the ONS response to the request for additional information. Enclosure 2 provides the ONS Emergency Action Level Technical Bases Document, Revision 0, that incorporates changes in response to the request for additional information and additional changes deemed necessary, as described in Enclosure 1.
This letter makes no new commitments or changes to any existing commitments. Should you have any questions regarding this request, please contact Chris Wasik, Regulatory Affairs Manager, at (864) 873-5789.

U.S. Nuclear Regulatory Commission February 4, 2016 Page 2 I declare under penalty of perjury that the foregoing is true and correct. Executed on the 4th day of February, 2016.
Sincerely, Scott L. Batson Vice President Oconee Nuclear Station I

==Enclosures:==
: 1. Response to Request for Additional Information
: 2. Oconee Nuclear Station Emergency Action Level Technical Bases Document, "Emergency Action Level Technical Bases" (Clean Version), Revision 0

U.S. Nuclear Regulatory Commission February 4, 2016 Page 3 cc w/ 

==Enclosures:==

Ms. Catherine Haney, Regional Administrator U. S. Nuclear Regulatory Commission - Region II Marquis One Tower 245 Peachtree Center Ave., NE, Suite 1200 Atlanta, GA 30303-1257 Mr. James Hall, Senior Project Manager (by electronic mail only)
Off ice of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission 11555 Rockville Pike Mail Stop 0-8 G9A Rockville, MD 20852 Mr. Jeffrey Whited, Project Manager (by electronic mail only)
Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission 11555 Rockville Pike Mail Stop 0-8 B1A Rockville, MD 20852 Mr. Eddy Crowe NRC Senior Resident Inspector Oconee Nuclear Site Ms. Susan Jenkins, Manager Radioactive & Infectious Waste Management Division of Waste Management South Carolina Department of Health and Environmental Control 2600 Bull St.
Columbia, SC 29201

U.S. Nuclear Regulatory Commission February 4, 2016 Page 4 cc w/o 

==Enclosures:==

Mr. Scott Krein, Director Oconee County Emergency Preparedness 415 South Pine Street Walhalla, SC 29691 Ms. Denise Kwiatek, Deputy Director Pickens County Emergency Preparedness 1509 Walhalla Highway Pickens, SC 29671 Mr. Kim Stenson, Director South Carolina Emergency Management Division 2779 Fish Hatchery Road West Columbia, SC 29172

ONS-201 6-008 Enclosure 1 ENCLOSURE I RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION LICENSE AMENDMENT REQUEST NO. 2015-04 OCONEE NUCLEAR POWER STATION, UNIT 1, 2, AND 3 DOCKET NOS 50-269, 50-270, and 50-287 RENEWED LICENSE NOS. DPR-38, DPR-47, and DPR-55 14 Pages Follow

ONS-201 6-008 Enclosure 1 Page 1 of 14 By letter dated June 23, 2015 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML15183A060), Duke Energy requested approval of a proposed change to the Emergency Action Levels (EALs) used at Oconee Nuclear Station (ONS) which would revise the current ONS EAL scheme to one based upon Nuclear Energy Institute (NEI) document NEI 99-01, "Development of Emergency Action Levels for Non-Passive Reactors," Revision 6 (ADAMS Accession No. ML12326A805). The NRC staff reviewed the request and determined that additional information is needed to complete their review. A letter requesting additional information was sent on December 15, 2015 (ADAMS Accession No. ML15345A398).
Duke Energy provides the following response to the request for additional information (RAI) regarding the License Amendment Request to revise the ONS EAL Scheme to one based on NEI 99-01, Revision 6. Changes to the ONS Technical Bases Document as a result of this RAI are identified within the ONS RAI Response in the table below. Changes were also made to the ONS Technical Bases Document that were not the result of an RAI. These additional changes are described in a table below the RAl response, with an explanation of why each change was deemed necessary. The revised ONS EAL Technical Bases Document is provided in Enclosure 2 with revision bars indicating changes from the original submittal.
RAI-ONS-#                                  Question                                             ONS Response Section 4.3, "Instrumentation Used for EALs," to NEI 99-01,    ONS has confirmed that setpoints and indications used in the proposed EAL Revision 6, states, in part, that: "Scheme developers should   scheme are within the calibrated range(s) of the stated instrumentation and ensure that specific values used as EAL setpoints are within   that the resolution of the instrumentation is appropriate for the the ca/ibratedrange of the referenced instrumentation."        setpoint/indication.
01        Please confirm that all setpoints and indications used in the proposed EAL scheme are within the calibrated range(s) of the stated instrumentation and that the resolution of the instrumentation is appropriate for the setpoint/indication.
Section 2.5, 'Technical Bases Information," of Duke's LAR     The ONS site specific and NEI 99-01    generic bases sections have been states, in part, that: "A Plant-specific basis section that   combined into a single bases section   for each EAL. Section 2.5 "Technical provides ONS-relevant information concerning the EAL.         Bases Information" has been revised    accordingly.
This is followed by a Generic basis section that provides a   Reudnbaswreplibehvbeneet.
description of the rationale for the EAL as provided in NEI   Rdnatbss              hr plcbe           aebedltd 02        99-01 Rev. 6." Due to the high probability that EAL decision-makers may be confused between these two sections when the information appears to be inconsistent, please justify the rational for two sections when it is acceptable to just have one basis section that is specific to the plant, or revise accordingly to eliminate potential confusion by user.

ONS-201 6-008 Enclosure 1 Page 2 of 14 RAI.ONS-#                               Question                                             ONS Response In Section 5.0, "Definitions," of Duke's LAR, the definition  Site Area Emergency was verified consistent with the NEI 99-01, Revision 6 for Site Area Emergency is not consistent with the definition definition.
in the NRC-endorsed guidance. Please revise accordingly, or provide further justification for inconsistency with the   Also, after further review, determined that the following definitions should be 03     NRC-endorsed guidance.                                        included in Section 5.0 to be consistent with NEI 99-01 Revision 6:
* Emergency Action Level (E-AL),          --
* Emergency Classification Level (ECL),
* Fission Product Barrier Threshold, and
* Initiating Condition For the following EALs, please explain why the listed         Deleted Note 3 from EALs RAI1.2, RS1 .2 and RG1 .2 04     NOTEs were included, or revise accordingly..,
* RAI.2- NOTE-3
* RSI.2-NOTE-3
_______          RG1.2  - NOTE-3

0NS-201 6-008 Enclosure 1 Page 3 of 14 RAI-ONS-#                           Question                                              ONS Response For EALs RU1, RA1, RS1 and RG1, please explain why the       The proposed ONS RGI .1, RSI .1 and RA1 .1 EAL threshold values are based values and thresholds developed for these EALs are           on NEI 99-0 1 Rev 6 and utilize the site specific URI/RASCAL dose inconsistent with what has already been approved for ONS,    assessment model (v2.0.0.1) for their development.
or revise accordingly.                                       The proposed ONS RUI.1 EAL threshold value is based on NEI 99-01 Rev 6 and utilizes 2 times the 0DCM limit for its development.
The current ONS RG1 .1 and RSI .1 EALs are based on NUMARC/NESP-007 Rev 2.
05 The current ONS RAI.1 EAL threshold value is-based on NUMARC/NESP-007 Rev 2 and utilizes 200 times the 0DCM limit for its development.
The current ONS RUI.1 EAL threshold value is based on NUMARC/NESP-007 Rev 2 and utilizes 2 times the 0DCM limit for its development.
The table below provides a comparison between the current and proposed ONS gaseous effluent EAL thresholds for the GE, SAE, Alert and Unusual Event emergency classification levels:
Monitor             GE                   SAE                      Alert                Unusual Event RA-6(p)NEP:            3.OOE+5  NEI:           3.O0E+4   NEI:          3.OOE+3            N/A RAI46(cm)NUMARC:       2.09E+I6 NUJMARC:       2.09E+5   NUMARC: 2.09E+4 RAI-45 (cpm)         N/A                   N/A                      N/A             NEI:          1.41E+5 NUJMARC:      9.35E+5

ONS-201 6-008 Enclosure 1 Page 4 of 14 RAI-ONS-#  Question                    ONS Response RGI.1 Differences The NEI 99-01 Rev 60ONS RGI.1 EAL threshold value utilizes the RASCAL based dose assessment model assuming a reactor accident source term and meteorology reflective of conditions consistent with a design basis accident and definition of the General Emergency classification level. The NUMARC/NESP-007 Rev 20ONS RGI .1 EAL threshold value utilizes an annual average X/Q in accordance with the NRC endorsed "Methodology for Development of Emergency Action Levels NUMARC/NESP-007 Revision 2 Questions and Answers," dated June 1993. The X/Q input alone accounts for a difference of more than one order of magnitude. Differences in assumed source term mix also contribute to the difference in threshold values. The basis and development methodologies for RGI.1 are completely different between 05,             NUMARC/NESP-007 Rev 2 and NEI 99-01 Rev 6.
continued RSI .1 Differences Both NEI 99-01 Rev 6 and the NUMARC/NESP-007 Rev 2 RS1I.1 EAL threshold values are established as 1/10 the RGI.1 threshold values. Thus, the differences described above for RGI.1 are the same for RS1.1.
RAI .1 Differences NEI 99-01 Rev 6 established the basis for RAI.1 as 1/10 the RSI.1 threshold value. NUMARC/NESP-007 Rev 2 established the basis for RA1.1 as 200 times the ODCM limit. The basis and development methodologies for RAI .1 are completely different between NUMARC/NESP-007 Rev 2 and NEI 99-01 Rev 6.

ONS-201 6-008 Enclosure 1 Page 5 of 14 RAI-ONS-#                             Question .ONS                                              Response RU1. 1 Differences Both the NEI 99-01 Rev 6 and the NUMARC/NESP-007 Rev 20ONS RUI .1 EAL threshold values are derived from the same the ODCM equation:
S~obody (cpm) =   4r2 f         *i(i 500*     ~ C          +bkg
                                                                                                                                    +_3
* Eq, RU 1.1 Differences, continued 05,                                                                                                         --
continued                                                                Both the NEI 99-01 Rev 6 and the NUMARC/NESP-007 Rev 20ONS RUI .1 EAL threshold values use the same input values with the exception of source term.
* The NUMARC/NESP-007 Rev 20ONS RUI.1 EAL threshold value is based on Xe-133.
* The NEI 99-01 Rev 60ONS RU1.1 EAL threshold value is based on NUREG-1 940 Table 1-2 normal RCS Noble Gas term source mix.
Neither NUMARC/NESP-007 Rev 2 nor NEI 99-01 Rev 6 provides source term guidance for the RU1.1 EAL threshold value. NUREG-1940 is used for the ONS RAI.1, RSI.1 and RG1.1 radiological effluent EAL threshold source term basis. NUREG-1940 was selected as the source term basis for the RUI.1 radiological effluent EAL threshold to maintain consistency of guidance document bases across the emergency classification thresholds.
For EAL RA2.2, the information in the NEI 99-01 Basis          Re-instated the following text to the RA2.2 bases:
section does  not contain to thisallparticular of the actual EAL.information from 06    NE! 99-01  germane                              Please explain "This EAL applies to irradiated fuel that is licensed for dry storage up to the why this information was omitted, or revise accordingly,       point that the loaded storage cask is sealed. Once sealed, damage to a loaded cask causing loss of the CONFINEMENT BOUNDARY is classified in accordance with EAL EU1.1 ."

0NS-201 6-008 Enclosure 1 Page 6 of 14 RAI-ONS-#                               Question                                             ONS Response For EAL RG2.I, the level value in the EAL, -23 feet, is       Corrected RG2.I to read, 'Spent fuel pooi level cannot be restored to at least inconsistent with the stated level in the ONS Basis, -23.5     -23.5 ft. for ->60 mai. (Note 1).'
feet. Please correct the discrepancy and ensure the correct 07    ONS Level 3 value is used in this EAL, or provide justification for this difference.
For EALs RA3.2 and HA5.1, both of these EALs are               ONS will maintain two tables, R-2 and H-2 for end-user usability.
applicable to the same areas; therefore, it is not required to NR-nosdgiacidcteththeiespcfcltoflntrmsr have two separate tables (Table R-2 for EAL RA3.2, and         NR-nosdgiacincteththeiespcfcltoflntrmsr Table H-2 for EAL HA5.I). In addition, the listed areas are    areas with entry-related mode applicability should include those rooms or 08    too all-encompassing and should be refined to just those       areas that contain equipment which require a manual/local action as specified specific areas where access is required as described in the    in operating procedures used for normal plant operation, cooldown and NRC ndosedguiance Plaserevse te aeas orshutdown.               EAL Technical Bases Document, Attachment 3, Safe Operation &
provide further justification.                                 Shutdown Rooms/Areas Tables R-3 & H-2 Bases, provides analysis for the areas listed. The Oconee plant is constructed with essentially a common Turbine Building and a common Auxiliary Building for all three units. Very little separation is provided for key areas joined by common hallways. For this reason, the areas provided in Table R-2/H-2 are necessarily all-encompassing.
For the description of Category E-Independent Spent            Revised the ISFSI category introduction:
09     Fuel Storage installation (ISFSI), please add wording related to how security events at the ISFSI are declared.      "The ONS ISFSI is contained wholly within the plant Protected Area. Therefore a security event related to the ISFSI would be applicable to EALs HUI .1, HAl.land HSI.I."

ONS-201 6-008 Enclosure 1 Page 7 of 14 RAI-ONS-# JQuestion                                                                      ONS Response For EALs CU2.1 and SA1.1, please confirm that the         The CU2.1 and SAI.1 listed AC Power Sources include:
listed alternating current (AC) power sources are timely and reliable (i.e., will they be available when needed?), Offsite:
and If not, please remove them from the list.                  0     Unit Normal Transformer (backcharged)
S Unit Startup Transformer (SWYD) a     Another Unit Startup Transformer (aligned) (SWYD)
S CT5 (Central/energizing Standby Bus)
Emergency:
* Unit Startup Transformer (Keowee)
* Another Unit Startup Transformer (aligned) (Keowee)
* CT4
* CT5 (dedicated line/energizing Standby Bus) 10                                                              The timeliness of the listed AC power sources is dependent upon the required alignment as indicated by the parenthetical following the source. For example, Unit Normal Transformer (backcharged) is only credited as a viable offsite source while it is aligned and the associated unit is on backcharge such that it is capable of powering the essential busses within 15 minutes, whether or not the buses are currently powered from it.
Routine maintenance and surveillance programs are in place to ensure the reliability of the offsite and emergency power paths. The important functions of these systems are monitored under the 10 CFR 50.65 Maintenance Rule program. In addition, Operations performs routine rounds to verify these paths are operable and/or available. While the Central switchyard is outside the scope of the Maintenance Rule, it is subject to the North American Electric Reliability Corporation (NERC) reliability standards.

ONS-201 6-008 Enclosure 1 Page 8 of 14 RAI-ONS-#                              Question                                            ONS Response The intent of EALs CA2.1, 551.1, SG1.1 and SG1.2 is to       Removed table references from CA2.1, SS1.1,-SGI.1, and SG1.2 as make the appropriate emergency classification upon a         suggested.
loss of ALL power sources. However, the list as 11    developed eliminates the possibility of alternative AC power sources energizing an essential bus, thus negating the need for declaring the applicable EAL.
Please remove the table of AC power sources, or provide further justification.
For EALs CU3.1 and CA3.1, the guidance in the ONS            Added new Note 10 to CU3.1 and CA3.1:
basis related to the process to follow When core exit              'oe1:I         h bec             frlal    C eprtr          niaincue 12 thermocouples (CETCs) are unavailable/unreliable                   byt 1:Ilossnc the dea of etremvlcaablityS           clpeassreifdication should sed 12 should be carried over as a NOTE for this EAL to ensure            basdo timeos    tofboily deata."      vlcpbliy lsiiato           hudb this information is presented to EAL decision-makers (i.e., on the EAL wallboard).
For EALs CU5.1 and SU7.1, both of these EALs are             The two separate tables are provided for end-user usability.
applicable to the same areas; therefore, it is not required  EOF Phones are not available for Onsite communications. Removed EOF to have two separate tables (Table C-5 for EAL CU5.1, and Table S-4 for EAL SU7.1). Please provide basis for       PoeSse             rmOst oun listing separate tables, or revise accordingly. In addition, Dedicated satellite phones are available for Offsite communications only.
please explain the following, and if necessary, revise       Removed Satellite phones from Onsite and NRC columns.
13    teeEsacodny:There                                                  are two dedicated satellite phones. One is in the EOF Offsite Agency
* How the EOF phone system can suffice for             Communicator Area and one is in the Offsite Monitoring Area.
onsite communications?
                  *Are dedicated satellite phones available for onsite, offsite, and NRC communications?
                  *How many dedicated satellite phones are there, as it may be likely that the NRC phone will need to be dedicated for NRC communications?

ONS-201 6-008 Enclosure 1 Page 9 of 14 RAI-ONS-#                             Question                                               ONS Response For EAL HU2.1, please explain why the use of the                Revised HU2 IC and EAL to use "aBE" vs. "DBE."
abbreviation DBE (design basis earthquake) and OBE (opratng eathqake asi ar inonsstetlyuse. Wile         Removed the detailed information related to the Strong Motion Accelerometer and Tendon Gallery Peak Acceleration Recorders, as they cannot be used to the guidance eventually states that these two terms are the     clsiyaemrny(15iutidcto).
same for ONS, please revise to be consistent with theclsiyaemrny(>5iutincto).
guidance or consistent with the ONS use of the equivalent term(s), or provide further justification. Also, please explain 14    the need for the detailed information related to the Strong Motion Accelerometer and Tendon Gallery Peak Acceleration Recorders, as it could be implied that, pending approval, the staff agrees that these recorders will suffice for determining this EAL, or revise accordingly.
For EAL HU3.2, please explain why EAL SA9.1 is not              Added escalation pathway of EAL SA9.1 to EAL HU3.2 bases.
also listed, along with EAL CA6.1, to ensure that the           During review of the escalation pathways on the Wallboards, it was noted that 15  escaion pathcordi   ally prtn.mdsaeadrseo                      HA2, HA3, and HA4 incorrectly referenced HA6.1 instead of CA6.1. This has revie acordigly.been                                corrected in the Wallboards.
For EAL HU3.5, please explain, in greater detail, what          The Jocassee Dam is located upstream of the Oconee Nuclear Station.
this EAL is intended for and how it meets the definition of     Historically, Condition B associated with the failure of the Jocassee Dam was an Unusual Event. In particular, please explain what            included in the Emergency Plan because it could cause the turbine building Condition B is, and how it related to the site radiological     basement to flood, disabling important equipment (e.g., main and emergency emergency plan                                                  feedwater pumps). Escalation of the event to a higher category was based on 16                                                                    the ability to maintain core cooling or shutdown functions. With the migration to NEI 99-01, Rev 6 guidance, symptom-based effects from flooding on the radiological emergency plan would be addressed through the loss of power EALs and the Natural/Technical Hazard EALs, based on actual losses of power or safety-related equipment. FERC Emergency Planning maintains evacuation protocols for the public based on dam failures; therefore, HU3.5 was removed.

ONS-201 6-008 Enclosure I Page 10 of 14 RAI-ONS-#                           Question                        ]ONS                       Response For EAL HS3.1, please explain, in greater detail, what     In the ONS submittal to adopt the NUMARC/NESP-007, Rev. 2, this EAL is intended to do and why it is at the Site Area  Classification Scheme, dated May 5, 1994, Duke Energy identified the Emergency classification level. Has there been an NRC      concerns associated with the failure of the Keowee Hydro Dam and commitment made related to ONS dam failures?               characterized this failure as a Site Area Emergency. A failure of Keowee Hydro would result in loss of the emergency AC power supply and the potential to lose the ultimate heat sink source. Some flooding of the site may result. Evaluation of the plant status following failure of the dam would determine the need to escalate emergency classifications. More recently, in a January 15, 2010 letter to the NRC, Duke Energy communicated measures in place to address postulated external flood threat issues. Within that letter, Duke Energy referred to the Jocassee EAP, which identifies two conditions related to the status of the dam: Condition A- Failure is Imminent or Has Occurred; Condition B- Potentially Hazardous Situation is Developing. Condition A initiates a call tree that notifies offsite agencies to implement specific actions to protect/warn the public as well as notifications to the ONS Unit 2 Control Room. The letter further communicated that once notification has been received at ONS, the ONS Emergency Plan and associated response procedures are implemented. Existing plant 17                                                               procedures included guidance to trip each of the ONS units and bring the units to a safe shutdown condition.
Although no specific commitments related to ONS dam failures were identified, previous versions of the ONS Emergency Plan and ongoing flood hazard discussions have included existing actions within the Site and Hydro Emergency Plans. NRC approval of this submittal would supersede previous Emergency Plans. Symptom-based affects from flooding on the radiological emergency plan would be addressed through the loss of power EALs and the Natural/Technical Hazard EALs, based on actual losses of power or safety-related equipment. The Federal Energy Regulatory Commission (FERC), as regulator for both Jocassee and Keowee dams, has Emergency Planning requirements and oversight for evacuation protocols for the public safety based on dam failures. Therefore, HS3.1 has been removed.

ONS-201 6-008 Enclosure 1 Page 11 of 14 RAI-ONS-#                             Question                                           ONS Response For EALs HU4.1 and HU4.2, the areas listed in Table H-     Table H-I Fire Areas are based on the ONS Fire Protection design basis 1 seem to be vague or too all-encompassing. Please         specification. Table H-i Fire Areas include those structures containing explain if the listed areas are all the areas that contain functions and systems required for safe operation, shutdown and cooldown 18    equipment needed for safe operation, safe shutdown and     of the plant (SAFETY SYSTEMS).
safet cool-downiand ifothese aea~s.cnb ietndt              A balance must be established between defining major plant structures limi fo thse  E~s.containing cosideatin                                     safe shutdown equipment as fire areas versus a detailed list of areas for every safety system component location. The Table H-I list of fire areas achieves that balance in support of timely and accurate emergency classification for the end-user.
For EALs HU4.3 and HU4.4, please confirm that the          The ONS ISFSI is contained wholly within the plant Protected Area. Therefore 19 Independent Spent Fuel Storage Installation (ISFSI)        the ISFSI would be applicable to EALs HU4.3 and HU4.4 for fires within the 19 would be an area applicable to these EALs, or revise        plant Protected Area accordingly.
For EAL HS6.1, please consider adding operating mode        Revised HS6.I mode applicability from ALL to Modes 1 - 6.
speifiityto ey afey fnctonslised he               n te EL. Revised HS6.1 based upon an assessment of applicable modes for each of the listed safety function as follows:
                                                                     'An event has resulted in plant control being transferred from the Control Room to the Auxiliary Shutdown Panels or Standby Shutdown Facility (SSF) 20                                                                       AND Control of any of the following key safety functions is not reestablished within 15 min. (Note 1):
* Reactivity (Modes 1, 2 and 3 only)
* Core Cooling
                                                                            *RCS heat removal"

ONS-201 6-008 Enclosure 1 Page 12 of 14 RAI-ONS-#                            Question                                        ONS Response Frevalsae the. andudesig plase ctonfir that  di    NSa  A conference call between ONS and NRC personnel was conducted on evaluatedntheio uniqed adesign apelcabltsof thesdigiALs  1/19/2016 to gain further understanding of this question. As a result of this call, instumetaton sed an th appicailiy o thse A~s      it was determined that the question specifically pertains to the digital Reactor tonfrtheapiabl faditnluE~re md.Ino wadditinepleased        Protective System (RPS) at Oconee and whether the design of this digital cponfirmthat additiontaltEAn arenot warrnted bsdsystem           introduced failure modes which affect applicability of EALs or warrant upondigtalinsrumntaton nd ontol.additional                  EALs.
As part of the RPS and Engineered Safeguards Protective System (ESPS) replacement project, a Failure Modes and Effects Analysis (FMEA) was performed. The FMEA, reviewed by NRC staff as part of the licensing submittal for the RPS/ESPS replacement [ML100220016, Ref 72], consisted of a 21                                                            systematic qualitative analysis of the RPS and ESPS design with the primary objectives of identifying all credible failure modes, evaluating the consequence and effects of failures, and verifying that the design satisfies single-failure criterion applicable to the replacement RPS/ESPS.
While the design of the Oconee RPS is now digital, the functions remain the same. Credible failures within the system have been analyzed. These failures cause the system to fail in a predefined safe state, as was true for the original analog RPS system. Based on an assessment of the FMEA analysis there is no impact on the scope of SU3.1 or SA3.1. The EALs for Oconee that the RPS is applicable to still apply as written, and no new initiating criteria are deemed to be warranted.

ONS-201 6-008 Enclosure 1 Page 13 of 14 RAI-ONS-#                           Question                                           ONS Response 22    For Table F-2 in the Fission Barrier Matrix, please       The values presented in Table F-2 were derived from Table 2 of the Oconee provide the information related to how this table was     Core Damage Assessment Guidelines (OSC-5283) as follows:
developed.                                                FC Loss Column FC Loss represents the expected Containment High Range Radiation Monitor (RIA-57 & RIA-58) response based on a LOCA for various periods of time after SID with 4% clad damage. The values in Table F-2 in this column were calculated first by multiplying the 100% clad damage values in Table 2 of OSC-5283 by .04 and then rounded to account for variation in monitor readings due to decay.
CMT Potential Loss Column CMT Potential Loss represents the expected Containment High Range Radiation Monitor (RIA-57 & RIA-58) response based on a LOCA for various periods of time after SID with 20% clad damage. The values in Table F-2 in this column were calculated first by multiplying the 100% clad damage values in Table 2 of OSC-5283 by 0.2 and then rounded to account for variation in monitor readings due to decay.
Under the Fission Product Barrier (FPB) Matrix, the cited As per response to RAI-ONS-02, the ONS site specific and NEI 99-01 NEI 99-01 Basis sections for several of the FPB criteria  Revision 6 bases have been unified.
23    are not from NRC-endorsed guidance (NEI 99-01, Revision 6). Please either revise to what has actually been endorsed, or (depending on respond to RAI-02) consider unifying the basis sections into one.

ONS-201 6-008 Enclosure 1 Page 14 of 14 Summary of EAL Changes Not Associated with RAI Response The table below summarizes the changes that have been incorporated into the EAL Technical Bases Document contained in Enclosure 2 that are not involved in the NRC RAI response.
EAL #                                                                         Description HUI .1      Separated HU1 .1 into HU1 .1, HU1 .2 and HU1 .3 to remove the "OR" qualifier in the NEI 99-01, Revision 6 template, as follows:
                        "HU1 .1 A SECURITY CONDITION that does not involve a HOSTILE ACTION as reported by the Site Security Supervision" "HU1 .2 Notification of a credible security threat directed at the site" "HUl1.3 A validated notification from the NRC providing information of an aircraft threat" HA1 .1      Separated HA1 .1 into HA1 .1 and HA1 .2 to remove the "OR" qualifier in the NEI 99-01, Revision 6 template, as follows:
                       "HAl .1 A HOSTILE ACTION is occurring or has occurred within the OWNER CONTROLLED AREA as reported by the Security Shift Supervision" "HA1 .2 A validated notification from NRC of an aircraft threat within 30 mai. of the site" RAl1.4      Administrative change to correct the definition for Site Boundary by changing "DPC" to "Duke Energy".
RG1.3       Administrative change to update ONS Basis Reference 1Ito be AD-EP-ALL-0203 which has superseded SH/0/B/2005/002.
SAl.1       Administrative change to correct the Mode Applicability to read "4-Hot Shutdown" instead of "3-Hot Shutdown".
SA6.1/SS6.1     Administrative change to correct the Subcategory to read "6 - RPS Failure" instead of "2 - RPS Failure".
HA7.1/HS7.1/    Administrative change to update the Operations Shift Manager acronym to be "SM" instead of "OSM".
HG7.1

ENCLOSURE 2 ONS EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT, REVISION 0 OCONEE NUCLEAR POWER STATION, UNIT 1, 2, AND 3 DOCKET NOS 50-269, 50-270, and 50-287 RENEWED LICENSE NOS. DPR-38, DPR-47, and DPR-55 240 Pages Follow

*,*DUKE         OCONEE NCERSTATION EMERGENCY ACTION LEVEL TECHNICAL BASES Revision 0 2/4/16 S[Document No.1         Rev. 0       Page 1 of 240

Table of Contents SECTION                                                                                                     PAGE 1.0  PURPOSE........................................................................................... 3 2.0 DISCUSSION ........................................................................................       3 2.1 Background........................................................................................      3 2.2 Fission Product Barriers .......................................................................... 4 2.3 Fission Product Barrier Classification Criteria...................................................       4 2.4 EAL Organization .................................................................................      5 2.5 Technical Bases Information......................................................................       7 2.6 Operating Mode Applicability ....... .............................................................       8 3.0 GUIDANCE ON MAKING EMERGENCY CLASSIFICATIONS .................................. 9 3.1 General Considerations........................................................................... 9 3.2 Classification M ethodology.........          J........................................10

==4.0 REFERENCES==
 ....................................  ................................................        13 4.1 Developmental................................................................................... 13 4.2 Implementing..................................................................................... 13 5.0 DEFINITIONS, ACRONYMS & ABBREVIATIONS............................................... 14 5.1 Definitions ........................................................................................ 14 5.2 Acronyms and Abbreviations .................................................................... 19 6.0  ONS TO NEI 99-01 Rev. 6 EAL CROSS-REFERENCE........................................ 22 7.0  ATTACHMENTS................................................................................... 26 1      Emergency Action Level Technical Bases .......................................... :.....27 Cateaorv R Abnormal Rad Release I Rad Effluent .............................. 27 Catecqorv E ISFSI................................................................... 63 Catecqorv C Cold Shutdown / Refueling System Malfunction.................... 66
_Cateaqorv H Hazards .............................................................. 101
_Cate~qorv S System Malfunction ................................................. 145 Cate~Qorv F Fission Product Barrier Degradation............................... 184 2      Fission Product Barrier Loss / Potential Loss Matrix and Bases ........................................................................ 189 3     Safe Operation & Shutdown Rooms/Areas Tables R-2 & H-2 Bases ............... 235 l[Document No.]                        1Rev.                  0                  [Page                 2 of 240

1.0      PURPOSE This document provides an explanation and rationale for each Emergency Action Level (EAL) included in the EAL Upgrade Project for Oconee Nuclear Station (ONS). It should be used to facilitate review of the ONS EALs and provide historical documentation for future reference.
Decision-makers responsible for implementation of RPI0/A/1 000/001, Emergency Classification, may use this document as a technical reference in support of EAL interpretation.
This information may assist the Emergency Coordinator in making classifications, particularly those involving judgment or multiple events. The basis information may also be useful in training and for explaining event classifications to off-site officials.
The expectation is that emergency classifications a~re to be made as soon as conditions are present and recognizable for the classification, but within 15 minutes or less in all cases of conditions present. Use of this document for assistance is not intended to delay the emergency classification.
Because the information in a basis document can a~ffect emergency classification decision-making (e.g., the Emergency Coordinator refers to it during an event), the NRC staff expects that changes to the basis document will be evaluated in accordance with the provisions of 10 CFR 50.54(q).
2.0      DISCUSSION 2.1      Background EALs are the plant-specific indications, conditions or instrument readings that are utilized to classify emergency conditions defined in the ONS Emergency Plan.
In 1992, the NRC endorsed NUMARC/NESP-007 "Methodology for Development of Emergency Action Levels" as an alternative to NUREG-0654 EAL guidance.
NEI 99-01 (NUMARC/NESP-007) Revisions 4 and 5 were subsequently issued for industry implementation. Enhancements over earlier revisions included:
* Consolidating the system malfunction initiating conditions and example emergency action levels which address conditions that may be postulated to occur during plant shutdown conditions.
* Initiating conditions and example emergency action levels that fully address conditions that may be postulated to occur at permanently Defueled Stations and Independent Spent Fuel Storage Installations (ISFSIs).
* Simplifying the fission product barrier EAL threshold for a Site Area Emergency.
Subsequently, Revision 6 of NEI 99-01 has been issued which incorporates resolutions to numerous implementation issues including the NRC EAL Frequently Asked Questions (FAQs).
Using NEI 99-01 Revision 6, "Methodology for the Development of Emergency Action Levels for Non-Passive Reactors," November 2012 (ADAMS Accession Number ML12326A805) (ref.
4.1.1), ONS conducted an EAL implementation upgrade project that produced the EALs discussed herein.
I[Document No.]                        IRev.0                                           Page 3of 240

2.2     Fission Product Barriers Fission product barrier thresholds represent threats to the defense in depth design concept that precludes the release of radioactive fission products to the environment. This concept relies on multiple physical barriers, any one of which, if maintained intact, precludes the release of significant amounts of radioactive fission products to the environment.
Many of the EALs derived from the NEI methodology are fission product barrier threshold based. That is, the conditions that define the EALs are based upon thresholds that represent the loss or potential loss of one or more of the three fission product barriers. "Loss" and "Potential Loss" signify the relative damage and threat of damage to the barrier. A "Loss" threshpld means the barrier no longer assures contain mentlof radioactive materials. A "Potential Loss" threshold implies an increased probability of barrier loss and decreased certainty of maintaining the barrier.
The primary fission product barriers are:
A. Fuel Clad (FC): The Fuel Clad Barrier consists of the cladding material that contains the fuel pellets.
B. Reactor Coolant System (RCS): The RCS Barrier includes the RCS primary side and its connections up to and including the pressurizer safety and relief valves, and other connections up to and including the primary isolation valves.
C. Containment (CMT): The Containment (Reactor Building) Barrier includes the Reactor Building and connections up to and including the outermost containment isolation valves. This barrier also includes the main steam, feedwater, and blowdown line extensions outside the Reactor Building up to and including the outermost secondary side isolation valve. Containment Barrier thresholds are used as criteria for escalation of the Emergency Classification Level (ECL) from Alert to a Site Area Emergency or a General Emergency.
2.3     Fission Product Barrier Classification Criteria The following criteria are the bases for event classification related to fission product barrier loss or potential loss:
Alert:
Any loss or any potential loss of either Fuel Clad or RCS Barrier Site Area Emergqency:
Loss or potential loss of any two barriers General Emergency:
Loss of any two barriersand loss or potentialloss of the third barrier S[Document No.]                                 Rev. 0                                  Page 4 of 240

2.4     EAL Organization The ONS EAL scheme includes the following features:
* Division of the EAL set into three broad groups:
o EALs applicable under any plant operating modes - This group would be reviewed by the EAL-user any time emergency classification is considered.
o EALs applicable only under hot operating modes - This group would only be reviewed by the EAL-user when the plant is in Hot Shutdown, Hot Standby, Startup, or Power Operation mode.
o E'ALs applicable only under col operating modes - This group would only be reviewed by the EAL-user when the plant is in Cold Shutdown, Refueling mode or No Mode.
The purpose of the groups is to avoid review of hot condition IEALs when the plant is in a cold condition and avoid review of cold condition EALs when the plant is in a hot condition. This approach significantly minimizes the total number of EALs that must be reviewed by the EAL-user for a given plant condition, reduces EAL-user reading burden and, thereby, speeds identification of the EAL that applies to the emergency.
* Within each group, assignment of EALs to categories and subcategories:
Category and subcategory titles are selected to represent conditions that are operationally significant to the EAL-user. The ONS EAL categories are aligned to and represent the NEI 99-01 "Recognition Categories." Subcategories are used in the ONS scheme as necessary to further divide the EALs of a category into logical sets of possible emergency classification thresholds. The ONS EAL categories and subcategories are listed below.
I[Document No.]                                   Rev. 0                               Page 5 of 240

EAL Groups, Categories and Subcategories EAL Group/Category                               EAL Subcategory Any Operating Mode:
R - Abnormal Rad Levels / Rad Effluent   1 - Radiological Effluent 2- Irradiated Fuel Event 3- Area Radiation Levels H - Hazards and Other Conditions         1 - Security, Affecting Plant Safety               2- Seismic Ekent 3 - Natural or Technological Hazard 4- Fire 5 - Hazardous Gas 6 -Control Ro~om Evacuation 7 - Emergency Coordinator Judgment E - Independent Spent Fuel Storage       1 - Confinement Boundary Installation (ISFSI)
Hot Conditions:
S - System Malfunction                   1 - Loss of Essential AC Power 2 - Loss of Vital DC Power 3 - Loss of Control Room Indications 4-  RCS  Activity 5-  RCS  Leakage 6-  RPS  Failure 7-  Loss of Communications 8 - Containment Failure 9 - Hazardous Event Affecting Safety Systems F - Fission Product Barrier Degradation  None Cold Conditions:
C - Cold Shutdown I Refueling System     1 - RCS Level Malfunction                          2 - Loss of Essential AC Power 3-  RCS Temperature 4-  Loss of Vital DC Power 5-  Loss of Communications 6-  Hazardous Event Affecting Safety Systems The primary tool for determining the emergency classification level is the EAL Classification Matrix. The user of the EAL Classification Matrix may (but is not required to) consult the EAL Technical Bases in order to obtain additional information concerning the EALs under classification consideration. The user should consult Section 3.0 and Attachments 1 & 2 of this document for such information.
I[Document No.]                           I        Rev. 0                                    Page 6 of 240

2.5     Technical Bases Information EAL technical bases are provided in Attachment 1 for each EAL according to EAL group (Any, Hot, Cold), EAL category (R, C, H, 5, E and F) and EAL subcategory. A summary explanation of each category and subcategory is given at the beginning of the technical bases discussions of the EALs included in the category. For each EAL, the following information is provided:
Category Letter & Title Subcateqiory Number & Title Initiating Condition (IC)
Site-specific description of the generic IC given in NEI 99-01 Rev. 6.
EAL Identifier (enclosed in rectanqle)
Each EAL is assigned a unique identifier to support accurate communication of the emergency classification to onsite and offsite personnel. Four characters define each EAL identifier:
: 1. First character (letter): Corresponds to the EAL category as described above (R, C, H, S, EorF)
: 2. Second character (letter): The emergency classification (G, 5, A or U)
G = General Emergency S = Site Area Emergency A = Alert U = Unusual Event
: 3. Third character (number): Subcategory number within the given category.
Subcategories are sequentially numbered beginning with the number one (1). If a category does not have a subcategory, this character is assigned the number one (1).
: 4. Fourth character (number): The numerical sequence of the EAL within the EAL subcategory. If the subcategory has only one EAL, it is given the number one (1).
Classification (enclosed in rectangle):
Unusual Event (U), Alert (A), Site Area Emergency (S) or General Emergency (G)
EAL (enclosed in rectanqle)
Exact wording of the EAL as it appears in the EAL Classification Matrix Mode Applicability One or more of the following plant operating conditions comprise the mode to which each EAL is applicable: 1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown, 5   -
Cold Shutdown, 6 - Refueling, NM      - No Mode, or Any. (See Section 2.6 for operating mode definitions)
Definitions:
If the EAL wording contains a defined term, the definition of the term is included in this section. These definitions can also be found in Section 5.1.
[Document No.]                                   Rev. 0                               Page 7 of 240

Basis:
A basis section that provides ONS-relevant information concerning the EAL as well as a description of the rationale for the EAL as provided in NEI 99-01 Rev. 6.
ONS Basis Reference(s):
Site-specific source documentation from which the EAL is derived 2.6     Operating Mode Applicability (ref. 4.1.6) 1     Power Operation Keff>~ 0.99 and reactor thermal power > 5%
2     Startup      i Keff > 0.99 and reactor thermal power < 5%
3     Hot Standby Keff < 0.99 and average coolant temperature > 250&deg;FI 4     Hot Shutdown Keff < 0.99 and average coolant temperature 250&deg;F    > Tavg > 2000 F and all reactor vessel head closure bolts fully tensioned 5     Cold Shutdown Keff < 0.99 and average coolant temperature < 200&deg;F and all reactor vessel head closure bolts fully tensioned 6     Refuelingq One or more reactor vessel head closure bolts less than fully tensioned NM No Mode Reactor vessel contains no irradiated fuel The plant operating mode that exists at the time that the event occurs (prior to any protective system or operator action being initiated in response to the condition) should be compared to the mode applicability of the EALs. If a lower or higher plant operating mode is reached before the emergency classification is made, the declaration shall be based on the mode that existed at the time the event occurred.
I[Document No.]                         IRev.         0            IPage                        8 of 240

3.0     GUIDANCE ON MAKING EMERGENCY CLASSIFICATIONS 3.1     General Considerations When making an emergency classification, the Emergency Coordinator must consider all information having a bearing on the proper assessment of an Initiating Condition (IC3). This includes the Emergency Action Level (EAL) plus the associated Operating Mode Applicability, Notes, and the informing basis information. In the Recognition Category F matrices, EALs are based on loss or potential loss of Fission Product Barrier Thresholds.
3.1.1    Classification Timeliness NRC3 regulations require the licensee to establish and maintain the capability to assess, classify, and declare an emergency condition within 15 minutes after the availability of indications to plant operators that an emergency action level has been exceeded and to promptly declare the emergency condition as soon as possible following identification of the appropriate emergency classification level. The NRC3 staff has provided guidance, on implementing this requirement in NSIR/DPR-ISG-01, "Interim Staff Guidance, Emergency Planning for Nuclear Power Plants" (ref. 4.1.9).
3.1.2 Valid Indications All emergency classification assessments shall be based upon valid indications, reports or conditions. A valid indication, report, or condition, is one that has been verified through appropriate means such that there is no doubt regarding the indicator's operability, the condition's existence, or the report's accuracy.
An indication, report, or condition is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment.
3.1 .3 Imminent Conditions For 1I3s and EALs that have a stipulated time duration (e.g., 15 minutes, 30 minutes, etc.), the Emergency Coordinator should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the condition has exceeded, or will likely exceed, the applicable time. If an ongoing radiological release is detected and the release start time is unknown, it should be assumed that the release duration specified in the 1IC/EAL has been exceeded, absent data to the contrary.
3.1.4 Planned vs. Unplanned Events A planned work activity that results in an expected event or condition which meets or exceeds an EAL does not warrant an emergency declaration provided that: 1) the activity proceeds as planned, and 2) the plant remains within the limits imposed by the operating license. Such activities include planned work to test, manipulate, repair, maintain or modify a system or component. In these cases, the controls associated with the planning, preparation and execution of the work will ensure that compliance is maintained with all aspects of the operating license provided that the activity proceeds and concludes as expected. Events or conditions of this type may be subject to the reporting requirements of 10 &sect; CFR 50.72 (ref.
4.1.4).
I[Document No.]                        *.          Rev. 0                                 Page 9 of 240

3.1 .5 Classification Based on Analysis The assessment of some EALs is based on the results of analyses that are necessary to ascertain whether a specific EAL threshold has been exceeded (e.g., dose assessments, chemistry sampling, RCS leak rate calculation, etc.). For these EALs, the EAL wording or the associated basis discussion will identify the necessary analysis. In these cases, the 15-minute declaration period starts with the availability of the analysis results that show the threshold to be exceeded (i.e., this is the time that the EAL information is first available). The NRC expects licensees to establish the capability to initiate and complete EAL-related analyses within a reasonable period of time (e.g., maintain the necessary expertise on-shift).
3.1.6 Emergency Coordinator Judgment While the EALs have been developed to address a full spectrum of possible events and conditions which may warrant emergency classification, a provision for classification based on operator/management experience and judgment is still necessary. The NEI 99-01 EAL scheme provides the Emergency Coordinator with the ability to classify events and conditions based upon judgment using EALs that are consistent with the ECL definitions (refer to Category H).
The Emergency Coordinator will need to determine if the effects or consequences of the event or condition reasonably meet or exceed a particular ECL definition. A similar provision is incorporated in the Fission Product Barrier Tables; judgment may be used to determine the status of a fission product barrier.
3.2     Classification Methodology To make an emergency classification, the user will compare an event or condition (i.e., the relevant plant indications and reports) to an EAL(s) and determine if the EAL has been met or exceeded. The evaluation of an EAL must be consistent with the related Operating Mode Applicability and Notes. If an EAL has been met or exceeded, the associated IC is likewise met, the emergency classification process "clock" starts, and the ECL must be declared in accordance with plant procedures no later than fifteen minutes after the process "clock" started.
When assessing an EAL that specifies a time duration for the off-normal condition, the "clock" for the EAL time duration runs concurrently with the emergency classification process "clock."
For a full discussion of this timing requirement, refer to NSlRIDPR-ISG-01 (ref. 4.1.9).
3.2.1   Classification of Multiple Events and Conditions When multiple emergency events or conditions are present, the user will identify all met or exceeded EALs. The highest applicable ECL identified during this review is declared. For example:
* If an Alert EAL and a Site Area Emergency EAL are met, whether at one unit or at two different units, a Site Area Emergency should be declared.
There is no "additive" effect from multiple EALs meeting the same ECL. For example:
* If two Alert EALs are met, whether at one unit or at two different units, an Alert should be declared.
Related guidance concerning classification of rapidly escalating events or conditions is provided in Regulatory Issue Summary (RIS) 2007-02, Clarificationof NRC Guidance for Emergency Notifications During Quickly Changing Events (ref. 4.1.2).
[Document No.]                                    Rev. 0                               Page 10. of 240

3.2.2 Consideration of Mode Changes During Classification The mode in effect at the time that an event or condition occurred, and prior to any plant or operator response, is the mode that determines whether or not an IC is applicable. If an event or condition occurs, and results in a mode change before the emergency is declared, the emergency classification level is still based on the mode that existed at the time that the event or condition was initiated (and not when it was declared). Once a different mode is reached, any new event or condition, not related to the original event or condition, requiring emergency classification should be evaluated against the ICs and EArs applicable to the operating mode at the time of the new event or condition.
For events that occur in Cold Shutdo~vn or Refueling, escalation is via EALs that are appliq*able in the Cold Shutdown or Refueling modes, even if Hot Shutdown (or a higher mode) is entered during the subsequent plant response. In particular, the fission product barrier EALs are applicable only to events that initiate in the Hot Shutdown mode or higher.
3.2.3 Classification of Imminent Cornditions Although EALs provide specific thresholds, the Emergency Coordinator must remain alert to events or conditions that could lead to meeting or exceeding an EAL within a relatively short period of time (i.e., a change in the ECL is IMMINENT). If, in the judgment of the Emergency Coordinator, meeting an EAL is IMMINENT, the emergency classification should be made as if the EAL has been met. While applicable to all emergency classification levels, this approach is particularly important at the higher emergency classification levels since it provides additional time for implementation of protective measures.
3.2.4 Emergency Classification Level Upgrading and Downgrading An ECL may be downgraded when the event or condition that meets the highest IC and EAL no longer exists, and other site-specific downgrading requirements are met. If downgrading the ECL is deemed appropriate, the new ECL would then be based on a lower applicable IC(s) and EAL(s). The ECL may also simply be terminated.
As noted above, guidance concerning classification of rapidly escalating events or conditions is provided in RIS 2007-02 (ref. 4.1.2).
3.2.5 Classification of Short-Lived Events Event-based ICs and EALs define a variety of specific occurrences that have potential or actual safety significance. By their nature, some of these events may be short-lived and, thus, over before the emergency classification assessment can be completed. If an event occurs that meets or exceeds an EAL, the associated ECL must be declared regardless of its continued presence at the time of declaration. Examples of such events include an earthquake or a failure of the reactor protection system to automatically trip the reactor followed by a successful manual trip.
3.2.6 Classification of Transient Conditions Many of the ICs and/or EALs employ time-based criteria. These criteria will require that the IC/EAL conditions be present for a defined period of time before an emergency declaration is warranted. In cases where no time-based criterion is specified, it is recognized that some transient conditions may cause an EAL to be met for a brief period of time (e.g., a few seconds to a few minutes). The following guidance should be applied to the classification of these conditions.
I[Document No.]                                   Rev. 0                                Page11o24

EAL    momentarily met during           met durinq an expected         expected (normal)        plant response - In instances where an EAL is briefly plant response, an emergency declaration is not warranted provided that associated systems and components are operating as expected, and operator actions are performed in accordance with procedures.
EAL momentarily met but the condition is corrected prior to an emergency declaration     -  Ifan operator takes prompt manual action to address a condition, and the action is successful in correcting the condition prior to the emergency declaration, then the applicable EAL is not considered met and the associated emergency declaration is not required. For illustrative purposes, consider the following example:
An ATWS occurs and the high pressur* ECCS systems fail to automatically start.
Reactor vessel level rapidly decreases and the plant enters an inadequate core cooling condition (a potential loss of both the Fuel Clad and RCS Barriers). If an operator manually starts a high pressure ECCS system in accordance with an EOP step and clears the inadequate core cooling condition prior to an emergency declaration, then the classification should be based on the ATWS only.
It is important to stress that the 15-minute emergency classification assessment period (process clock) is not a "grace period" during which a classification may be delayed to allow the performance of a corrective action that would obviate the need to classify the event.
Emergency classification assessments must be deliberate and timely, with no undue delays.
The provision discussed above addresses only those rapidly evolving situations when an operator is able to take a successful corrective action prior to the Emergency Coordinator completing the review and steps necessary to make the emergency declaration. This provision is included to ensure that any public protective actions resulting from the emergency classification are truly warranted by the plant conditions.
3.2.7 After-the-Fact Discovery of an Emergency Event or Condition In some cases, an EAL may be met but the emergency classification was not made at the time of the event or condition. This situation can occur when personnel discover that an event or condition existed which met an EAL, but no emergency was declared, and the event or condition no longer exists at the time of discovery. This may be due to the event or condition not being recognized at the time or an error that was made in the emergency classification process.
In these cases, no emergency declaration is warranted; however, the guidance contained in NUREG-1 022 (ref. 4.1.3) is applicable. Specifically, the event should be reported to the NRC in accordance with 10 CFR &sect; 50.72 (ref. 4.1.4) within one hour of the discovery of the undeclared event or condition. The licensee should also notify appropriate State and local agencies in accordance with the agreed upon arrangements.
3.2.8 Retraction of an Emergency Declaration Guidance on the retraction of an emergency declaration reported to the NRC is discussed in NUREG-1022 (ref. 4.1.3).
S[Document No.]                                    Rev. 0                             Page 12 of 240

==4.0   REFERENCES==

4.1   Developmental 4.1.1    NEI 99-0 1 Revision 6, Methodology for the Development of Emergency Action Levels for Non-Passive Reactors, ADAMS Accession Number ML12326A805 4.1.2  'RIS 2007-02 Clarification of NRC Guidance for Emergency Notifications During Quickly Changing Events, February 2, 2007.
4.1.3    NUREG-1022 Event Reporting Guidelines: 10CFR5O.72 and 50.73 4.1.4    10 &sect; CFR 50.72 Immediate Notification Requirements for Operating Nuclear Power Reactors 4.1.5    10 &sect; CFR 50.73 License Event Report System 4.1.6   Technical Specifications Table 1.1-1 Modes 4.1.7    OP/i1,2 ,3/A/1502/000 Containment Clo~sure Control 4.1.8    Procedure Writer's Manual, Revision 012 4.1.9    NSIR/DPR-ISG-01 Interim Staff Guidance; Emergency Planning for Nuclear Power Plants 4.1.10   Oconee Nuclear Site Emergency Plan 4.1.11   S.D.1.3.5 Shutdown Protection Plan 4.1.12   Duke Energy Physical Security Plan for ONS 4.2   Implementing 4.2.1 RPIOIA/1 000/00 1 Emergency Classification 4.2.2 NEI 99-01 Rev. 6 to ONS EAL Comparison Matrix 4.2.30ONS EAL Matrix
[Document No.]                              Rev. 0                             Page 13 of 240

5.0     DEFINITIONS, ACRONYMS & ABBREVIATIONS 5.1     Definitions (ref. 4.1 .1 except as noted)
Selected terms used in Initiating Condition and Emergency Action Level statements are set in all capital letters (e.g., ALL CAPS). These words are defined terms that have specific meanings as used in this document. The definitions of these terms are provided below.
Alert Events are in progress, or have occurred, which involve an actual or potential substantial degradation of the level of safety of the plant or a security event that involves probable life threatenirqg risk to site personnel or damage to site equipment *ecause of hostile action. Any releases are expected to be small fractions of the EPA Protective Action Guideline exposure levels.
Confinement Barrier The barrier(s) between spent fuel and the environment once the spent fuel is processed for dry storage. As related to the ONS ISFSI, Confinement Boundary is comprised of the DSC (dry shielded canister) shell, inner bottom cover plate, inner top cover plate, siphon & vent block, siphon & vent port cover plate, and the welds that join them together.
Containment Closure The action to secure containment and its associated structures, systems, and components as a functional barrier to fission product release under-all plant conditions up to and including a loss of decay heat removal or fuel handling accident inside containment (ref. 4.1 .11).
As applied to ONS, Containment Closure is established when the requirements of OP/1,2,3/A/1502/000, Containment Closure Control, are met (ref. 4.1.7).
Emergency Action Level (EAL)
A pre-determined, site-specific, observable threshold for an Initiating Condition that, when met or exceeded, places the plant in a given emergency classification level.
Emergency Classification Level (ECL)
One of a set of names or titles established by the US Nuclear Regulatory Commission (NRC) for grouping off-normal events or conditions according to (1) potential or actual effects or consequences, and (2) resulting onsite and offsite response actions. The emergency classification levels, in ascending order of severity, are:
* Unusual Event (UE)
* Alert
* Site Area Emergency (SAE)
* General Emergency (GE)
EPA PAGs Environment Protection Agency Protective Action Guidelines. The EPA PAGs are expressed in terms of dose commitment: 1 Rem TEDE or 5 Rem CDE Thyroid. Actual or projected offsite exposures in excess of the EPA PAGs requires ONS to recommend protective actions for the general public to offsite planning agencies.
I[Document     NO.]                                Rev. 0          IPage                      14 of 240

Explosion A rapid, violent and catastrophic failure of a piece of equipment due to combustion, chemical reaction or overpressurization. A release of steam (from high energy lines or components) or an electrical component failure (caused by short circuits, grounding, arcing, etc.) should not automatically be considered an explosion. Such events require a post-event inspection to determine if the attributes of an explosion are present.
Faulted The term applied to a steam generator that has a steam leak on the secondary side of sufficient size to cause an uncontrolled drop in steam generator pressure or the steam generator to becom~e completely depressurized.
Fire Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred but is NOT required if large quantities of smoke and heat are observed.
Fission Product Barrier Threshold A pre-determined, site-specific, observable threshold indicating the loss or potential loss of a fission product barrier.
Flooding A condition where water is entering a room or area faster than installed equipment is capable of removal, resulting in a rise of water level within the room or area.
General Emergency Events are in progress or have occurred which involve actual or imminent substantial core degradation or melting with potential for loss of containment integrity or hostile actions that result 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.
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 ONS 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 ONS. 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.
[Document No.]                                    Rev.O0                              Page15 &deg;f240O

Imminent The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions.
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.
Impede(d)
Personnel access to a room or area is hindered to an extent that extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., requiring use of protective equipment, such as SCBAs, that is not routinely employed).
Initiating Condition (IC)
An event or condition that aligns with the definition of one of the four emergency classification levels by virtue of the potential or actual effects or consequences.
Intrusion The act of entering without authorization. Discovery of a bomb in a specified area is indication of intrusion into that area by a hostile force.
Maintain Take appropriate action to hold the value of an identified parameter within specified limits.
Normal Levels As applied to radiological IC/EALs, the highest reading in the past twenty-four hours excluding the current peak value.
Owner Controlled Area Area outside the PROTECTED AREA fence that immediately surrounds the plant. Access to this area is generally restricted to those entering on official business.
Projectile An object directed toward a Nuclear Power Plant that could cause concern for its continued operability, reliability, or personnel safety.
Protected Area That part of the nuclear station property consisting of the Reactor, Auxiliary, Turbine, and Service Building and grounds, contained within the owner controlled security fence (ref.
4.1.10).
RCS Intact The RCS should be considered intact when the RCS pressure boundary is in its normal condition for the cold shutdown mode of operation (e.g., no freeze seals or nozzle dams, pressurizer manway and safeties installed).
Reduced Inventory Condition with fuel in the reactor vessel and the level lower than approximately three feet below the reactor vessel flange (RCS level < 50" on LT-5) (ref. 4.1.11 ).
S[Document No.]                                   Rev. 0                              Page 16 of 240

Refueling Pathway The spent fuel pool and/or fuel transfer canal comprise the refueling pathway.
Restore Take the appropriate action required to return the value of an identified parameter to the applicable limits.
Ruptured The condition of a steam generator in which primary-to-secondary leakage is of sufficient magnitude to require a safety injection.
Safety SystemII A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2):I Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:
(1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.
Security Condition Any security event as listed in the approved security contingency plan that constitutes a threat/compromise to site security, threat/risk to site personnel, or a potential degradation to the level of safety of the plant. A security condition does not involve a hostile action.
Site Area Emergency Events are in progress or have occurred which involve actual or likely major failures of plant functions needed for protection of the public or hostile actions that result 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 Guidelines exposure levels beyond the site boundary.
Site Boundary That area, including the PROTECTED AREA, in which Duke Energy has the authority to control all activities including exclusion or removal of personnel and property (1 mile radius) from the center of Unit 2. (ref. 4.1.10).
Unisolable An open or breached system line that cannot be isolated, remotely or locally.
Unplanned A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.
((Document No.]                                   Rev. 0           IPage                       17 &deg;f240

Unusual Event Events are in progress or have occurred which indicate a potential degradation in the level of safety of the plant or indicate a security threat to facility protection has been initiated. No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of safety systems occurs.
Valid An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the condition's existence, or the report's accurac~y is removed, Implicit in this definition is the need for timely assessment.
Visible Damage Damage to a component or structure that is' readily observable without measurements, testing, or analysis. The visual impact of the damage is sufficient to cause concern regarding the operability or reliability of the affected component or structure.
.[Document NO.]                                   Rev. 0                                  Page 18 of 240

5.2      Acronyms and Abbreviations 0F .......................................................................... Degrees Fahrenheit
 &deg;..................................................................................... Degrees AC .......................................................................... Alternating Current AP............................................................. Abnormal Operating Procedure ATWS................................................... Anticipated Transient Without Scram BWST ........................................................... Borated Water Storage Tank CETC ...............................................                         4................. Core Exit Thermocouple CDE............................................................... Committed Dose Equivalent CFR ............................................................. Code of Federal Regulations CMT ...........................................                              ..................................... Containment DBA..................................................................... Design Basis Accident DBE.................................................................. Design Basis Earthquake DC ................................................................................                               Direct Current DSC ..................................................................... Dry Shielded Canister EAL ...........           i........................................................ Emergency Action Level ECCS....................................................... Emergency Core Cooling System ECL........................................................... Emergency Classification Level EOF ........................................................... Emergency Operations Facility EOP........................................................ .Emergency Operating Procedure EPA ........................................................ Environmental Protection Agency ERG.......................................................... Emergency Response Guideline EPIP .............................................. Emergency Plan Implementing Procedure ESF ................................................................ Engineered Safety Feature FAA............................                   ,............................... Federal Aviation Administration FBI............................................................ Federal Bureau of Investigation FEMA ............................................. Federal Emergency Management Agency GE......................................................................... General Emergency HPI..................................................................... High Pressure Injection IC............................................................................ Initiating Condition IPEEE............. Individual Plant Examination of External Events (Generic Letter 88-20)
ISFSI ........................................... Independent Spent Fuel Storage Installation Keff ...................................................... Effective Neutron Multiplication Factor LCO.............................................................. Limiting Condition of Operation
[Document No.]                                                    Rev. 0                     *IPage                             19 of 240

LEO ................................................................. Law Enforcement Center LER...................................................................... Licensee Event Report LOCA ................................................................ Loss of Coolant Accident LWR ....................................................................... Light Water Reactor MPC ......................... Maximum Permissible Concentration/Multi-Purpose Canister mR, mRem, mrem, mREM.................................. milli-Roentgen Equivalent Man MSL ........................................................................... Main Steam Line iMW............................................                      *....................................... Megawatt NEI.................................................................... Nuclear Energy Institute NESP................................................. National Environmental Studies Project SNM ............................................                   !......................................... No Mode NPP....................................................................... Nuclear Power Plant NRC.......................................................... Nuclear Regulatory Commission NORAD..................................... North American Aerospace Defense Command (NO)UE.......................................................... Notification of Unusual Event OBE .............................................................. Operating Basis Earthquake OCA .....................................................  .........             ..... Owner Controlled Area ODCM....................................................... Off-site Dose Calculation Manual ORO........................................................... Offsite Response Organization PA ...............................................................................                     Protected Area PAG................................................................ Protective Action Guideline PRA ............................................................ Probabilistic Risk Assessment PSA .. ....................................................... Probabilistic Safety Assessment PWR............................................................... Pressurized Water Reactor PSIG .......................................................... Pounds per Square Inch Gauge PSW.................................................................. Protected Service Water R .......................................................................................                     Roentgen RCS .................................................................. Reactor Coolant System Rem, rem, REM ................................................... Roentgen Equivalent Man Rep CET........................................... Representative Core Exit Thermocouples RETS ......................................... Radiological Effluent Technical Specifications RPS ................................................................ Reactor Protective System RV................................................................................ Reactor Vessel RVLIS ............................................... Reactor Vessel Level Indicating System I[Document NO.]                                            Rev. 0                                                 Page 20 of 240

SAR ...................................................................... Safety Analysis Report SBO.............................................................................. Station Blackout SCBA..................................................... Self-Contained Breathing Apparatus SG.............................................................................. Steam Generator SLCO............................................................ Selected License Commitment SPDS........................................................ Safety Parameter Display System SRO .................................................................... Senior Reactor Operator TEDE . .. .L..........................................................................Total Effective Dose Equivalent TSC................................................................... Technical Support Center UFSAR.................................................. Updated Final Safety Analysis Report S[Document No.].                                                     Rev. 0                                    Page 2l of 240

6.0   ONS-TO-NEI 99-01 Rev. 6 EAL CROSS-REFERENCE This cross-reference is provided to facilitate association and location of an ONS EAL within the NEI 99-0 1 IC/EAL identification scheme. Further information regarding the development of the ONS EALs based on the NEI guidance can be found in the EAL Comparison Matrix.
ONS           NEI 99-01 Rev. 6 EAL            IC        Example EAL RU1.1           AU1          1, 2 RU1.2          AU1           3 RU2.1          AU2           1 RA1.1          AA1           1 RA1.2          AA1           2 RA1.3          AA1           3 RA1.4          AA1           4 RA2.1          AA2           1 RA2.2          AA2           2 RA2.3          AA2           3 RA3.1          AA3           1 RA3.2          AA3           2 RSI.1          AS1           1 RS1.2          AS1           2 RS1.3          AS1           3 RS2.1          AS2           1 RGI.1           AG1           1 RG1.2           AG1           2 RG1.3           AG1           3 RG2.1           AG2           1 CU1.1           CU1           1 CU1.2           CU1           2 S[Document No.]                                  Rev. 0                            Page 22 of 240

ONS      NEI 99-01 Rev. 6 EAL  IC       Example EAL ICEAL CU2.1     CU2           1 CU3.1     CU3           1 CU3.2     CU3          2 CU4.1     CU4           1 CU5.1     CU5        1,2, 3 CAl.1     CA1           1 CA1.2     CAl          2 CA2.1     CA2           1 CA3.1     CA3         1, 2 CA6.1     CA6           1 CS1.1     CS1          3 CGI.1     CG1          2 FAI.1     FA1           1 FSI.1     FS1           1 FGI.1     FG1           1 HUI.1     HU1           1 HU1.2     HU1          2 HU1.3     HlU1          3 HU2.1     HU2           1 HU3.1     HU3           1 HU3.2     HU3          2 HU3o3     HU3           3 HU3.4     HU3          4 HU4.1     HU4           1 HU4.2     HU4          2
[Document No.]            Rev. 0           Page 23 of 240

ONS        NEI 99-01 Rev. 6 EAL  IC       Example EAL ICEAL HU4.3      HU4           3 HU4.4      HU4           4 HU7.1      HU7           1 HAl.1      HA1           1 HA1.2      HA1           2 HA*5.1     HA5           1 HA6.1      HA6           1 HA7.1      HA7           1 HSl.1      HSl           1 HS6.1      HS6           I HS7.1      HS7           1 HG1.1      HG1           1 HG7.1      HG7           1 SUI.1      SU1           1 SU3.1      SU2           1 SU4.1      SU3           2 SU5.1      SU4        1, 2, 3 SU6.1      SU5           1 SU6.2      SU5           2 SU7.1      SU6        1,2, 3 SU8.1      SU7          1,2 SA1.1      SA1           1 SA3.1      SA2           1 SA6.1      SA5           1 SA9.1      SA9           1 S[Document No.]             Rev. 0            Page 24 of 24.0

ONS      NEI 99-01 Rev. 6 EAL  IC       Example EAL ICEAL SS1.1     SS11 SS2.1     SS8          1 SS6.1     SS5          1 SG1.1     SG1          1 SG1.2     SG8          1 EUI.1     EU1          1
*I[Document No.]           Rev. 0          IPage 25 of 240

7.0   ATTACHMENTS 7.1   Attachment 1, Emergency Action Level Technical Bases 7.2   Attachment 2, Fission Product Barrier Matrix and Ba'sis 7.3   Attachment 3, Safe Operation & Shutdown Areas Tables R-2 & H-2 Bases S[Document No.]                             Rev. 0                         Page 26 of 240

ATTACHMENT 1 EAL Bases Category R - Abnormal Rad Release / Rad Effluent EAL Group:       ANY (EALs in this category are applicable to any plant condition, hot or cold.)
Many EALs are based on actual or potential degradation of fission product barriers because of the elevated potential for offsite radioactivity release. Degradation of fission product barriers though is not always apparent via non-radiological symptoms. Therefore, direct indication of elevated radiological effluents or area radiation levels are appropriate symptoms for emlergency classification.
At lower levels, abnormal radioactivity releases may be indicative of a failure of containment systems or precursors to more significant releases. At higher release rates, offsite radiological conditions may result which require offsite protective actions. Elevated area radiation levels in the plant may also be indicative of the failure of containment systems or preclude access to plant vital equipment necessary to ensure plant safety.
Events of this category pertain to the following subcategories:
: 1. Radiological Effluent Direct indication of effluent radiation monitoring systems provides a rapid assessment mechanism to determine releases in excess of classifiable limits. Projected offsite doses, actual offsite field measurements or measured release rates via sampling indicate doses or dose rates above classifiable limits.
: 2. Irradiated Fuel Event Conditions indicative of a loss of adequate shielding or damage to irradiated fuel may preclude access to vital plant areas or result in radiological releases that warrant emergency classification.
: 3. Area Radiation Levels Sustained general area radiation levels which may preclude access to areas requiring continuous occupancy also warrant emergency classification.
[Document No.]                                  Rev. 0                             Page 27 of 240

ATTACHMENT 1 EAL Bases Category:                          R - Abnormal Rad Levels / Rad Effluent Subcategory:                       1 - Radiological Effluent Initiating Condition:              Release of gaseous or liquid radioactivity greater than 2 times the SLCITS limits for 60 minutes or longer EAL:
RUI.1                Unusual Event Reading on an~y Table R-1 effluent radiation monitor > column "UE"ifor --60 min.
(Notes 1,2, 3)
Note 1:       The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
Note 2:       If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.
Note 3:       If the effluent flow past an effluent monitor is known to have stopped, indicating that the release path is isolated, the effluent monitor reading is no longer VALID for classification purposes.
Table R-.1      Effluent Monitor Classification Thresholds_____]
Release Point            [Monitor [            GE        [       SAE         [     Alert       ]        UE     ]
  =, Unit 112/3 Plant Vent              RIA-45            .......                                            1.41 E+5 cpm 0
Unit 1/2/3 Plant Vent             RIA-46       3.O0E+5 cprm       3.00E+4 cpm       3.00E+3 cpm              ..
  ,m. Liquid Radwaste Discharge        RIA-33            .......                                            4.79E+5 cpm Mode Applicability:
All Definition(s):
None Basis:
The column "UE" release values in Table R-1 represent two times the appropriate SLC and Technical Specification release rate and concentration limits associated with the specified monitors (ref. 1, 2,3, 4, 5,6).
Gaseous Releases Instrumentation that may be used to assess this EAL: (ref. 1):
      . Unit 1/2/3 Plant Vent Noble Gas Low Monitor                -  RIA-45(L)
S[Document No.]                                              Rev. 0                                  Page 28 of 240

ATTACHMENT 1 EAL Bases Liquid Releases Instrumentation that may be used to assess this EAL: (ref. 1):
      *Liquid Radwaste Discharge Monitor     -  RIA-33 (batch release)
This IC addresses a potential decrease in the level of safety of the plant as indicated by a low-level radiological release that exceeds regulatory commitments for an extended period of time (e.g., an uncontrolled release). It includes any gaseous or liquid radiological release, monitored or un-monitored, including those for which a radioactivity discharge permit is normally prepared.        II Nuclear power plants incorporate design features intended to control the release of radioactive effluents to the environment. Further, there are administrative controls established to prevent unintentional releases, and to control and monitor intentional releases. The occurrence of an extended, uncontrolled radioactive release to the environment is indicative of degradation in these features and/or controls.
Radio logical effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.
Classification based on effluent monitor readings assumes that a release path to the environment is established. Ifthe effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.
Releases should not be prorated or averaged. For example, a release exceeding 4 times release limits for 30 minutes does not meet the EAL.
This EAL addresses normally occurring continuous radioactivity releases from monitored gaseous or liquid effluent pathways.
Escalation of the emergency classification level would be via IC RAl.
ONS Basis Reference(s):
: 1. UFSAR Section 11.5, Process and Effluent Radiological Monitoring and Sampling Systems
: 2. Oconee Nuclear Station Units 1, 2 and 30Offsite Dose Calculation Manual
: 3. ONS-SLC 16.11.1 Radioactive Liquid Effluents
: 4. ONS-SLC 16.11.2 Radioactive Gaseous Effluents
: 5. EP-EALCALC-ONS-1401 ONS Radiological Effluent EAL Values, Rev. 0
: 6. Technical Specification Section 5.5.5
: 7. NEI 99-01 AU1
[Document No.]                                 Rev. 0                            Page 29 of 240

ATTACHMENT 1 EAL Bases Category:                    R  - Abnormal Rad Levels / Rad Effluent Subcategory:                 1 - Radiological Effluent Initiating Condition:        Release of gaseous or liquid radioactivity greater than 2 times the SLCITS limits for 60 minutes or longer.
EAL:
RUI.2            Unusual Event Sample analysis for a gaseous or liquid release indicates a concentration or release rate
> 2 x SLC/TS limits for > 60 mmn. (Notes 1, 2)
Note 1:   The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.I Note 2:   If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.
Mode Applicability:
All Definition(s):
None Basis:
This IC addresses a potential decrease in the level of safety of the plant as indicated by a low-level radiological release that exceeds regulatory commitments for an extended period of time (e.g., an uncontrolled release). It includes any gaseous or liquid radiological release, monitored or un-monitored, including those for which a radioactivity discharge permit is normally prepared.
Nuclear power plants incorporate design features intended to control the release of radioactive effluents to the environment. Further, there are administrative controls established to prevent unintentional releases, and to control and monitor intentional releases. The occurrence of an extended, uncontrolled radioactive release to the environment is indicative of degradation in these features and/or controls.
Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.
Releases should not be prorated or averaged. For example, a release exceeding 4 times release limits for 30 minutes does not meet the EAL.
This EAL addresses uncontrolled gaseous or liquid releases that are detected by sample analyses or environmental surveys, particularly on unmonitored pathways (e.g., spills of radioactive liquids into storm drains, heat exchanger leakage in river water systems, etc.).
Escalation of the emergency classification level would be via IC RAI.
[Document No.1                                        Rev. 0             IPage                       30 of 2401

ATTACHMENT 1 EAL Bases ONS Basis Reference(s):
: 1. UFSAR Section 11.5, Process and Effluent Radiological Monitoring and Sampling Systems
: 2. Oconee Nuclear Station Units 1, 2 and 30Offsite Dose Calculation Manual
: 3. ONS-SLO 16.11.1 Radioactive Liquid Effluents
: 4. ONS-SLC 16.11.2 Radioactive Gaseous Effluents
: 5. AD-RP-ALL-2003 Investigation of Unusual Radiological Occurrences
: 6. NEI 99-01 AU1
[Document No.]                            Rev. 0                           Page 31 of 240.

ATTACHMENT 1 EAL Bases Category:                        R - Abnormal Rad Levels / Rad Effluent Subcategory:                     1 - Radiological Effluent Initiating Condition:            Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid ODE EAL:
RA1.1              Alert Reading on any Table R-1 effluent radiation monitor > column "ALERT" for > 15 min.
(Notes 1, 2, 3,4)
Note 1:     The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
Note 2:     If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.
Note 3:     If the effluent flow past an effluent monitor is known to have stopped, indicating that the release path is isolated, the effluent monitor reading is no longer VALID for classification purposes.
Note 4      The pre-calculated effluent monitor values presented in EALs RAI.1, RSI.1 and RGI.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.
Table R-1       Effluent Monitor Classification Thresholds                 _____
        - Release Paint            ]Monitor              GE        (       SAE        ]      Alert       ]        UE
  = Unit 1/2/3 Plant Vent              RIA-45            .........                                          1.41 E+5 cpm
    *=Unit 1/2/3 Plant Vent            RIA-46       3.00E+5 cpm        3.00E+4 cpm       3.00E+3 cpm              ..
  *- Liquid Radwaste Discharge         RIA-33            .........                                           4.79E+5 cpm Mode Applicability:
All Definition(s):
None Basis:
This EAL addresses gaseous radioactivity releases, that for whatever reason, cause effluent radiation monitor readings corresponding to site boundary doses that exceed either:
* 10mRem TEDE
* 50 mRem ODE Thyroid I[Document No.]                                             Rev. 0           IPage                          32 of 240I

ATTACHMENT 1 EAL Bases The column "ALERT" gaseous effluent release values in Table R-1 correspond to calculated doses of 1% (10% of the SAE thresholds) of the EPA Protective Action Guidelines (TEDE or ODE Thyroid) (ref. 1, 2, 3, 4).
Instrumentation that may be used to assess this EAL: (ref. 1):
* Unit 1/2/3 Plant Vent Noble Gas Medium Monitor- RIA-46(M)
This IC addresses a release of gaseous or liquid radioactivity that results in projected or actual offsite doses greater than or equal to 1% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude represent an actual or potential substantial degradation of the level of safety of the plant as indicated by a radiological release that significantly exceeds regulatory limits (e.g., a significant uncontrolled release).
Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.
The TEDE dose is set at 1% of the EPA PAG of 1,000 mrem while the 50 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid ODE.
Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.
Escalation of the emergency classification level would be via IC RSI.
ONS Basis Reference(s):
: 1. UFSAR Section 11.5 Process and Effluent Radiological Monitoring and Sampling Systems
: 2. Oconee Nuclear Station Units 1, 2 and 30Offsite Dose Calculation Manual
: 3. EP-EALCALC-ONS-1401 ONS Radiological Effluent EAL Values, Rev. 0
: 4. SDQA-70400-COM, "Unified RASCAL Interface (URI)"
: 5. NEI 99-01 AA1 I[Document No.]                        I        Rev. 0                               Page 33 of 240I

ATTACHMENT 1 EAL Bases Category:                    R - Abnormal Rad Levels / Rad Effluent Subcategory:                 1 - Radiological Effluent Initiating Condition:        Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid CDE EAL:
RA1.2           Alert Dose assessment using actual meteorology indicates dos~es > 10 mrem TEDE or 50 mrem thyroid ODE at or beyond the SITE BOUNDARY (Note 4)
Note 4:    The pre-calculated effluent monitor values presented in EALs RAIl,; RSI.l and RGI.1 should be used Sfor emergency classification assessments until the results from a dose assessment using actual meteorology are available.
Mode Applicability:
All Definition(s):
SITE BOUNDARY- That area, including the PROTECTED AREA, in which Duke Energy has the authority to control all activities including exclusion or removal of personnel and property (1 mile radius) from the center of Unit 2.
Basis:
Dose assessments are performed by computer-based methods (ref. 1, 2).
This IC addresses a release of gaseous or liquid radioactivity that results in projected or actual offsite doses greater than or equal to 1% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude represent an actual or potential substantial degradation of the level of safety of the plant as indicated by a radiological release that significantly exceeds regulatory limits (e.g., a significant uncontrolled release).
Radiological effluent EAL~s are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.
The TEDE dose is set at 1% of the EPA PAG of 1,000 mrem while the 50 mrem thyroid ODE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid ODE.
Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.
Escalation of the emergency classification level would be via IC RSI.
S[Document Np.]                                       Rev. 0                               Page 34 of 240

ATTACHMENT 1 EAL Bases ONS Basis Reference(s):
: 1. RP/0/A/1000/001 Emergency Classification
: 2. AD-EP-ALL-0202 Emergency Response Offsite Dose Assessment
: 3. NEI 99-01 AA1
((Document No.]                           Rev. 0              Page 35 of 240

ATTACHMENT 1 EAL Bases Category:                    R - Abnormal Rad Levels / Rad Effluent Subcategory:                 1 - Radiological Effluent Initiating Condition:        Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid ODE EAL:
RA1.3            Alert Analysis of a liquid effluent sample indicates a concentration or release ratle that would result in doses > 10 mrem TEDE or 50 mrem thyroid ODE at or beyond the SITE BOUNDARY for 60 mai. of exposure (Notes 1, 2)
Note 1:   The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
Note 2:   If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.
Mode Applicability:
All Definition(s):
SITE BOUNDARY - That area, including the PROTECTED AREA, in which Duke Energy has the authority to control all activities including exclusion or removal of personnel and property (1 mile radius) from the center of Unit 2.
Basis:
Dose assessments based on liquid releases are performed per Offsite Dose Calculation Manual (ref. 1).
This IC addresses a release of gaseous or liquid radioactivity that results in projected or actual offsite doses greater than or equal to 1% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude represent an actual or potential substantial degradation of the level of safety of the plant as indicated by a radiological release that significantly exceeds regulatory limits (e.g., a significant uncontrolled release).
Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.
The TEDE dose is set at 1% of the EPA PAG of 1,000 mrem while the 50 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.
Classification based on effluent monitor readings assumes that a release path to the environment is established. Ifthe effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.
[Document No.]                                       Rev. 0                                  Page.36 of 240I

ATTACHMENT 1 EAL Bases Escalation of the emergency classification level would be via IC RS1.
ONS Basis Reference(s):
: 1. Oconee Nuclear Station Units 1, 2 and 30Offsite Dose Calculation Manual
: 2. NEI 99-01 AA1 S[Document No.]                               Rev. 0                        Page 37 of 24.0

ATTACHMENT 1 EAL Bases Category:                     R - Abnormal Rad Levels / Rad Effluent Subcategory:                  1 - Radiological Effluent Initiating Condition:         Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid CDE EAL:
RA1.4             Alert Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY:
* Closed window dose rates > 10 mR/hr expected to continue for > 60 min.
* Analyses of field survey samples indicate thyroid CDE > 50 mrem for 60 min. of inhalation.
(Notes 1,2)
Note 1:   The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
Note 2:    If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.
Mode Applicability:
All Definition(s):
SITE BOUNDARY- That area, including the PROTECTED AREA, in which Duke Energy has the authority to control all activities including exclusion or removal of personnel and property (1 mile radius) from the center of Unit 2.
Basis:
AD-EP-ALL-0203 Protocol for the Field Monitoring Coordinator During Emergency Conditions provides guidance for emergency or post-accident radiological environmental monitoring (ref.
1).
This IC addresses a release of gaseous or liquid radioactivity that results in projected or actual offsite doses greater than or equal to 1% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude represent an actual or potential substantial degradation of the level of safety of the plant as indicated by a radiological release that significantly exceeds regulatory limits (e.g., a significant uncontrolled release).
Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.
The TEDE dose is set at 1% of the EPA PAG of 1,000 mrem while the 50 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.
S[Document No.]                                       Rev. 0                                 Page 38 of 240I

ATTACHMENT 1 EAL Bases Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.
Escalation of the emergency classification level would be via IC RSI.
ONS Basis Reference(s):
: 1. AD-EP-ALL-0203 Protocol for the Field Monitoring Coordinator During Emergency Conditions
: 2. NEI 99-01 AA1 S[Document No.1                                 Rev. 0                            Page  39 of 2401

ATTACHMENT 1 EAL Bases Category:                        R - Abnormal Rad Levels / Rad Effluent Subcategory:                     1 - Radiological Effluent Initiating Condition:            Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem thyroid CDE EAL:
RSI.1              Site Area Emergency Reading on any Table R-1 effluent radiation monitor > column "SAE" for ->15 min.
(Notes 1, 2,3, 4)
Note 1:     The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
Note 2:     If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.
Note 3:     If the effluent flow past an effluent monitor is known to have stopped, indicating that the release path is isolated, the effluent monitor reading is no longer VALID for classification purposes.
Note 4:     The pre-calculated effluent monitor values presented in EALs RAI.1, RSI.1 and-RG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.
Table R-1        Effluent Monitor Classification Thresholds_____
__   Release Point            IMonitor I             GE          I     SAE       I       Alert      f        UE
    =Unit 1/2/3 Plant Vent             RIA-45             .........                                         1.41 E+5 cpm t*Unit 1/2/3 Plant Vent            RIA-46        3.00E+5 cpm        3.00E+4 cpm       3.00E+3 cpm--
 "0
  ~.mLiquid Radwaste Discharge         RIA-33             ..........-                                       4.79E+5 cpm Mode Applicability:
All Definition(s):
None I[Do~cument No.]                                             Rev. 0           IPage                         40 of 240

ATTACHMENT 1 EAL Bases Basis:
This EAL addresses gaseous radioactivity releases, that for whatever reason, cause effluent radiation monitor readings corresponding to site boundary doses that exceed either:
* 100mRemTEDE
* 500 mRem ODE Thyroid The column "SAE" gaseous effluent release value in Table R-1 corresponds to calculated doses of 10% of the EPA Protective Action Guidelines (TEDE or ODE Thyroid) (ref. 1, 3).
Instrumentation that may be used to assess this EAL: (ref. 2):
* Unit 1/2/3 Plant Vent Noble Gas Medium Monitor       - RIA-46(M)
This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to 10% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public.
Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.
The TEDE dose is set at 10% of the EPA PAG of 1,000 mrem while the 500 mrem thyroid ODE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid ODE.
Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.
Escalation of the emergency classification level would be via IC RG1.
ONS Basis Reference(s):
I. EP-EALCALC-ONS-1401 ONS Radiological Effluent EAL Values, Rev. 0
: 2. UFSAR Section 11.5 Process and Effluent Radiological Monitoring and Sampling Systems
: 3. SDQA-70400-COM, "Unified RASCAL Interface (URl)"
: 4. NEI 99-01 AS1
[Document No.]                                 Rev. 0                            Page 41of 240O

ATTACHMENT 1 EAL Bases Category:                   R - Abnormal Rad Levels / Rad Effluent Subcategory:                1 - Radiological Effluent Initiating Condition:       Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem thyroid CDE EAL:
Note 4:   The pre-calculated effluent monitor values presented in EALs RAI .1, RS1 .1 and RGI .1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.
Mode Applicability:
All Definition(s):
SITE BOUNDARY- That area, including the PROTECTED AREA, in which Duke Energy has the authority to control all activities including exclusion or removal of personnel and property (1 mile radius) from the center of Unit 2.
Basis:
Dose assessments are performed by computer-based methods (ref. 1, 2).
This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to 10% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public.
Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both planit condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.
The TEDE dose is set at 10% of the EPA PAG of 1,000 mrem while the 500 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.
Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.
Escalation of the emergency classification level would be via IC RG1.
S[Document No.]                                      Rev. 0                                 Page 42 of 240

ATTACHMENT 1 EAL Bases ONS Basis Reference(s):
: 1. RP/0/A/1 000/001 Emergency Classification
: 2. AD-EP-ALL-0202 Emergency Response Offsite Dose Assessment
: 3. NEI 99-01 AS1 l[Document No.]                            Rev. 0   1Page     43 of 240

ATTACHMENT 1 EAL Bases Category:                     R - Abnormal Rad Levels / Rad Effluent Subcategory:                  1 - Radiological Effluent Initiating Condition:         Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem thyroid ODE EAL:
RS1.3             Site Area Emergency Field survey results indicate IEITHER of the following at or beyond the SITE BOUNDARY:
* Closed window dose rates > 100 mR/hr expected to continue for > 60 min.
* Analyses of field survey samples indicate thyroid ODE > 500 mrem for 60 min. of inhalation.
(Notes 1,2)
Note 1:   The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
Note 2:    If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.
Mode Applicability:
All Definition(s):
SITE BOUNDARY- That area, including the PROTECTED AREA, in which Duke Energy has the authority to control all activities including exclusion or removal of personnel and property (1 mile radius) from the center of Unit 2.
Basis:
AD-EP-ALL-0203 Protocol for the Field Monitoring Coordinator During Emergency Conditions provides guidance for emergency or post-accident radiological environmental monitoring (ref.
1).
This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to 10% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public.
Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.
The TEDE dose is set at 10% of the EPA PAG of 1,000 mrem while the 500 mrem thyroid ODE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid ODE.
Escalation of the emergency classification level would be via IC RGI.
r[Document No.]                                        Rev. 0                                  Page 44 of 240

ATTACHMENT 1 EAL Bases ONS Basis Reference(s):
: 1. AD-EP-ALL-0203 Protocol for the Field Monitoring Coordinator During Emergency Conditions
: 2. NEI 99-01 AS1 S[Document No.]                           Rev. 0                         Page 45 of 240I

ATTACHMENT 1 EAL Bases Category:                             R - Abnormal Rad Levels / Rad Effluent Subcategory:                          1 - Radiological Effluent Initiating Condition:                 Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid ODE EAL:
RGI.1                   General Emergency Reading on any Table R-1 effluent radiation monitor > column "GE" for ->15 min.
(Notes 1, 2,3, 4)
Note 1:         The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
Note 2:          If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.
Note 3:          If the effluent flow past an effluent monitor is known to have stopped, indicating that the release path is isolated, the effluent monitor reading is no longer VALID for classification purposes.
Note 4:         The pre-calculated effluent monitor values presented in EALs RAI.1, RSI.1 and RGI.1 should be used for emergency classification assessments until the results from a dose assessment using actual.
Table R-1       Effluent Monitor Classification Thresholds                _____
         -Release       Paint          IMonitor ]            GE        [       SAE       I       Alert       ]        UE
   *oUnit 1/2/3 Plant Vent                 RIA-45            * .....                                            1.41 E+5 cpm uUnit      1/2/3 Plant Vent             RIA-46       3.00E+5 cpm        3.00E+4 cpm       3.00E+3 cpm              ..
  ~.Liquid Radwaste Discharge              RIA-33             ......                                             4.79E+5 cpm Mode Applicability:
All Definition(s):
None Basis:
This EAL addresses gaseous radioactivity releases, that for whatever reason, cause effluent radiation monitor readings corresponding to site boundary doses that exceed either:
* 1000 mRemTEDE
* 5000 mRem ODE Thyroid The column "GE" gaseous effluent release values in Table R-1 correspond to calculated doses of 100% of the EPA Protective Action Guidelines (TEDE or ODE Thyroid) (ref. 1, 3).
S[Document No.]                                                 Rev. 0                                  Page 46 of 240

ATTACHMENT 1 EAL Bases Instrumentation that may be used to assess this EAL: (ref. 2):
* Unit 1/2/3 Plant Vent Noble Gas Medium Monitor        - RIA-46(M)
This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude will require implementation of protective actions for the public.
Radiological effluent EALs are also included to provide a basis for classifying events and cqonditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.
The TEDE dose is set at the EPA PAG of 1,000 mrem while the 5,000 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.
Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.
ONS Basis Reference(s):
: 1. EP-EALCALC-ONS-1 401 ONS Radiological Effluent EAL Values, Rev. 0
: 2. UFSAR Section 11.5 Process and Effluent Radiological Monitoring and Sampling Systems
: 3. SDQA-70400-COM, "Unified RASCAL Interface (URI)"
: 4. NEI 99-01 AG1 S[Document No.]                                Rev. 0                             Page 47 of 240

ATTACHMENT 1 EAL Bases Category:                  R - Abnormal Rad Levels / Rad Effluent Subcategory:               1 - Radiological Effluent Initiating Condition:      Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid CDE EAL:
RG1.2          General Emergency Dose assessment using actual meteorology indicates doses > I1,000 mrem TEDE or 5,000 mrem thyroid CDE at or beyond the SITE BOUNDARY (Note 4)
Note 4:  The pre-calculated effluent monitor values presented in EALs RAI .1, RSI .1 and RGI .1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.
Mode Applicability:
All Definition(s):
SITE BOUNDARY- That area, including the PROTECTED AREA, in which Duke Energy has the authority to control all activities including exclusion or removal of personnel and property (1 mile radius) from the center of Unit 2.
Basis:
Dose assessments are performed by computer-based methods (ref. 1, 2).
This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude will require implementation of protective actions for the public.
Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.
The TEDE dose is set at the EPA PAG of 1,000 mrem while the 5,000 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.
Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.
I[Document No.]                                      Rev. 0                                 Page 48 of 240

ATTACHMENT 1 EAL Bases ONS Basis Reference(s):
: 1. RP/O/A/1 000/00 1 Emergency Classification
: 2. AD-EP-ALL-0202 Emergency Response Offsite Dose Assessment
: 3. NE! 99-01 AG1 S[Document No.]                            Rev. 0            Page 49 of 240

ATTACHMENT 1 EAL Bases Category:                    R  - Abnormal Rad Levels / Rad Effluent Subcategory:                 1 - Radiological Effluent Initiating Condition:        Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid ODE EAL:
RG1.3             General Emergency Field survey results indicate EITHER of the following at or beyond the *ITE BOUNDARY:
* Closed window dose rates > 1,000 mR/hr expected to continue for > 60 min.
* Analyses ofq field survey samples indicate thyroid ODE > 5,000 mrem*for 60 mmn. of inhalation.
(Notes 1,2)
Note 1:    The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
Note 2:    Ifan ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.
Mode Applicability:
All Definition(s):
SITE BOUNDARY- That area, including the PROTECTED AREA, in which Duke Energy has the authority to control all activities including exclusion or removal of personnel and property (1 mile radius) from the center of Unit 2.
Basis:
AD-EP-ALL-0203 Protocol for the Field Monitoring Coordinator During Emergency Conditions provides guidance for emergency or post-accident radiological environmental monitoring (ref.
1).
This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude will require implementation of protective actions for the public.
Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.
The TEDE dose is set at the EPA PAG of 1,000 mrem while the 5,000 mrem thyroid ODE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.
[Document No.]                                        Rev. 0                                 Page 50 of 240

ATTACHMENT 1 EAL Bases ONS Basis Reference(s):
: 1. AD-EP-ALL-0203 Protocol for the Field Monitoring Coordinator During Emergency Conditions
: 2. NEI 99-01 AGI
[Document No.]                            Rev. 0                          Page 51 of 240

ATTACHMENT 1 EAL Bases Category:                  R - Abnormal Rad Levels / Rad Effluent Subcategory:               2 - Irradiated Fuel Event Initiating Condition:      UNPLANNED loss of water level above irradiated fuel EAL:
RU2.1           Unusual Event UNPLANNED water level drop in the REFUELING PATHWAY as indicated by low water level alarm or indicationI AND UNPLANNED rise in corresponding area radiation levels as indicated by any of the following radiation monitors:
* RIA-3 RB Refueling Deck Shield Wall
* RIA-6 Spent Fuel Building Wall
* Portable area monitors on the main bridge or SFP bridge Mode Applicability:
All Definition(s):
UNPLANNED -. A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.
REFUELING PATHWAY- The spent fuel pool and/or fuel transfer canal comprise the refueling pathway.
Basis:
The spent fuel pool low water level alarm setpoint is actuated at -1.8 ft. below normal level (ref.
1). Water level restoration instructions are performed in accordance with Abnormal Operating Procedures (APs) (ref. 2).
The specified radiation monitors are those expected to see increase area radiation levels as a result of a loss of REFUELING PATHWAY inventory (ref. 3). Increasing radiation indications on these monitors in the absence of indications of decreasing water level are not classifiable under this EAL. Radiation levels in the Reactor Building refueling area are monitored by RIA-3. Radiation levels in the Spent Fuel Pool area are monitored RIA-6. When a fuel bridge is being used to handle fuel, radiation levels are monitored by a portable area monitor mounted on the bridge. (ref. 3, 4)
When the spent fuel pool and reactor cavity are connected, there could exist the possibility of uncovering irradiated fuel. Therefore, this EAL is applicable for conditions in which irradiated fuel is being transferred to and from the reactor vessel and spent fuel pool.
S[Document No.]                                 Rev. 0                            Page 52 of 240

ATTACHMENT 1 EAL Bases This IC addresses a decrease in water level above irradiated fuel sufficient to cause elevated radiation levels. This condition could be a precursor to a more serious event and is also indicative of a minor loss in the ability to control radiation levels within the plant. It is therefore a potential degradation in the level of safety of the plant.
A water level decrease will be primarily determined by indications from available level instrumentation. Other sources of level indications may include reports from plant personnel (e.g., from a refueling crew) or video camera observations (if available). A significant drop in the water level may also cause an increase in the radiation levels of adjacent areas that can be detected by monitors in those locations.        I The effects of planned evolutions should be considered. For example, a refueling bridge area radiation monitor reading may increase due to planned evolutions such as lifting of the reactor vessel head or movement of a fuel assembly. Note that this EAL is applicable only in cases where the elevated reading is due to an unplanned loss of water level.
A drop in water level above irradiated fuel within the reactor vessel may be classified in accordance Recognition Category C during the Cold Shutdown and Refueling modes.
Escalation of the emergency classification level would be via IC RA2.
ONS Basis Reference(s):
: 1. OP/I1/A/61 01/009 Alarm Response Guide 1SA-09, A-5; OP/2/A/61 02/009; OP/3/A/61 03/009
: 2. AP/I1-2,3/Ni1700/035 Loss of SPF Cooling and/or Level
: 3. UFSAR Table 12-3 Area Radiation Monitors
: 4. OP/I1,2,3/A/1 502/007, Enclosure 1, Defueling/Refueling Prerequisites
: 5. NEI 99-01 AU2 S[Document No.]                                   Rev. 0                              Page 53 of 240I

ATTACHMENT 1 EAL Bases Category:                 R - Abnormal Rad Levels / Rad Effluent Subcategory:              2 - Irradiated Fuel Event Initiating Condition:     Significant lowering of water level above, or damage to, irradiated fuel EAL:
RA2.1          Alert Uncovery of irradiated fuel in the REFUELING PATHWAY Mode Applicability:
All I Definition(s):
REFUELING PATHWAY- The spent fuel pool and/or fuel transfer canal comprise the refueling pathway.
Basis:
This IC addresses events that have caused imminent or actual damage to an irradiated fuel assembly, or a significant lowering of water level within the spent fuel pool. These events present radiological safety challenges to plant personnel and are precursors to a release of radioactivity to the environment. As such, they represent an actual or potential substantial degradation of the level of safety of the plant.
This EAL escalates from RU2.1 in that the loss of level, in the affected portion of the REFUELING PATHWAY, is of sufficient magnitude to have resulted in uncovery of irradiated fuel. Indications of irradiated fuel uncovery may include direct or indirect visual observation (e.g., reports from personnel or camera images), as well as significant changes in water and radiation levels, or other plant parameters. Computational aids may also be used (e.g., a boil-off curve). Classification of an event using this EAL should be based on the totality of available indications, reports and observations.
While an area radiation monitor could detect an increase in a dose rate due to a lowering of water level in some portion of the REFUELING PATHWAY, the reading may not be a reliable indication of whether or not the fuel is actually uncovered. To the degree possible, readings should be considered in combination with other available indications of inventory loss.
A drop in water level above irradiated fuel within the reactor vessel may be classified in accordance Recognition Category C during the Cold Shutdown and Refueling modes.
Escalation of the emergency classification level would be via IC RSI.
ONS Basis Reference(s):
: 1. AP/1 -2,3/NI1700/035 Loss of SPF Cooling and/or Level
: 2. NEI 99-01 AA2 S[Document No.]                                 Rev. 0                            Page 54 of 240

ATTACHMENT I EAL Bases Category:                R - Abnormal Rad Levels / Rad Effluent Subcategory:             2 - Irradiated Fuel Event Initiating Condition:    Significant lowering of water level above, or damage to, irradiated fuel EAL:
RA2.2          Alert Damage to irradiated fuel resulting in a release of radioactivity AND HIGH alarm on any of the following radiation monitors:
* RIA-3 RB Refueling Deck Shield Wall
* RIA-6 Spent Fuel Building Wall
* RIA-41 Spend Fuel Pool Gas
* RIA-49 RB Gas
* Portable area monitors on the main bridge or SEP bridge Mode Applicability:
All Definition(s):
None Basis:
The specified radiation monitors are those expected to see increase area radiation levels as a result of damage to irradiated fuel. Radiation levels in the Reactor Building refueling area are monitored by RIA-3. Radiation levels in the Spent Fuel Pool area are monitored RIA-6. When a fuel bridge is being used to handle fuel, radiation levels are monitored by a portable area monitor mounted on the bridge. (ref. 1, 2, 3)
The HIGH alarm for RIA-3 (containment area monitor) and RIA-49 (RB gaseous process monitor) corresponds to the setpoints established to assure that 10 CFR 20 limits are not exceeded.
The HIGH alarm setpoint for RIA-6 (SFP bridge area monitor) is designed to make operators aware of increased readings above 10 CFR 20 limits. The HIGH alarm setpoint for RIA-41 (Spent Fuel Pool gaseous atmosphere) is set to alarm if 4 times the limits of 10 CFR 20 are exceeded based upon Xe-I133. RIA-49 monitors the reactor building gas. Portable monitors are established during refueling outages and are located on the main bridge, and the spent fuel pool bridge.
((Document No.].                               Rev. 0                             Page 55 of 240[

ATTACHMENT 1 EAL Bases This IC addresses events that have caused imminent or actual damage to an irradiated fuel assembly, or a significant lowering of water level within the spent fuel pool. These events present radiological safety challenges to plant personnel and are precursors to a release of radioactivity to the environment. As such, they represent an actual or potential substantial degradation of the level of safety of the plant.
This EAL applies to irradiated fuel that is licensed for dry storage up to the point that the loaded storage cask is sealed. Once sealed, damage to a loaded cask causing loss of the CONFINEMENT BOUNDARY is classified in accordance with EAL EU1 .1.
Escalation of th* emergency would be based on either Recognition d~ategory R or C ICs.
This EAL addresses a release of radioactive material caused by mechanical damage to irradiated fuel. Damaging events may include the dropping, bumping or binding of an assembly, or dropping a heavy load onto an assembly. A rise in readings on radiation monitors should be considered in conjunction with in-plant reports or observations of a potential fuel damaging event (e.g., a fuel handling accident).
Escalation of the emergency classification level would be via IC RSI.
ONS Basis Reference(s):
: 1. OP/1/AI6101/O08, Alarm Response Guide 1SA-08 B-9; OP/2/AI6101/O08; OP/3/A/6101/008
: 2. AP/1 ,2,3/A/1 700/018, Abnormal Release of Radioactivity
: 3. OP/1,2,3/A/1502/007, Enclosure 1, Defueling/Refueling Prerequisites
: 4. NEI 99-01 AA2 S[Document No.]                                 Rev. 0                             page 56 of 240

ATTACHMENT 1 EAL Bases Category:                 R - Abnormal Rad Levels / Rad Effluent Subcategory:              2 - Irradiated Fuel Event Initiating Condition:     Significant lowering of water level above, or damage to, irradiated fuel EAL:
RA2.3          Alert Lowering of spent fuel Pool level to -13.5 ft.
Mode Applicability:
All Definition(s):I None Basis:
This EAL is applicable once the post-Fukushima level instrumentation becomes operational on its associated unit.
Post-Fukushima order EA-12-051 required the installation of reliable SFP level indication capable of identifying normal level (Level 1), SFP level 10 ft. above the top of the fuel racks (Level 2) and SEP level at the top of the fuel racks (Level 3).
SFP level instruments 1/2/3SFP0010 (primary) and 011 (backup) measure SFP level relative to normal water level (El. 840 ft.) from  + 1 ft. to -23.5 ft. (El. 816.4 ft).
For ONS Level 2 corresponds to an indicated water level of -13.5 ft. (El. 826.5 ft.) (ref. 1).
This IC addresses events that have caused imminent or actual damage to an irradiated fuel assembly, or a significant lowering of water level within the spent fuel pool. These events present radiological safety challenges to plant personnel and are precursors to a release of radioactivity to the environment. As such, they represent an actual or potential substantial degradation of the level of safety of the plant.
Escalation of the emergency would be based .on either Recognition Category R or C ICs.
Spent fuel pool water level at this value is within the lower end of the level range necessary to prevent significant dose consequences from direct gamma radiation to personnel performing operations in the vicinity of the spent fuel pool. This condition reflects a significant loss of spent fuel pool water inventory and thus it is also a precursor to a loss of the ability to adequately cool the irradiated fuel assembles stored in the pool.
Escalation of the emergency classification level would be via IC RSI.
ONS Basis Reference(s):
: 1. Engineering Change EC 105805 & 105806
: 2. NEI 99-01 AA2
[Document No.]                         I         Rev. 0            IPage                  57of 240

ATTACHMENT 1 EAL Bases Category:                 R - Abnormal Rad Levels / Rad Effluent Subcategory:              2 - Irradiated Fuel Event Initiating Condition:     Spent fuel pool level at the top of the fuel racks EAL:
RS2.1          Site Area Emergency Lowering of spent fuel pool level to -23.5 ft.
Mode Applicability:
All Definition(s):
None Basis:
This EAL is applicable once the post-Fukushima level instrumentation becomes operational on its associated unit.
Post-Fukushima order EA-1 2-051 required the installation of reliable SFP level indication capable of identifying normal level (Level 1), SEP level 10 ft. above the top of the fuel racks (Level 2) and SFP level at the top of the fuel racks (Level 3).
SEP level instruments 1/2/3SFP0010 (primary) and 011 (backup) measure SEP level relative to normal water level (El. 840 ft.) from + 1 ft. to -23.5 ft. (El. 816.4 ft).
For ONS Level 3 corresponds to an indicated water level of -23.5 ft. (El. 816.5 ft.) (ref. 1).
This EAL addresses a significant loss of spent fuel pool inventory control and makeup capability leading to IMMINENT fuel damage. This condition entails major failures of plant functions needed for protection of the public and thus warrant a Site Area Emergency declaration.
It is recognized that this IC would likely not be met until well after another Site Area Emergency IC was met; however, it is included to provide classification diversity.
Escalation of the emergency classification level would be via IC AGI or RG2.
ONS Basis Reference(s):
: 1. Engineering Change EC 105805 & 105806
: 2. NEI 99-01 AS2 i[Document No.]                                  Rev. 0                              Page 58 of 240'

ATTACHMENT 1 EAL Bases Category:                  R - Abnormal Rad Levels/IRad Effluent Subcategory:               2 - Irradiated Fuel Event Initiating Condition:      Spent fuel pool level cannot be restored to at least the top of the fuel racks for 60 minutes or longer EAL:
SRG2.1          General Emergency Spent fuel pool level cannot be restored t*o at least -23.5 ft. for > 60 min. (Note 1)I Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
Mode Applicability:
All Definition(s):
None Basis:
This EAL is applicable once the post-Fukushima level instrumentation becomes operational on its associated unit.
Post-Fukushima order EA-12-051 required the installation of reliable SFP level indication capable of identifying normal level (Level 1), SEP level 10 ft. above the top of the fuel racks (Level 2) and SEP level at the top of the fuel racks (Level 3).
SEP level instruments 1/2/3SFP0010 (primary) and 011 (backup) measure SFP level relative to normal water level (El. 840 ft.) from + 1 ft. to -23.5 ft. (El. 816.4 ft).
For ONS Level 3 corresponds to an indicated water level of -23.5 ft. (El. 816.5 ft.) (ref. 1).
This EAL addresses a significant loss of spent fuel pool inventory control and makeup capability leading to a prolonged uncovery of spent fuel. This condition will lead to fuel damage and a radiological release to the environment.
It is recognized that this IC would likely not be met until well after another General Emergency IC was met; however, it is included to provide classification diversity.
ONS Basis Reference(s):
: 1. Engineering Change EC 105805 & 105806
: 2. NEI 99-01 AG2
- [Document No.]                                      Rev. 0                              Page 59 of 240

ATTACHMENT 1 EAL Bases Category:                  R - Abnormal Rad Levels / Rad Effluent S ubcategory:             3 - Area Radiation Levels Initiating Condition:      Radiation levels that IMPEDE access to equipment necessary for normal plant operations, cooldown or shutdown EAL:
RA3.1           Alert Dose rates > 15 mR/hr in EITHER of the following areas:
* Control Room (RIA-1)
* Central Alarm Station (by survey)
Mode Applicability:
All Definition(s):
IMPEDE(D) - Personnel access to a room or area is hindered to an extent that extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., requiri~ng use of protective equipment, such as SCBAs, that is not routinely employed).
Basis:
Areas that meet this threshold include the Control Room (CR) and the Central Alarm Station (CAS). RIA-1 monitors the Control room for area radiation (ref. 1). The CAS is included in this EAL because of its importance to permitting access to areas required to assure safe plant operations.
There are no permanently installed area radiation monitors in the CAS that may be used to assess this EAL threshold. Therefore, this threshold is evaluated using local radiation survey for this area (ref. 1).
This IC addresses elevated radiation levels in certain plant rooms/areas sufficient to preclude or impede personnel from performing actions necessary to maintain normal plant operation, or to perform a normal plant cooldown and shutdown. As such, it represents an actual or potential substantial degradation of the level of safety of the plant. The Emergency Coordinator should consider the cause of the increased radiation levels and determine if another IC may be applicable.
Escalation of the emergency classification level would be via Recognition Category R, C or F ICs.
ONS Basis Reference(s):
: 1. UFSAR Table 12-3 Area Radiation Monitors
: 2. NEI 99-01 AA3
* [Document No.]                                  Rev. 0.                         Page 60 of 240I

ATTACHMENT 1 EAL Bases Category:                      R - Abnormal Rad Levels / Rad Effluent Subcategory:                   3 - Area Radiation Levels Initiating Condition:          Radiation levels that IMPEDE access to equipment necessary for normal plant operations, cooldown or shutdown EAL:
RA3.2            Alert An UNPLANNED event results in radiation levels tilat prohibit or IMPEDE access to any Table R-2 rooms or areas (Note 5)
Note 5:   Ifthe equipment in the listed room or area was already inoperable or out-of-service before the event occurred, then no emergency classification is warranted.I Ii         Table R-2           Safe Operation &Shutdown Rooms/Areas                                      1 Room/Area                                        Mode Applicability Turbine Building                                                                        1,2, 3 Equipment and Cable Rooms                                                               1,2, 3 Auxiliary Building                                                                   1, 2, 3,4, 5 Reactor Buildings                                                                       3, 4, 5 Mode Applicability:
All Definition(s):
IMPEDE(D) - Personnel access to a room or area is hindered to an extent that extraordinary*
measures are necessary to facilitate entry of personnel into the affected room/area (e.g., requiring use of protective equipment, such as SCBAs, that is not routinely employed).
UNPLANNED - A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.
Basis:
If the equipment in the listed room or area was already inoperable, or out-of-service, before the event occurred, then no emergency should be declared since the event will have no adverse impact beyond that already allowed by Technical Specifications at the time of the event.
The list of plant rooms or areas with entry-related mode applicability identified specify those rooms or areas that contain equipment which require a manual/local action as specified in operating procedures used for normal plant operation, cooldown and shutdown. Rooms or areas in which actions of a contingent or emergency nature would be performed (e.g., an action to address an off-normal or emergency condition such as emergency repairs, corrective measures or emergency operations) are not included. In addition, the list specifies the plant mode(s) during which entry would be required for each room or area (ref. 1).
I[Document No.]                                           Rev. 0                                        Page 61 of 240

ATTACHMENT 1 EAL Bases This IC addresses elevated radiation levels in certain plant rooms/areas sufficient to preclude or impede personnel from performing actions necessary to maintain normal plant operation, or to perform a normal plant cooldown and shutdown. As such, it represents an actual or potential substantial degradation of the level of safety of the plant. The Emergency Coordinator should consider the cause of the increased radiation levels and determine if another IC may be applicable.
For RA3.2, an Alert declaration is warranted if entry into the affected room/area is, or may be, procedlurally required during the plant operating mode in effect at the time of the elevated radiation levels. The emergency classification is not contingent upon whether entry is actually necessary at the time of the increased radiation levels. Access should be considered as impeded if extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., installing temporary shielding, requiring use of non-routine protective equipment, requesting an extension in dose limits beyond normal administrative limits).
An emergency declaration is not warranted if any of the following conditions apply:
* The plant is in an operating mode different than the mode specified for the affected room/area (i.e., entry is not required during the operating mode in effect at the time of the elevated radiation levels). For example, the plant is in Mode 1 when the radiation increase occurs, and the procedures used for normal operation, cooldown and shutdown do not require entry into the affected room until Mode 4.
* The increased radiation levels are a result of a planned activity that includes compensatory measures which address the temporary inaccessibility of a room or area (e.g., radiography, spent filter or resin transfer, etc.).
* The action for which room/area entry is required is of an administrative or record keeping nature (e.g., normal rounds or routine inspections).
* The access control measures are of a conservative or precautionary nature, and would not actually prevent or impede a required action.
* Ifthe equipment in the listed room or area was already inoperable, or out-of-service, before the event occurred, then no emergency should be declared since the event will have no adverse impact beyond that already allowed by Technical Specifications at the time of the event.
Escalation of the emergency classification level would be via Recognition Category R, C or F ICs.
ONS Basis Reference(s):
: 1. Attachment 3 Safe Operation & Shutdown Rooms/Areas Tables R-3 & H-2 Bases
: 2. NEI 99-01 AA3 l[Document No.]                                   Rev. 0                           Page 62 of 240

ATTACHMENT 1 EAL Bases Categlory E - Independent Spent Fuel Storage Installation (IS FSI)
EAL Group: ANY (EALs in this category are applicable to any plant condition, hot or cold.)
An independent spent fuel storage installation (ISESI) is a complex that is designed and constructed for the interim storage of spent nuclear fuel and other radioactive materials associated with spent fuel storage. A significant amount of the radioactive material contained within a canister must escape its packaging and enter the biosphere for there to be a significant environmental effect resulting from an accident involving tthe dry storage of spent nuclear fuel.
A Notification of Unusual Event is declared on the basis of the occurrence of an event of sufficient magnitude that a loaded cask confinement boundary is damaged or violated.
The ONS ISFSI is contained wholly within the plant Protected Area. Therefore a security event related to the ISESI would be applicable to EALs HU1 .1, HA1 .land HSI1.1 I[Document No.]                                 Rev. 0                             Page 63 of 240

ATTACHMENT 1 EAL Bases Category:                 ISFSI S ubcategory:             Confinement Boundary Initiating Condition:     Damage to a loaded cask CONFINEMENT BOUNDARY EAL:
EUI,1          Unusual Event Damage to a loaded canister CONFINEMENT BOUNDARY as indicated by an on-contact radiation reading on the surface of a loaded spent fuel cask > any Table E-1* ISESI dose limit S    Table E-1   ISFSI Dose Limits Location                 24PHB             37PTH             69BTH HSM front bird screen       1,050 mremfhr     1,050 mrem/hr      500 mrem/hr Outside HSM door              40 mrem/hr        4 mrem/hr         4 mrem/hr End shield wall exterior     550 mrem/hr        8 mrem/hr         8 mrem/hr Mode Applicability:
All Definition(s):
CONFINEMENT BOUNDARY - The barrier(s) between spent fuel and the environment once the spent fuel is processed for dry storage. As related to the ONS ISFSI, Confinement Boundary is comprised of the DSC (dry shielded canister) shell, inner bottom cover plate, inner top cover plate, siphon & vent block, siphon & vent port cover plate, and the welds that join them together.
INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI) - A complex that is designed and constructed for the interim storage of spent nuclear fuel and other radioactive materials associated with spent fuel storage.
Basis:
The ONS ISFSI utilizes the NUHOMS System dry spent fuel storage system for dry spent fuel storage.
The Standardized NUHOMS System is a horizontal canister system composed of a steel dry shielded canister (DSC), a reinforced concrete horizontal storage module (HSM), and a transfer cask (TC). The welded DSC provides confinement and criticality control for the storage and transfer of irradiated fuel. The concrete module provides radiation shielding while allowing cooling of the DSC and fuel by natural convection during storage (ref. 1, 2). The ONS ISESI utilizes the 24PHB, 37PTH and 69BTH DSC designs.
I[Document No.]                                Rev. 0         IPage                   64 of 240

ATTACHMENT 1 EAL Bases Confinement boundary is defined as the barrier(s) between areas containing radioactive substances and the environment. Therefore, damage to a confinement boundary must be a confirmed physical breach between the spent fuel and the environment for the TSC.
The Table E-1 values shown are 2 times the limits specified in the ISESI Certificate of Compliance Technical Specifications for radiation external to the applicable loaded DSC (ref.
1,2).
This IC addresses an event that results in damage to the CONFINEMENT BOUNDARY of a storage cask containing spent fuel. It applies to irradiated fuel that is licensed for dry storage beginning at the point that the 15aded storage cask is sealed. The issues of concernP are the creation of a potential or actual release path to the environment, degradation of one or more fuel assemblies due to environmental factors, and configuration changes which could cause challenges in removing the cask or fuel from storage.
The existence of "damage" is determined by radiological survey. The technical specification multiple of "2 times", which is also used in Recognition Category R IC RU1, is used here to distinguish between non-emergency and emergency conditions. The emphasis for this classification is the degradation in the level of safety of the spent fuel cask and not the magnitude of the associated dose or dose rate. It is recognized that in the case of extreme damage to a loaded cask, the fact that the "on-contact" dose rate limit is exceeded may be determined based on measurement of a dose rate at some distance from the cask.
Security-related events for ISFSIs are covered under l~s HU1 and HA1.
ONS Basis Reference(s):
: 1. USNRC Certificate of Compliance for Spent Fuel Storage Casks, No. 1004, Amendment 13, Attachment A, Technical Specifications for Transnuclear, Inc., Standardized NUHOMS Horizontal Modular Storage System
: 2. OSC-8716, Oconee ISFSI Dose Rate Evaluations, Rev. 0 (4/29/05)
: 3. NEI 99-01 E-HU1
{[Document No.]                        1Rev.          0                             Page 65 of 240.1

ATTACHMENT 1 EAL Bases Category C - Cold Shutdown / Refueling System Malfunction EAL Group: Cold Conditions (RCS temperature -<200&deg;F); EALs in this category are applicable only in one or more cold operating modes.
Category C EALs are directly associated with cold shutdown or refueling system safety functions. Given the variability of plant configurations (e.g., systems out-of-service for maintenance, containment open, reduced AC power redundancy, time since shutdown) during these periods, the consequences of any given initiating event can vary greatly. For example, a loss of decay heat removal capability thMt occurs at the end of an extended outage has less significance than a similar loss occurring during the first week after shutdown. Compounding these events is the likelihood that instrumentation necessary for assessment may also be inoperable. The cold shutdown and refueling system malfunction EALs are based on performance capability to the extent possible with consideration given to RCS integrity, containment closure, and fuel clad integrity for the applicable operating modes (5 - Cold Shutdown, 6 - Refueling, NM - No Mode).
The events of this category pertain to the following subcategories:
: 1. RCS Level RCS water level is directly related to the status of adequate core cooling and, therefore, fuel clad integrity.
: 2. Loss of Essential AC Power Loss of essential plant electrical power can compromise plant safety system operability including decay heat removal and emergency core cooling systems which may be necessary to ensure fission product barrier integrity. This category includes loss of onsite and offsite power sources for 41 60V AC essential buses.
: 3. RCS Temperature Uncontrolled or inadvertent temperature or pressure increases are indicative of a potential loss of safety functions.
: 4. Loss of Vital DC Power Loss of emergency plant electrical power can compromise plant safety system operability including decay heat removal and emergency core cooling systems which may be necessary to ensure fission product barrier integrity. This category includes loss of power to or degraded voltage on the 125V DC vital buses.
: 5. Loss of Communications Certain events that degrade plant operator ability to effectively communicate with essential personnel within or external to the plant warrant emergency classification.
: 6. Hazardous Event Affectincq Safety Systems Certain hazardous natural and technological events may result in visible damage to or degraded performance of safety systems warranting classification.
S[Document No.]                                  Rev. 0                            Page 66 of 240I

ATTACHMENT 1 EAL Bases Category:                  C - Cold Shutdown I Refueling System Malfunction Subcategory:               1 - RCS Level Initiating Condition:      UNPLANNED loss of RCS inventory for 15 minutes or longer EAL:
Note 1:   The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
Mode Applicability:
5 - Cold Shutdown, 6 - Refueling Definition(s):
UNPLANNED - A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.
Basis:
RCS water level less than a required lower limit is meant to be less than the lower end of the level control band being procedurally maintained for the current condition or evolution.
RCS water level instrumentation requirements to begin an RCS inventory reduction with fuel in the core to below 80" (lowered inventory) or 50" (reduced inventory) are the following (ref. 1):
* Both channels of LT-5 prior to reducing ROS inventory below 80".
* Both channels of LT-5 and both hot leg and cold leg ultrasonic monitors prior to reducing RCS inventory below 50".
This IC addresses the inability to restore and maintain water level to a required minimum level (or the lower limit of a level band), or a loss of the ability to monitor RCS level concurrent with indications of coolant leakage. Either of these conditions is considered to be a potential degradation of the level of safety of the plant.
Refueling evolutions that decrease RCS water inventory are carefully planned and controlled.
An UNPLANNED event that results in water level decreasing below a procedurally required limit warrants the declaration of an Unusual Event due to the reduced water inventory that is available to keep the core covered.
This EAL recognizes that the minimum required RCS level can change several times during the course of a refueling outage as different plant configurations and system lineups are implemented. This EAL is met if the minimum level, specified for the current plant conditions, cannot be maintained for 15 minutes or longer. The minimum level is typically specified in the applicable operating procedure but may be specified in another controlling document.
S[Document No.]                                       Rev. 0                             Page 67 of 240

ATTACHMENT 1 EAL Bases The 15-minute threshold duration allows sufficient time for prompt operator actions to restore and maintain the expected water level. This criterion excludes transient conditions causing a brief lowering of water level.
Continued loss of RCS inventory may result in escalation to the Alert emergency classification level via either IC CA1 or CA3.
ONS Basis Reference(s):
: 1. S. D. 1.3.5 Shutdown Protection Plan, Section 5.2.7
: 2. NEI 99-01 CU1 S[Document No.]                             Rev. 0                             Page 68 of 240I

ATTACHMENT I EAL Bases Category:                 C - Cold Shutdown / Refueling System Malfunction Subcategory:              1 - RCS Level Initiating Condition:     UNPLANNED loss of RCS inventory for 15 minutes or longer EAL:
CU1.2          Unusual Event RC3S level cannot be monitored AND EITHER
* UNPLANNED increase in any Table C3-I sump/tank level due to loss of RC3S inventory
* Visual observation of UNISOLABLE RCS leakage Table C-I     Sumps / Tanks
* RB Normal Sumps
* RB Emergency Sumps
* Core Flood Tank
* Quench Tank
* Low Activity Waste Tank
* High Activity Waste Tank
* Miscellaneous Waste Holdup Tank
* LPI Room Sumps Mode Applicability:
5 - Cold Shutdown, 6 - Refueling Definition(s):
UNISOLABLE - An open or breached system line that cannot be isolated, remotely or locally.
UNPLANNED - A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.
Basis:
A RC3S leak should be considered UNISOLABLE if the leak cannot be isolated within 15 min.
In Cold Shutdown mode, the RC3S will normally be intact and standard RC3S level monitoring means are available.
In the Refuel mode, the RC3S is not intact and reactor vessel level may be monitored by different means, including the ability to monitor level visually.
In this EAL, all water level indication is unavailable and the RC3S inventory loss must be detected by indirect leakage indications (Table C3-I). Level increases must be evaluated S[Document No.]                                 Rev. 0                           Page 69 of 240

ATTACHMENT 1 EAL Bases against other potential sources of leakage such as cooling water sources inside the containment to ensure they are indicative of RCS leakage. If the make-up rate to the RCS unexplainably rises above the pre-established rate, a loss of RCS inventory may be occurring even if the source of the leakage cannot be immediately identified. Visual observation of significant leakage from systems connected to the RCS that cannot be isolated could also be indicative of a loss of RCS inventory (ref. 1, 2).
This IC addresses the inability to restore and maintain water level to a required minimum level (or the lower limit of a level band), or a loss of the ability to monitor RCS level concurrent with indications of' coolant leakage. Either of these conditions is considered to be a potential degradation of the level of safety of the plant.
Refueling evolutions that decrease RCS water inventory are carefully planned and controlled.
' An UNPLANNED event that results in water level decreasing below a procedurally required limit warrants the declaration of an Unusual Event due to the redu~ced water inventory that is available to keep the core covered.
This EAL addresses a condition where all means to determine level have been lost. In this condition, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels (Table C-I). Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the RCS.
Continued loss of RCS inventory may result in escalation to the Alert emergency classification level via either IC CAl or CA3.
ONS Basis Reference(s):
: 1. AP/1 ,2,3/A/1 700/002 Excessive RCS Leakage
: 2. AP/1 -2,3/NI1700/030 Auxiliary Building Flood
: 3. NEI 99-01 CU1                7 l[Document No.]                                  Rev. 0                              Page 70 of 240I

ATTACHMENT 1 EAL Bases Category:                  C - Cold Shutdown / Refueling System Malfunction Subcategory:               1 - RCS Level Initiating Condition:      Loss of RCS inventory EAL:
CA1.1          Alert Loss of RCS inventory as indicated by RCS level       < 10" (LT-5)
Mode Applicability:
5 - Cold Shutdown, 6 - Refueling Definition(s):           I None Basis:
RCS water level of 10" as indicated on LT-5 is the lowest level for continued operation of LPI pumps for decay heat removal (ref. 1). Two LPI pumps and two coolers normally perform the' decay heat removal function for each unit (ref. 2).
The threshold was chosen because a loss of suction to decay heat removal systems may occur. The inability to restore and maintain level after reaching this setpoint infers a failure of the RCS Barrier.
This IC addresses conditions that are precursors to a loss of the ability to adequately cool irradiated fuel (i.e., a precursor to a challenge to the fuel clad barrier). This condition represents a potential substantial reduction in the level of plant safety.
For this EAL, a lowering of RCS water level below 10 in. indicates that operator actions have not been successful in restoring and maintaining RCS water level. The heat-up rate of the coolant will increase as the available water inventory is reduced. A continuing decrease in water level will lead to core uncovery.
Although related, this EAL is concerned with the loss of RCS inventory and not the potential concurrent effects on systems needed for decay heat removal (e.g., loss of a Decay Heat Removal suction point). An increase in RCS temperature caused by a loss of decay heat removal capability is evaluated under IC CA3.
If RCS water level continues to lower, then escalation to Site Area Emergency would be via IC CSI.
ONS Basis Reference(s):
: 1. AP/1 ,2,3/A/1700/026 Loss of Decay Heat Removal
: 2. UFSAR Section 9.3.3 Low Pressure Injection System
: 3. NEI 99-01 CA1 i[Document No.]                                  Rev. 0                              Page. 71 of 240I

ATTACHMENT 1 EAL Bases Category:                     C - Cold Shutdown / Refueling System Malfunction Subcategory:                  1 - RCS Level Initiating Condition:         Loss of RCS inventory EAL:
CA1.2           Alert RCS level cannot be monitored for >- 15 mai.         (Note 1)
AND EITHER
* UNPLANNED increase in any Table C-i Sump / Tank level due to a loss of RCS inventory
* Visual observation of UNISOLABLE ROS leakage exceeded, or will likely be exceeded.
Table C-I      Sumps I Tanks
* RB Normal Sumps
* RB Emergency Sumps
* Core Flood Tank
* Quench Tank
* Low Activity Waste Tank
* High Activity Waste Tank
* Miscellaneous Waste Holdup Tank
* LPI Room Sumps Mode Applicability:
5 - Cold Shutdown, 6 - Refueling Definition(s):
UNISOLABLE - An open or breached system line that cannot be isolated, remotely or locally.
UNPLANNED - A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.
Basis:
A RCS leak should be considered UNISOLABLE if the leak cannot be isolated within 15 minutes.
In Cold Shutdown mode, the RCS will normally be intact and standard RCS level monitoring means are available.
I[Document No.]                            I         Rev. 0                     Page 72o&deg;240

ATTACHMENT 1 EAL Bases In the Refuel mode, the RCS is not intact and RCS level may be monitored by different means, including the ability to monitor level visually.
In this EAL, all RCS water level indication would be unavailable for greater than 15 minutes, and the RCS inventory loss must be detected by indirect leakage indications (Table C-1). Level increases must be evaluated against other potential sources of leakage such as cooling water sources inside the containment to ensure they are indicative of RCS leakage. If the make-up rate to the RCS unexplainably rises above the pre-established rate, a loss of RCS inventory may be occurring even if the source of the leakage cannot be immediately identified. Visual observation of significant leakage from systems connected to the RCS that cannot be isolated could also be indicative of a loss of RCS inventory (ref. 1, 2).
This IC addresses conditions that are precursors to a loss of the ability to adequately cool irradiated fuel (i.e., a precursor to a ch~allenge to the fuel clad barrier). This condition represents a potential substantial reduction in the level of plant safety.
For this EAL, the inability to monitor RCS level may be caused by instrumentation and/or power failures, or water level dropping below the range of available instrumentation. Ifwater level cannot be monitored, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the RCS.
The 15-minute duration for the loss of level indication was chosen because it is half of the EAL duration specified in IC CSI.
If the RCS inventory level continues to lower, then escalation to Site Area Emergency would be via IC CSI.
ONS Basis Reference(s):
: 1. AP/1 ,2,3/A/1700/002 Excessive RCS Leakage
: 2. AP/1 -2,3/NI1700/030 Auxiliary Building Flood
: 3. NEI 99-01 CA1
((Document No.]                                   Rev. 0                              Page 73 of 240I

ATTACHMENT 1 EAL Bases Category:                       C - Cold Shutdown / Refueling System Malfunction Subcategory:                    1 - RCS Level Initiating Condition:           Loss of RCS inventory affecting core decay heat removal capability EAL:
CSl.1           Site Area Emergency RCS level cannot be monitored for > 30 min. (Note I)
AND Core uncovery is indicated by any of the following:
* UNPLANNED increase in any Table C-I sump/tank level
* Visual observation of UNISOLABLE RCS leakage
* High alarm on RIA-3 RB Refueling Deck Shield Wall
* Erratic Source Range Monitor Indication Note 1:   The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
Table C-I     Sumps / Tanks
* RB Normal Sumps
* RB Emergency Sumps
* Core Flood Tank
* Quench Tank
* Low Activity Waste Tank
* High Activity Waste Tank
* Miscellaneous Waste Holdup Tank
* LPI Room Sumps Mode Applicability:
5 -  Cold Shutdown, 6 - Refueling Definition(s):
UNISOLABLE - An open or breached system line that cannot be isolated, remotely or locally.
UNPLANNED - A parameter change or an event that is not I) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.
Basis:
A RCS leak should be considered UNISOLABLE if the leak cannot be isolated within 15 minutes.
[Dcmn o]Rev.                                        0                                     Page 74 of 240

ATTACHMENT 1 EAL Bases In Cold Shutdown mode, the RCS will normally be intact and standard RCS level monitoring means are available.
In the Refueling mode, the RCS is not intact and RCS level may be monitored by different means, including the ability to monitor level visually.
In this EAL, all RCS water level indication would be unavailable for greater than 30 minutes, and the RCS inventory loss must be detected by indirect leakage indications (Table C-I). Level increases must be evaluated against other potential sources of leakage such as cooling water sources inside the containment to ensure they are indicative of RCS leakage. If the make-up rafe to the RCS unexplainably rises above the pre-established rate, a loss of RCS inventory may be occurring even if the source of the leakage cannot be immediately identified. Visual observation of significant leakage from systems connected to the RCS that cannot be isolated could*also be indicative of a loss of RCS inventory (ref.! 1, 2).
Sump or tank level increases should be of a magnitude that correlates to a volume sufficient to indicate fuel has been uncovered or uncovery is imminent.
The Reactor Vessel inventory loss may be detected by a reduction in water shielding that causes a high alarm on the Refueling Deck Shield Wall area radiation monitor (ref. 3).
Post-TM I accident studies indicated that the installed PWR nuclear instrumentation will operate erratically when the core is uncovered and that this should be used as a tool for making such determinations (ref. 4, 5, 6).
This IC addresses a significant and prolonged loss of reactor vessel/ROS inventory control and makeup capability leading to IMMINENT fuel damage. The lost inventory may be due to a RCS component failure, a loss of configuration control or prolonged boiling of reactor coolant.
These conditions entail major failures of plant functions needed for protection of the public and thus warrant a Site Area Emergency declaration.
Following an extended loss of core decay heat removal and inventory makeup, decay heat will cause reactor coolant boiling and a further reduction in reactor vessel level. If RCS level cannot be restored, fuel damage is probable.
The 30-minute criterion is tied to a readily recognizable event start time (i.e., the total loss of ability to monitor level), and allows sufficient time to monitor, assess and correlate reactor and plant conditions to determine if core uncovery has actually occurred (i.e., to account for various accident progression and instrumentation uncertainties). It also allows sufficient time for performance of actions to terminate leakage, recover inventory control/makeup equipment and/or restore level monitoring.
The inability to monitor RCS level may be caused by instrumentation and/or power failures, or water level dropping below the range of available instrumentation. Ifwater level cannot be monitored, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the RCS.
This EAL addresses concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal; SECY 91-283, Evaluation of Shutdown and Low Power Risk Issues; NUREG-1 449, 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.
S[Document.No.]                                 Rev. 0                             Page 75 of 240

ATTACHMENT 1 EAL Bases Escalation of the emergency classification level would be via IC OG1 or RG1.
ONS Basis Reference(s):
: 1. API/1,2,3/A/1700/002 Excessive RCS Leakage
: 2. AP/1-2,3/A/1700/030 Auxiliary Building Flood
: 3. UFSAR Table 12-3 Area Radiation Monitors
: 4. UFSAR Section 7.4.1 Nuclear Instrumentation
: 5. OP/1,2,3/A/5102/002 Alarm Response Guide 1,2,3SA-02, A-6
: 6. Nuclear Safety Analysis Center (NSAC), 1980, "Analysis of Three Mile Island  - Unit 2 Accident," NSAC-1
: 7. NEI 99-01 CS1 S[Document No.]                               Rev. 0                          Page 76 of 240

ATTACHMENT 1 EAL Bases Category:                   C - Cold Shutdown / Refueling System Malfunction Subcategory:                1 - RCS Level Initiating Condition:       Loss of RCS inventory affecting fuel clad integrity with containment challenged EAL:
CGI.1           General Emergency RCS level cannot be monitored for ->30 min. (Note 1)
AND Core uncovery is indicated by any of the following:
* UNPLANNED increase in any Table C-i sump/tank level due to a loss of RCS inventory
* Visual observation of UNISOLABLE RCS leakage
* High alarm on RIA-3 RB Refueling Deck Shield Wall
* Erratic Source Range Monitor Indication AND Any Containment Challenge indication, Table C-2 Note 1:   The been Emergency exceeded, Coordinator    should declare the event promptly upon determining that time limit has or will likely be exceeded.
Note 6:   If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, declaration of a General Emergency is not required.
Table C-I       Sumps I Tanks
* RB Normal Sumps
* RB Emergency Sum ps
* Core Flood Tank
* Quench Tank
* Low Activity Waste Tank
* High Activity Waste Tank
* Miscellaneous Waste Holdup Tank
* LPI Room Sumps I[Document No.]                                        Rev. 0                                 Page 77 of 240

ATTACHMENT 1 EAL Bases TablC-2   Containment Challenge TablC-2    Indications
* CONTAINMENT CLOSURE not established (Note 6)
* Containment hydrogen concentration
                                  > 4%
                                *,Un planned rise in containment pressure Mode Applicability:
5 - Cold Shutdown, 6 - Refueling Definition(s):
CONTAINMENT CLOSURE - The action to secure containment and its associated structures, systems, and components as a functional barrier to fission product release under all plant conditions up to and including a loss of decay heat removal or fuel handling accident inside containment.
As applied to ONS, Containment Closure is established when the requirements of OP/i1,2,3/A/1502/009, Containment Closure Control, are met.
UNISOLABLE - An open or breached system line that cannot be isolated, remotely or locally.
UNPLANNED - A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the paraimeter change or event may be known or unknown.
Basis:
A RCS leak should be considered UNISOLABLE if the leak cannot be isolated within 15 minutes.
In Cold Shutdown mode, the RCS will normally be intact and standard RCS level monitoring means are available.
In the Refueling mode, the RCS is not intact and RCS level may be monitored by different means, including the ability to monitor level visually.
In this EAL, all RCS water level indication would be unavailable for greater than 30 minutes, and the RCS inventory loss must be detected by indirect leakage indications (Table C-I). Level increases must be evaluated against other potential sources of leakage such as cooling water sources inside the containment to ensure they are indicative of RCS leakage. If the make-up rate to the RCS unexplainably rises above the pre-established rate, a loss of RCS inventory may be occurring even if the source of the leakage cannot be immediately identified. Visual observation of significant leakage from systems connected to the RCS that cannot be isolated could also be indicative of a loss of RCS inventory (ref. 1, 2).
S[Document No.]                                 Rev. 0                           Page 78 of 2401

ATTACHMENT 1 EAL Bases Sump or tank level increases should be of a magnitude that correlates to a volume sufficient to indicate fuel has been uncovered or uncovery is imminent.
The Reactor Vessel inventory loss may be detected by a reduction in water shielding that causes a high alarm on the Refueling Deck Shield Wall area radiation monitor (ref. 3).
Post-TMI accident studies indicated that the installed PWR nuclear instrumentation will operate erratically when the core is uncovered and that this should be used as a tool for making such determinations (ref. 4, 5, 6).
Three conditions are associated witl] a challenge to Containment integrity:
: 1. CONTAINMENT CLOSURE not established - The status of containment closure is tracked if plant conditions change that could raise the risk of a fission product release as a result of a loss of decay heat removal (ref. 7). If containment closure is re-established prior to exceeding the 30 minute core uncovery time limit then escalation to GE would not occur.
: 2. Containment hydrogen > 4% - The 4% hydrogen concentration threshold is generally considered the lower limit for hydrogen combustion. ONS is equipped with a Containment Hydrogen Monitoring System (CHMS) that provides continuous indication of hydrogen concentration in the containment atmosphere. The measurement capability is provided over the range of 0% to 10%. A continuous indication of the hydrogen concentration is not required in the control room at all times during normal operation. If continuous indication of the hydrogen concentration is not available at all times, continuous indication and recording shall be functioning within 90 minutes of the initiation of the safety injection. (ref. 8, 9)
: 3. UNPLANNED rise in containment pressure - An unplanned pressure rise in containment while in cold shutdown or refueling modes can threaten Containment Closure capability and thus containment potentially cannot be relied upon as a barrier to fission product release.
This IC addresses the inability to restore and maintain reactor vessel level above the top of active fuel with containment challenged. This condition represents actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity.
Releases can be reasonably expected to exceed EPA PAG exposure levels offsite for more than the immediate site area.
Following an extended loss of core decay heat removal and inventory makeup, decay heat will cause reactor coolant boiling and a further reduction in reactor vessel level. If RCS level cannot be restored, fuel damage is probable.
With CONTAINMENT CLOSURE not established, there is a high potential for a direct and unmonitored release of radioactivity to the environment. If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, then declaration of a General Emergency is not required.
The existence of an explosive mixture means, at a minimum, that the containment atmospheric hydrogen concentration is sufficient to support a hydrogen burn (i.e., at the lower deflagration limit). A hydrogen burn will raise containment pressure and could result in collateral equipment damage leading to a loss of containment integrity. It therefore represents a challenge to Containment integrity.
S[Document No.]                                    Rev. 0                          Page 79 of 240

ATTACHMENT 1 EAL Bases In the early stages of a core uncovery event, it is unlikely that hydrogen buildup due to a core uncovery could result in an explosive gas mixture in containment. If all installed hydrogen gas monitors are out-of-service during an event leading to fuel cladding damage, it may not be possible to obtain a containment hydrogen gas concentration reading as ambient conditions within the containment will preclude personnel access. During periods when installed containment hydrogen gas monitors are out-of-service, operators may use the other listed indications to assess whether or not containment is challenged.
The 30-minute criterion is tied to a readily recognizable event start time (i.e., the total loss of ability to monitor level), and allows sufficient time to monitor, assess and correlate reactor1 and plant conditions to determine if core uncovery has actually occurred (i.e., to account for various accident progression and instrumentation uncertainties). It also allows sufficient time for performance of actions to terminate leakage, recover inventory control/makeup equipment and/or restore level monitoring.       I The inability to monitor RCS level may be caused by instrumentation and/or power failures, or water level dropping below the range of available instrumentation. Ifwater level cannot be monitored, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the RCS.
This EAL addresses concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal; SECY 91-283, Evaluation of Shutdown and Low Power Risk Issues; NUREG-1 449, 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.
ONS Basis Reference(s):
: 1. AP/1 ,2,3/A/1 700/002 Excessive RCS Leakage
: 2. AP/1-2,3/A/1700/030 Auxiliary Building Flood
: 3. UFSAR Table 12-3 Area Radiation Monitors
: 4. UFSAR Section 7.4.1 Nuclear Instrumentation
: 5. OP/1/A/6101/002; OP/2/A/6102/002; OP/3/A/6103/002 Alarm Response Guide 1,2,3SA-02, A-6
: 6. Nuclear Safety Analysis Center (NSAC), 1980, "Analysis of Three Mile Island         - Unit 2 Accident," NSAC-1
: 7. OP/i1,2,3/A/1 502/009 Containment Closure Control
: 8. UFSAR Section 9.3.7 Containment Hydrogen Monitoring System
: 9. UFSAR Section 15.16.3 Evaluation of Hydrogen Concentrations
: 10. NEI 99-01 CG1 S[Document No.]                                 Rev. 0                             Page 80 of 240

ATTACHMENT 1 EAL Bases Category:                  C - Cold Shutdown / Refueling System Malfunction Subcategory:               2 - Loss of Essential AC Power Initiating Condition:      Loss of all but one AC power source to essential buses for 15 minutes or longer EAL:
CU2.1           Unusual Event AC power capability, Table C-3, to essential 4160 V buses MFB-1 and MFB-2 reduced to a single power source for > 15 min. (Note 1)
AND Any additional single power source failure wil result in loss of all AC power to SAFETY SYSTEMS Note 1:   The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
Table C-3 AC Power Sources Offsite:
* Unit Normal Transformer (backcharged)
* Unit Startup Transformer (SWYD)
* Another Unit Startup Transformer (aligned) (SWYD)
* CT5 (Central/energizing Standby Bus)
Emergency:
* Unit Startup Transformer (Keowee)
* Another Unit Startup Transformer (aligned) (Keowee)
* CT4
* CT5 (dedicated line/energizing Standby Bus)
Mode Applicability:
5 - Cold Shutdown, 6 - Refueling, NM           - No Mode Definition(s):
SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2):
Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:
(1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition;*
S[Document No.]                                      Rev. 0                             Page 81 of 240

ATTACHMENT 1 EAL Bases (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.
Basis:
For emergency classification purposes, "capability" means that an AC power source is available to and capable of powering the emergency bus(es) within 15 min, whether or not the buses are currently powered from it.
The condition indicated by this EAL is the degradation of the offsite and onsite power sources such that any additional single failure would result in a loss of all AC power to the emergency b~uses. 4160 V buses MFB-1 and MFB-2 are the essential buses (ref. 1).
Each unit is provided with two physically independent circuits from the switching station. One is the circuit from the 230 kV switching station through the startup transformer, which is designed to be available within a few seconds following a loss Of coolant accident. The second circuit is the path from the switchyard through the main step-up transformer, the main generator bus and the unit auxiliary transformer with the generator disconnected from the main bus. The second circuit is currently used during refueling as an additional power feed for the shutdown unit(s) from the 230 kV switchyard. Whenever there is inadequate power from the generating units, the 230 kV switching station and the hydro units, power is available to the standby power buses either directly from the 100 kV Central Tie Substation or from Lee Steam Station via a Transformer CT5 (ref. 2).
Upon loss of power from the Oconee generating unit and 230 kV switchyard, power is supplied from both Keowee Hydro Station units through two separate and independent routes. One route is an underground feeder to Transformer CT4 which supplies the two redundant Main Feeder Buses (MFB-1 and MFB-2). The other route is an overhead feeder to the 230 KV switching station which supplies each unit startup transformer (ref. 3).
The Standby Shutdown Facility (SSF) consists of standby systems for use in an extreme emergency and is equipped with a manually started diesel generator that can supply power necessary to maintain hot shutdown of the reactors of each unit in the event of loss of power.
However, the SSF cannot supply power to the essential buses and therefore not credited in this EAL (ref. 3).
This cold condition EAL is equivalent to the hot condition EAL SAlI.1.
This IC describes a significant degradation of offsite and onsite AC power sources such that any additional single failure would result in a loss of all AC power to SAFETY SYSTEMS. In this condition, the sole AC power source may be powering one, or more than one, train of safety-related equipment.
When in the cold shutdown, refueling, or no mode, this condition is not classified as an Alert because of the increased time available to restore another power source to service. Additional time is available due to the reduced core decay heat load, and the lower temperatures and pressures in various plant systems. Thus, when in these modes, this condition is considered to be a potential degradation of the level of safety of the plant.
An "AC power source" is a source recognized in APs and EOPs, and capable of supplying required power to an essential bus. Some examples of this condition are presented below.
* A loss of all offsite power with a concurrent failure of all but one essential power source S[Docume~nt No.]                                 Rev. 0                              Page 82 of 2401

ATTACHMENT 1 EAL Bases (e.g., CT4, CT5, CT1 (Keowee).
      *A loss of essential power sources (e.g., CT4, CT5, CT1, 2, 3 (Keowee)) with a single train of essential buses being back-fed from an offsite power source.
Fifteen minutes was selected as a threshold to exclude transient or momentary losses of power.
The subsequent loss of the remaining single power source would escalate the event to an Alert in accordance with IC CA2.
ONS Bas~is Reference(s):I
: 1. UFSAR Figure 8.1 Single Line Diagram
: 2. UFSAR Section 8.20Offsite Power System
: 3. UFSAR Section 8.30Onsite Power Systems
: 4. NEI 99-01 CU2 S[Document No.]                               Rev. 0                         Page 83 of 240

ATTACHMENT 1 EAL Bases Category:                     C - Cold Shutdown / Refueling System Malfunction Subcategory:                  2 - Loss of Essential AC Power Initiating Condition:         Loss of all offsite and all emergency AC power to essential buses for 15 minutes or longer EAL:
CA2,1           Alert Loss of all offsiie and all emergency AC power capability to essential 4160 V buses MFB-1 and MFB-2 for > 15 min. (Note 1)
Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
Mode Applicability:
5 - Cold Shutdown, 6 - Refueling, NM          -  No Mode Basis:
For emergency classification purposes, "capability" means that an AC power source is available to and capable of powering the emergency bus(es) within 15 mai, whether or not the buses are currently powered from it.
The condition indicated by this EAL is the degradation of the offsite and onsite power sources resulting in a loss of all AC power to the emergency buses. 4160 V buses MFB-1 and MFB-2 are the essential buses (ref. 1).
Each unit is provided with two physically independent circuits from the switching station. One is the circuit from the 230 kV switching station through the startup transformer, which is designed to be available within a few seconds following a loss of coolant accident. The second circuit is the path from the switchyard through the main step-up transformer, the main generator bus and the unit auxiliary transformer with the generator disconnected from the main bus. The second circuit is currently used during refueling as an additional power feed for the shutdown unit(s) from the 230 kV switchyard. Whenever there is inadequate power from the generating units, the 230 kV switching station and the hydro units, power is available to the standby power buses either directly from the 100 kV Central Tie Substation or from Lee Steam Station via a Transformer CT5. (ref. 2)
Upon loss of power from the Oconee generating unit and 230 kV switchyard, power is supplied from both Keowee Hydro Station units through two separate and independent routes. One route is an underground feeder to Transformer CT4 which supplies the two redundant Main Feeder Buses (MFB-1 and MFB-2). The other route is an overhead feeder to the 230 KV switching station which supplies each unit startup transformer. (ref. 3)
The Standby Shutdown Facility (SSF) consists of standby systems for use in an extreme emergency and is equipped with a manually started diesel generator that can supply power necessary to maintain hot shutdown of the reactors of each unit in the event of loss of power.
However, the SSF cannot supply power to the essential buses and therefore not credited in this EAL (ref. 3).
I[Document No.]                                      Rev. 0         IPage                         84 of 240

ATTACHMENT 1 EAL Bases This cold condition EAL is equivalent to the hot condition loss of all offsite AC power EAL SS1.1.
This IC addresses a total loss of AC power that compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink.
When in the cold shutdown, refueling, or no mode, this condition is not classified as a Site Area Emergency because of the increased time available to restore an emergency bus to service. Additional time is available due to the reduced core decay heat load, and the lower temperatures and press*Jres in various plant systems. Thus, when in these mfodes, this condition represents an actual or potential substantial degradation of the level of safety of the plant.
Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.
Escalation of the emergency classification level would be via IC CS1 or RS1.
ONS Basis Reference(s):
: 1. UFSAR Figure 8.1 Single Line Diagram
: 2. UFSAR Section 8.20Offsite Power System
: 3. UFSAR Section 8.30Onsite Power Systems
: 4. NEI 99-01 CA2
[Document No.]                                Rev. 0                              Page 85 of 240

ATTACHMENT 1 EAL Bases Category:                      C - Cold Shutdown / Refueling System Malfunction Subcategory:                   3 - RCS Temperature Initiating Condition:          UNPLANNED increase in RCS temperature EAL:
Note 10: In the absence of reliable RCS temperature indication caused by the loss of decay heat removal capability, classification should be based on time to boil data.
Mode Applicability:
5 - Cold Shutdown, 6 - Refueling Definition(s):
UNPLANNED - A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.
Basis:
Several instruments are capable of providing indication of RCS temperature with respect to the Technical Specification cold shutdown temperature limit (200&deg;F, ref. 1). These include cold leg (Tc) temperature indications, hot leg (Th) temperature indications with RCPs running, CETCs and LPI cooler outlet temperature indications (ref. 2).
However, if Low Pressure Injection (LPI) flow is lost, the normal temperature elements used to monitor RCS temperature are not accurate indicators of RCS temperature. The CETCs are the design instruments for these conditions. For some periods of time the CETCs may not be available. The current practices concerning determining time to boil can be used in the evaluation of these EALs. Without CETC indication and with a loss of LPI flow the following guidance should be used (ref. 2):
* Use the predetermined "time to boil" data for evaluating these EALs. This approach reflects the relatively small numerical difference between the typical Technical Specification cold shutdown temperature limit of 2000 F and the boiling temperature of RCS water with the plant in Mode 5 or 6.
* Alternately, the Control Room staff may use a procedure or user aid to determine when RCS temperature will likely exceed 200&deg;F given the actual plant conditions (e.g., using a heat-up curve).
This IC addresses an UNPLANNED increase in RCS temperature above the Technical Specification cold shutdown temperature limit and represents a potential degradation of the level of safety of the plant. Ifthe RCS is not intact and CONTAINMENT CLOSURE is not established during this event, the Emergency Coordinator should also refer to IC CA3.
A momentary UNPLANNED excursion above the Technical Specification cold shutdown temperature limit when the heat removal function is available does not warrant a classification.
r[Document No.]                                        Rev.O                                Page 86 of 240

ATTACHMENT 1 EAL Bases This EAL involves a loss of decay heat removal capability, or an addition of heat to the RCS in excess of that which can currently be removed, such that reactor coolant temperature cannot be maintained below the cold shutdown temperature limit specified in Technical Specifications.
During this condition, there is no immediate threat of fuel damage because the core decay heat load has been reduced since the cessation of power operation.
During an outage, the level in the reactor vessel will normally be maintained at or above the reactor vessel flange. Refueling evolutions that lower water level below the reactor vessel flange are carefully planned and controlled. A loss of forced decay heat removal at reduced inventory may result in a rapid increase iij reactor coolant temperature depending on the timel after shutdown.
Escalation to Alert would be via IC CA1 based on an inventory loss or IC CA3 based on exceeding plant configuration-specific time criteria.
ONS Basis Reference(s):
: 1. ONS Technical Specifications Table 1.1-1
: 2. AP/l ,2,3/N1 700/026 Loss of Decay Heat Removal
: 3. NEI 99-01 CU3 I [Document No.]                                 Rev. 0                          Page   87 of 240 I
[DcmetNo]Re.0                                                                    ae,7of20       I

ATTACHMENT 1 EAL Bases Category:                 C - Cold Shutdown / Refueling System Malfunction Subcategory:              3 - RCS Temperature Initiating Condition:     UNPLANNED increase in RCS temperature EAL:
CU3.2          Unusual Event Loss of all RCS temperature and RCS level indication for ->15 min. (Note 1)
Note 1:  The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
Mode Applicability:
5 - Cold Shutdown, 6- Refueling Definition(s):
None Basis:
Several instruments are capable of providing indication of RCS temperature with respect to the Technical Specification cold shutdown temperature limit (200&deg;F, ref. 1). These include cold leg (Ta) temperature indications, hot leg (Th) temperature indications with RCPs running, CETCs and LPI cooler outlet temperature indications (ref. 2).
Several instruments are capable of providing indication of RCS level including pressurizer level, RVLIS, LT-5 and local monitor (ref. 3).
This EAL addresses the inability to determine RCS temperature and level, and represents a potential degradation of the level of safety of the plant. Ifthe RCS is not intact and CONTAINMENT CLOSURE is not established during this event, the Emergency Coordinator should also refer to IC CA3.
This EAL reflects. a condition where there has been a significant loss of instrumentation capability necessary to monitor RCS conditions and operators would be unable to monitor key parameters necessary to assure core decay heat removal. During this condition, there is no immediate threat of fuel damage because the core decay heat load has been reduced since the cessation of power operation.
Fifteen minutes was selected as a threshold to exclude transient or momentary losses of indication.
Escalation to Alert would be via IC CA1 based on an inventory loss or IC CA3 based on exceeding plant configuration-specific time criteria.
I[Document No.]                                      Rev. 0                              Page 88 of 240

ATTACHMENT 1 EAL Bases ONS Basis Reference(s):
: 1. ONS Technical Specifications Table 1.1-1
: 2. AP/1 ,2,3/A/1700/026 Loss of Decay Heat Removal
: 3. UFSAR Section 7.5.2.2 Inadequate Core Cooling Instruments
: 4. NEI 99-01 CU3 S[Document NO.]                             Rev. 0           Page 89 of 240

ATTACHMENT 1 EAL Bases Category:                      C  -  Cold Shutdown / Refueling System Malfunction Subcategory:                   3  -  RCS Temperature Initiating Condition:           Inability to maintain plant in cold shutdown EAL:
CA3.1 Alert UNPLANNED increase in RCS temperature to                  >  200&deg;F for > Table C-4 duration (Noteds 1, 10)
OR UNPLANNED RCS pressure increase > 10 psig due to a loss of RCS cooling (this EAL does* not apply during water-solid plant conditions)                     I Note 1:      The Emergency Coordinator should declare the event prom'ptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
Note 10:     In the absence of reliable RCS temperature indication caused by the loss of decay heat removal capability, classification should be based on time to boil data.
Table C-4: RCS Heat-up Duration Thresholds CONTAINMENT RCS Status                     CLOSURE Status                Heat-up Duration Intact (but not REDUCED                                          N/A                         60 main.*
INVENTORY)
Not intact                                   established                     20 min.*
OR REDUCED INVENTORY                          not established                    0 mai.
* If an RCS heat removal system is in operation within this time frame and RCS temperature is being reduced, the EAL is not applicable.
Mode Applicability:
5 - Cold Shutdown, 6 - Refueling Definition(s):
CONTAINMENT CLOSURE - The action to secure containment and its associated structures, systems, and components as a functional barrier to fission product release under all plant conditions up to and including a loss of decay heat removal or fuel handling accident inside containment.
As applied to ONS, Containment Closure is established when the requirements of OP/I1,2,3/A/1502/009, Containment Closure Control, are met.
UNPLANNED -. A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.
S[Document No.]                                         Rev. 0                                Page 90 of 240

ATTACHMENT 1 EAL Bases REDUCED INVENTORY - Condition with fuel in the reactor vessel and the level lower than three feet below the reactor vessel flange (RCS level < 50" on LT-5)
Basis:
Several instruments are capable of providing indication of RCS temperature with respect to the Technical Specification cold shutdown temperature limit (200&deg;F, ref. 1). These include cold leg (Ta) temperature indications, hot leg (Th) temperature indications with RCPs running, CETCs and LPI cooler outlet temperature indications (ref. 2).
However, if L~w Pressure Injection (LPI) flow is lost, the normal termperature elements used to monitor RCS temperature are not accurate indicators of RCS temperature. The CETCs are the design instruments for these conditions. For some periods of time the CETCs may not be available. The current practices concerning determining time to boil can be used in the evaluation of these EALs. Without CETC indication and with a loss of LPI flow the following guidance should be used (ref. 2):
* Use the predetermined "time to boil" data for evaluating these EALs. This approach reflects the relatively small numerical difference between the typical Technical Specification cold shutdown temperature limit of 200&deg;F and the boiling temperature of RCS water with the plant in Mode 5 or 6.
* Alternately, the Control Room staff may use a procedure or user aid to determine when ROS temperature will likely exceed 200&deg;F given the actual plant conditions (e.g., using a heat-up curve).
Numerous RCS pressure instruments are capable of measuring pressure to less than 10 psia including RCS low range cooldown pressure indicators RC-P-0086A/B (ref. 3).
This IC addresses conditions involving a loss of decay heat removal capability or an addition of heat to the RCS in excess of that which can currently be removed. Either condition represents an actual or potential substantial degradation of the level of safety of the plant.
A momentary UNPLANNED excursion above the Technical Specification cold shutdown temperature limit when the heat removal function is available does not warrant a classification.
The RCS Heat-up Duration Thresholds table addresses an increase in RCS temperature when CONTAINMENT CLOSURE is established but the RCS is not intact, or RCS inventory is reduced (e.g., mid-loop operation). The 20-minute criterion was included to allow time for operator action to address the temperature increase.
The RCS Heat-up Duration Thresholds table also addresses an increase in RCS temperature with the RCS intact. The status of CONTAINMENT CLOSURE is not crucial in this condition since the intact RCS is providing a high pressure barrier to a fission product release. The 60-minute time frame should allow sufficient time to address the temperature increase without a substantial degradation in plant safety.
Finally, in the case where there is an increase in RCS temperature, the RCS is not intact or is at reduced inventory, and CONTAINMENT CLOSURE is not established, no heat-up duration is allowed (i.e., 0 minutes). This is because 1) the evaporated reactor coolant may be released directly into the containment atmosphere and subsequently to the environment, and
: 2) there is reduced reactor coolant inventory above the top of irradiated fuel.
[Document No.]                                   Rev. 0                            Page 91 of 240

ATTACHMENT 1 EAL Bases The RCS pressure increase threshold provides a pressure-based indication of RCS heat-up in the absence of RCS temperature monitoring capability.
Escalation of the emergency classification level would be via IC OS1 or RSI.
ONS Basis Reference(s):
: 1. ONS Technical Specifications Table 1.1-1
: 2. AP/I1,2,3/A/1 700/026 Loss of Decay Heat Removal
: 3. IP/1 ,2,3/A/0200/047A Reactor Coolant System LTOP Instrument Calibration
: 4. NEI 99-01 CA3 S[Document No.]                              Rev. 0        1Page                92 of 240

ATTACHMENT 1 EAL Bases Category:                  C  -  Cold Shutdown / Refueling System Malfunction Subcategory:              4 - Loss of Vital DC Power Initiating Condition:      Loss of Vital DC power for 15 minutes or longer EAL:
CU4.1          Unusual Event Indicated voltage is < 10O5VDC on vital DC buses required by Technical Specifications for
  >15 min. (Note 1)I Note 1:   The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
Mode Applicability:
5 - Cold Shutdown, 6 - Refueling Definition(s):
None Basis:
The purpose of this EAL 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. The fifteen minute interval is intended to exclude transient or momentary power losses.
For each unit, two independent and physically separated 125 volt DC batteries and DC buses are provided for the vital instrumentation and control power system. (ref. 1, 2). Minimum DC bus voltage is 110 VDC (ref. 3).
This EAL is the cold condition equivalent of the hot condition loss of DC power EAL SS2.1.
This IC addresses a loss of vital DC power which compromises the ability to monitor and control operable SAFETY SYSTEMS when the plant is in the cold shutdown or refueling mode.
In these modes, the core decay heat load has been significantly reduced, and coolant system temperatures and pressures are lower; these conditions increase the time available to restore a vital DC bus to service. Thus, this condition is considered to be a potential degradation of the level of safety of the plant.
As used in this EAL, "required" means the vital DC buses necessary to support operation of the in-service, or operable, train or trains of SAFETY SYSTEM equipment. For example, if Train A is out-of-service (inoperable) for scheduled outage maintenance work and Train B is in-service (operable), then a loss of Vital DC power affecting Train B would require the declaration of an Unusual Event. A loss of Vital DC power to Train A would not warrant an emergency classification.
Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.
Depending upon the event, escalation of the emergency classification level would be via IC CA1 or CA3, or an IC in Recognition Category R.
I[Document No.]                                       Rev. 0                              Page 93 of 2401

ATTACHMENT 1 EAL Bases ONS Basis Reference(s):
: 1. UFSAR Figure 8.5 Typical DC and AC Vital Power System - Single Line
: 2. UFSAR Section 8.3.2 DC Power Systems
: 3. EP/*/AI1 800/001 Blackout Tab
: 4. Technical Specifications 3.8.4 DC Sources - Shutdown
: 5. NEI 99-01 CU4 S[Document No.]                    [      Rev.. 0       [Page           94 of 2401

ATTACHMENT 1 EAL Bases Category:                  C  -  Coid Shutdown / Refueling System Malfunction Subcategory:                5 -  Loss of Communications Initiating Condition:      Loss of all onsite or offsite communications capabilities EAL:
CU5.1          Unusual Event Loss of all Table C-5 onsite communication methods OR Loss of all Table C-5 offsite communication methods OR Loss of all Table C-5 NRC communication methods Table C-5    Communication Methods System                        Onsite    Offsite     NRC Commercial phone service                               X          X         X ONS site phone system                                  X          X         X EOF phone system                                                  X         X Public Address system                                   X Onsite radio system                                    X DEMNET                                                            X Offsite. radio systemX NRC Emergency Telephone System                                              X Satellite Phone                                                   X Mode Applicability:
5 - Cold Shutdown, 6 - Refueling, NM      -  No Mode Definition(s):
None S[Do~cument No.]                                  Rev.O0                            Page 95 of 240I

ATTACHMENT 1 EAL Bases Basis:
Onsite, offsite and NRC communications include one or more of the systems listed in Table 0-5 (ref. 1).
: 1. Commercial phone service The Commercial phone service does not go through the site telephone system.
: 2. ONS site phone system Trhe site phone system is generator and battery backed with:
* Fiber-Optic to Charlotte GO (65 lines)
* Telephone line to Easley (6 circuits)
I     Anderson (4 lines)
* Six Mile (4 lines)
* Site Telephone System: Inward and outward direct dial available from the Control Room, TSC, and OSC
: 3. EOF phone system The emergency communications systems at the Charlotte EOF are designed to ensure the reliable, timely flow of information between all parties having an emergency response role.
: 4. Public Address (Paging) system The paging system provides paging and party line communications between stations located throughout the plant. Inside and outside type wall and desk-mounted stations are used to communicate between roaming personnel and fixed work locations. Plant-wide instructions are issued using the paging feature.
: 5. Onsite radio system The onsite radio system receives emergency backup power from Keowee Hydro Units supporting communications with: Control Room 1&2, 3, Fire Brigade, Chemistry, Safety, Radiation Protection, Maintenance, Medical Emergency Response Team, and Hazardous Materials Response Team.
: 6. DEMNET DEMNET is the primary means of offsite communication. This circuit allows intercommunication among the EOF, TSC, control room, counties, and states. DEMNET operates as an internet based (VolP) communications system with a satellite back-up. Should the internet transfer rate become slow or unavailable, the DEMNET will automatically transfer to satellite mode.
: 7. Offsite radio system The offsite radio system is battery backed supporting communications with: Control Room Units 1&2, TSC, Field Monitoring Teams, EOF, counties and State of South Carolina.
I[Document No.]                                Rev. 0                           Page 96o&deg;240

ATTACHMENT 1 EAL Bases
: 8. NRC Emergency Telephone System (ETS)
The NRC uses a Duke Energy dedicated telephone line which allows direct telephone communications from the plant to NRC regional and national offices. The Duke Energy communications line provides a link independent of the local public telephone network.
Telephones connected to this network are located in the Oconee Control Rooms, Technical Support Center, and Emergency Operations Facility and can be used to establish NRC Emergency Notification System (ENS) and Health Physics Network (HPN) capability.
: 9. Satellitq Phone Satellite Phones can be used for external communications This EAL is the cold condition equivalent of the hot condition EAL SU7.1.
This IC addresses a significant loss of on-site or offsite communications capabilities. While not a direct challenge to plant or personnel safety, this event warrants prompt notifications to OROs and the NRC.
This IC should be assessed only when extraordinary means are being utilized to make communications possible (e.g., use of non-plant, privately owned equipment, relaying of on-site information via individuals or multiple radio transmission points, individuals being sent to offsite locations, etc.).
The first EAL condition addresses a total loss of the communications methods used in support of routine plant operations.
The second EAL condition addresses a total loss of the communications methods used to notify all OROs of an emergency declaration. The OROs referred to here are the State EOC and FEO, Pickens County LEC and EOC, and Oconee County LEC and EOC.
.The third EAL addresses a total loss of the communications methods used to notify the NRC of an emergency declaration.
ONS Basis Reference(s):
: 1. ONS Emergency Plan, Section 7.2 Communications Systems
: 2. NEI 99-01 CU5 S[Document No.]                                  Rev. 0                             Page 97 of 240

ATTACHMENT 1 EAL Bases Category:                C - Cold Shutdown / Refueling System Malfunction Subcategory:             6 - Hazardous Event Affecting Safety Systems Initiating Condition:    Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode EAL:
CA6.1          Alert The occurrence of ahiy Table C-6 hazardous event AND EITHER:
          *Event damage has caused indications of degraded performance in at least one train of a SAFETY SYSTEM needed for the current operating mfode
          *The event has caused VISIBLE DAMAGE to a SAFETY SYSTEM component or structure needed for the current operating mode Table 0-6    Hazardous Events
* Seismic event (earthquake)
* Internal or external FLOODING event
* High winds or tornado strike
* FIRE
* EXPLOSION
* Other events with similar hazard characteristics as determined by the Shift Manager Mode Applicability:
5 - Cold Shutdown, 6 - Refueling Definition(s):
EXPLOSION - A rapid, violent and catastrophic failure of a piece of equipment due to combustion, chemical reaction or overpressurization. A release of steam (from high energy lines or components) or an electrical component failure (caused by short circuits, grounding, arcing, etc.) should not automatically be considered an explosion. Such events require a post-event inspection to determine if the attributes of an explosion are present.
FIRE - Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred but is NOT required if large quantities of smoke and heat are observed.
FLOODING - A condition where water is entering a room or area faster than installed equipment is capable of removal, resulting in a rise of water level within the room or area.
((Document No.]                                 Rev. 0                            Page 98 of 240

ATTACHMENT 1 EAL Bases SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10OCFR50.2):
Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:
(1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which 6ould result in potential offsite exposures.
VISIBLE DAMAGE - Damage to a component or structure that is readily observable without measurements, testing, or analysis. The visual impact of the damage is sufficient ,to cause concern regarding the operability or reliability of the affected component or structure.
Basis:
* The significance of seismic events are discussed under EAL HU2.1 (ref. 1).
* Internal FLOODING may be caused by events such as component failures, equipment misalignment, or outage activity mishaps.
   *External flooding at ONS is highly unlikely since the probable maximum flood (PMF) would be contained by the Keowee Reservoir. Plant grade elevation is 796.0 ft MSL. The minimum external access elevation for the Auxiliary, Turbine, and Service Buildings is 796.5 ft MSL which provides a 6 inch water sill. (ref. 2)
   *High winds in excess of design (95 mph) or tornado strikes can cause significant structural damage (ref. 3).
   *Areas containing functions and systems required for safe shutdown of the plant are identified by fire area (ref. 4, 5).
   *An explosion that degrades the performance of a SAFETY SYSTEM train or visibly damages a SAFETY SYSTEM component or structure would be classified under this EAL.
This IC addresses a hazardous event that causes damage to a SAFETY SYSTEM, or a structure containing SAFETY SYSTEM components, needed for the current operating mode.
This condition significantly reduces the margin to a loss or potential loss of a fission product barrier, and therefore represents an actual or potential substantial degradation of the level of safety of the plant.
The first conditional addresses damage to a SAFETY SYSTEM train that is in service/operation since indications for it will be readily available. The indications of degraded performance should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train.
The second conditional addresses damage to a SAFETY SYSTEM component that is not in service/operation or readily apparent through indications alone, or to a structure containing SAFETY SYSTEM components. Operators will make this determination based on the totality of available event and damage report information. This is intended to be a brief assessment not requiring lengthy analysis or quantification of the damage.
I[Document NO.]                                  Rev. 0          ]Page99o24

ATTACHMENT 1 EAL Bases Escalation of the emergency classification level would be via IC CS1 or RSI.
ONS Basis Reference(s):
: 1. AP/0/A/1 700/005 Earthquake
: 2. UFSAR Section 3.4.1.1 Flood Protection Measures for Seismic Class I Structures
: 3. UFSAR Section 3.3.1.1 Design Wind Velocity
: 4. OSS-0254.00-00-4008 Design Bases Specification for Fire Protection
: 5. AP/1 ,2,3/A/1 700/050 Challenging Plant Fire
: 6. NEI 99-01 CA6             i S[Document No.]                              Rev. 0                          Page 100 of 240

ATTACHMENT 1 EAL Bases Category H - Hazards and Other Conditions Affecting Plant Safety EAL Group: ANY (EALs in this category are applicable to any plant condition, hot or cold.)
Hazards are non-plant, system-related events that can directly or indirectly affect plant operation, reactor plant safety or personnel safety.
: 1. Security Unauthorized entry attempts into tlie PROTECTED AREA, bomb threats, sabotage attempts, and actual security compromises threatening loss of physical control of the plant.
: 2. Seismic Event Natural events such as earthquakes have potential to cause plant structure or equipment damage of sufficient magnitude to threaten personnel or plant safety.
: 3. Natural or Technology Hazard Other natural and non-naturally occurring events that can cause damage to plant facilities include tornados, FLOODING, hazardous material releases and events restricting site access warranting classification.
: 4. Fire Fires can pose significant hazards to personnel and reactor safety. Appropriate for classification are fires within the Plant PROTECTED AREA or which may affect operability of equipment needed for safe shutdown
: 5. Hazardous Gas Toxic, corrosive, asphyxiant or flammable gas leaks can affect normal plant operations or preclude access to plant areas required to safely shutdown the plant.
: 6. Control Room Evacuation Events that are indicative of loss of Control Room habitability. If the Control Room must be evacuated, additional support for monitoring and controlling plant functions is necessary through the emergency response facilities.
: 7. Emergency Coordinator Judgment The EALs defined in other categories specify the predetermined symptoms or events that are indicative of emergency or potential emergency conditions and thus warrant classification. While these EALs have been developed to address the full spectrum of possible emergency conditions which may warrant classification and subsequent implementation of the Emergency Plan, a provision for classification of emergencies based on operator/management experience and judgment is still necessary. The EALs of this category provide the Emergency Coordinator the latitude to classify emergency conditions consistent with the established classification criteria based upon Emergency Coordinator judgment.
I[Document No.]                        I         Rev. 0        IPage                   101lof 240

ATTACHMENT 1 EAL Bases Category:                 H - Hazards Subcategory:              1 - Security Initiating Condition:     Confirmed SECURITY CONDITION or threat EAL:
HUI.1          Unusual Event A SECURITY CONDITION that does not involve a HOSTILE ACTION as reported by the Security Shift Supervision Mode Applicability:
All Definition(s):
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.
HOSTILE ACTION - An act toward ONS 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 ONS. 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).
Basis:
This EAL is based on the Duke Energy Physical Security Plan for ONS (ref. 1).
This IC addresses events that pose a threat to plant personnel or SAFETY SYSTEM equipment, and thus represent a potential degradation in the level of plant safety. Security events which do not meet one of these EALs are adequately addressed by the requirements of 10 CFR &sect; 73.71 or 10 CFR &sect; 50.72. Security events assessed as HOSTILE ACTIONS are classifiable under ICs HA1, HS1 and HGI.
Timely and accurate communications between Security Shift Supervision and the Control Room is essential for proper classification of a security-related event. Classification of these events will initiate appropriate threat-related notifications to plant personnel and Offsite Response Organizations.
Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan [and Independent Spent Fuel Storage InstallationSecurity Program].
This EAL references the Security Shift Supervision because these are the individuals trained to confirm that a security event is occurring or has occurred. Training on security event confirmation and classification is controlled due to the nature of Safeguards and 10 CER &sect; S[Document No.]                                 Rev. 0                            Page 102 of 240

ATTACHMENT 1 EAL Bases 2.39 information.
Emergency plans and implementing procedures are public documents; therefore, EALs should not incorporate Security-sensitive information. This includes information that may be advantageous to a potential adversary, such as the particulars concerning a specific threat or threat location. Security-sensitive information should be contained in non-public documents such as the Duke Energy Physical Security Plan for ONS (ref. 1).
Escalation of the emergency classification level would be via IC HAl.
ONlS Basis Reference(s):
: 1. Duke Energy Physical Security Plan for ONS
: 2. AP/0/A/1 700/045 Site Security Threats
: 3. NEI 99-01 HUI S[Document No.]                               Rev. 0                          Page 103 of 240

ATTACHMENT 1 EAL Bases Category:                 H - Hazards Subcategory:              1 - Security Initiating Condition:     Confirmed SECURITY CONDITION or threat EAL:
HUI.2           Unusual Event Notification of a credible security threat directed at the site Mode App~licability:
All Definition(s):
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.
Basis:
This EAL is based on the Duke Energy Physical Security Plan for ONS (ref. 1).
This IC addresses events that pose a threat to plant personnel or SAFETY SYSTEM equipment, and thus represent a potential degradation in the level of plant safety. Security events which do not meet one of these EALs are adequately addressed by the requirements of 10 CFR &sect; 73.71 or 10 CFR &sect; 50.72. Security events assessed as HOSTILE ACTIONS are classifiable under l~s HAl, HS1 and HGI.
Timely and accurate communications between Security Shift Supervision and the Control Room is essential for proper classification of a security-related event. Classification of these events will initiate appropriate threat-related notifications to plant personnel and Offsite Response Organizations.
Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan.[and Independent Spent Fuel Storage Installation Security Program].
This EAL addresses the receipt of a credible security threat. The credibility of the threat is assessed in accordance with the Duke Energy Physical Security Plan for ONS.
Emergency plans and implementing procedures are public documents; therefore, EALs should not incorporate Security-sensitive information. This includes information that may be advantageous to a potential adversary, such as the particulars concerning a specific threat or threat location. Security-sensitive information should be contained in non-public documents such as the Duke Energy Physical Security Plan for ONS (ref. 1).
Escalation of the emergency classification level would be via IC HA1.
S[Document No.]                                 Rev. 0                            Page 104 of 240

ATTACHMENT 1 EAL Bases ONS Basis Reference(s):
: 1. Duke Energy Physical Security Plan for ONS
: 2. AP/0/A/1 700/045 Site Security Threats
: 3. NEI 99-01 HU1 S[Document No.]                    I        Rev. 0  Pag~e 105 of 240

ATTACHMENT 1 EAL Bases Category:                 H - Hazards Subcategory:              1 - Security Initiating Condition:     Confirmed SECURITY CONDITION or threat EAL:
HUI.3           Unusual Event A validated notification from the NRC providing information of an aircraft threat Mode Applicability:
All Definition(s):
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.
Basis:
This EAL is based on the Duke Energy Physical Security Plan for ONS (ref. 1).
This IC addresses events that pose a threat to plant personnel or SAFETY SYSTEM equipment, and thus represent a potential degradation in the level of plant safety. Security events which do not meet one of these EALs are adequately addressed by the requirements of 10 CFR &sect; 73.71 or 10 CFR &sect; 50.72. Security events assessed as HOSTILE ACTIONS are classifiable under ICs HAl, HS1 and HG1.
Timely and accurate communications between Security Shift Supervision and the Control Room is essential for proper classification of a security-related event. Classification of these events will initiate appropriate threat-related notifications to plant personnel and Offsite Response Organizations.
Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification.Plan,Safeguards Contingency Plan [and Independent Spent Fuel Storage InstallationSecurity Program]..
This EAL addresses the threat from the impact of an aircraft on the plant. The NRC Headquarters Operations Officer (HOO) will communicate to the licensee if the threat involves an aircraft. The status and size of the plane may also be provided by NORAD through the NRC. Validation of the threat is performed in accordance with AP/OIAI17001045 Site Security Threats (ref. 2).
Emergency plans and implementing procedures are public documents; therefore, EALs should not incorporate Security-sensitive information. This includes information that may be advantageous to a potential adversary, such as the particulars concerning a specific threat or threat location. Security-sensitive information should be contained in non-public documents such as the Duke Energy Physical Security Plan for ONS (ref. 1).
Escalation of the emergency classification level would be via IC HAl.
I[Document No.]                         .        Rev. 0                            Page 106 of.240

ATTACHMENT 1 EAL Bases ONS Basis Reference(s):
: 1. Duke Energy Physical Security Plan for ONS
: 2. AP/0/A/1700/045 Site Security Threats
: 3. NEI 99-01 HU1 S[Document No.]                             Rev. 0 Page 107 of 240I

ATTACHMENT 1 EAL Bases Category:                 H - Hazards Subcategory:              1 - Security Initiating Condition:     HOSTILE ACTION within the OWNER CONTROLLED AREA or airborne attack threat within 30 minutes EAL:
HAl.1          Alert A HOSTILE ACTION is occurring or has occdurred within the OWNER CONTROLLED AREA as reported by the Security Shift Supervision Mode Applicability:
All Definition(s):
HOSTILE ACTION - An act toward ONS 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 ONS. 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).
OWNER CONTROLLED AREA - Area outside the PROTECTED AREA fence that immediately surrounds the plant. Access to this area is generally restricted to those entering on official business.
Basis:
This IC addresses the occurrence of a HOSTILE ACTION within the OWNER CONTROLLED AREA or notification of an aircraft attack threat. This event will require rapid response and assistance due to the possibility of the attack progressing to the PROTECTED AREA, or the need to prepare the plant and staff for a potential aircraft impact.
Timely and accurate communications between Security Shift Supervision and the Control Room is essential for proper classification of a security-related event.
Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan [and In dependent Spent Fuel Storage Installation Security Program].
As time and conditions allow, these events require a heightened state of readiness by the plant staff and implementation of onsite protective measures (e.g., evacuation, dispersal or sheltering). The Alert declaration will also heighten the awareness of Offsite Response Organizations (OROs), allowing them to be better prepared should it be necessary to consider further actions.
This IC does not apply to incidents that are accidental events, acts of civil disobedience, or otherwise are not a HOSTILE ACTION perpetrated by a HOSTILE FORCE. Examples include I[Document No.]                        I        Rev.O0                            Page 108 of 2401

ATTACHMENT 1 EAL Bases the crash of a small aircraft, shots from hunters, physical disputes between employees, etc.
Reporting of these types of events is adequately addressed by other EALs, or the requirements of 10 CFR &sect; 73.71 or 10 CFR &sect; 50.72.
This EAL is applicable for any HOSTILE ACTION occurring, or that has occurred, in the OWNER CONTROLLED AREA. This includes any action directed against an ISFSI that is located outside the plant PROTECTED AREA.
Emergency plans and implementing procedures are public documents; therefore, EALs should not incorporate Security-sensitive information. This includes information that may be tadvantageous to a potential adversary, such as the Ij*articulars concerning a specific threat or threat location. Security-sensitive information should be contained in non-public documents such as the Duke Energy Physical Security Plan for ONS (ref. 1).
SEscalation of the emergency classification level would be via IC HSI.
ONS Basis Reference(s):
: 1. Duke Energy Physical Security Plan for ONS
: 2. AP/O/NI1700/045 Site Security Threats
: 3. NEI 99-01 HAl I[Document No.]                                Rev. 0          IPage                  109 of 240I

ATTACHMENT 1 EAL Bases Category:                 H - Hazards Subcategory:              1 - Security Initiating Condition:     HOSTILE ACTION within the OWNER CONTROLLED AREA or airborne attack threat within 30 minutes EAL:
HA1.2          Alert iAvalidated notification from NRC of an aircraft attaclk threat within 30 min. of the site Mode Applicability:
*All Definition(s):
HOSTILE ACTION - An act toward ONS 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 ONS. 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).
Basis:
This IC addresses the occurrence of a HOSTILE ACTION within the OWNER CONTROLLED AREA or notification of an aircraft attack threat. This event will require rapid response and assistance due to the possibility of the attack progressing to the PROTECTED AREA, or the need to prepare the plant and staff for a potential aircraft impact.
Timely and accurate communications between Security Shift Supervision and the Control Room is essential for proper classification of a security-related event.
Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and QualificationPlan, Safeguards Contingency Plan [and Independent Spent Fuel Storage Installation Security Program].
As time and conditions allow, these events require a heightened state of readiness by the plant staff and implementation of onsite protective measures (e.g., evacuation, dispersal or sheltering). The Alert declaration will also heighten the awareness of Offsite Response Organizations (OROs), allowing them to be better prepared should it be necessary to consider further actions.
This IC does not apply to incidents that are accidental events, acts of civil disobedience, or otherwise are not a HOSTILE ACTION perpetrated by a HOSTILE FORCE. Examples include the crash of a small aircraft, shots from hunters, physical disputes between employees, etc.
Reporting of these types of events is adequately addressed by other EALs, or the requirements of 10 CFR &sect; 73.71 or 10 CFR &sect; 50.72.
I[Document No.1                                 Rev. 0                            Page 110 of 240I

ATTACHMENT 1 EAL Bases This EAL addresses the threat from the impact of an aircraft on the plant, and the anticipated arrival time is within 30 minutes. The intent of this EAL is to ensure that threat-related notifications are made in a timely manner so that plant personnel and OROs are in a heightened state of readiness. This EAL is met when the threat-related information has been validated in accordance with AP/0/AII1700/045 Site Security Threats (ref. 2).
The NRC Headquarters Operations Officer (HOO) will communicate to the licensee if the threat involves an aircraft. The status and size of the plane may be provided by NORAD through the NRC.
In some Icases, it may not be readily apparent if an aircraft imlbact within the OWNER CONTROLLED AREA was intentional (i.e., a HOSTILE ACTION). It is expected, although not certain, that notification by an appropriate Federal agency to the site would clarify this point. In this case, the appropriate federal agency is intended to be NORAD, FBI, FAA or NRC. The emergency declaration, including one based on other ICs/EALs, should not be unduly delayed while awaiting notification by) a Federal agency.
Emergency plans and implementing procedures are public documents; therefore, EALs should not incorporate Security-sensitive information. This includes information that may be advantageous to a potential adversary, such as the particulars concerning a specific threat or threat location. Security-sensitive information should be contained in non-public documents such as the Duke Energy Physical Security Plan for ONS (ref. I).
Escalation of the emergency classification level would be via IC HSl.
ONS Basis Reference(s):
: 1. Duke Energy Physical Security Plan for ONS
: 2. AP/0/A/1700/045 Site Security Threats
: 3. NEI 99-01 HA1 S[Document No.] .Rev.                               0                            Page 111 of 240

ATTACHMENT 1 EAL Bases Category:                 H - Hazards Subcategory:              1 - Security Initiating Condition:     HOSTILE ACTION within the PROTECTED AREA EAL:
HSI.1          Site Area Emergency A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by the Security Shift Supervision Mode Applicability:
All Definition(s):
HOSTILE ACTION - An act toward ONS 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 ONS. 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).
PROTECTED AREA - That part of the nuclear station property consisting of the Reactor, Auxiliary, Turbine, and Service Building and grounds, contained within the owner controlled security fence.
Basis:
The Security Shift Supervision are the designated on-site personnel 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 Duke Energy Physical Security Contingency Plan for ONS (Safeguards) information. (ref. 1)
This IC addresses the occurrence of a HOSTILE ACTION within the PROTECTED AREA.
This event will require rapid response and assistance due to the possibility for damage to plant equipment.
Timely and accurate communications between Security Shift Supervision and the Control Room is essential for proper classification of a security-related event (ref. 1, 2).
Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan [and Independent Spent Fuel Storage Installation Security Program].
As time and conditions allow, these events require a heightened state of readiness by the plant staff and implementation of onsite protective measures (e.g., evacuation, dispersaror sheltering). The Site Area Emergency declaration will mobilize Offsite Response Organization (ORO) resources and have them available to develop and implement public protective actions in the unlikely event that the attack is successful in impairing multiple safety functions.
I[Document No.]                                  Rev. 0                            Page 112 of 240I

ATTACHMENT 1 EAL Bases This IC does not apply to a HOSTILE ACTION directed at an ISFSI PROTECTED AREA located outside the PROTECTED AREA; such an attack should be assessed using IC HAl. It also does not apply to incidents that are accidental events, acts of civil disobedience, or otherwise are not a HOSTILE ACTION perpetrated by a HOSTILE FORCE. Examples include the crash of a small aircraft, shots from hunters, physical disputes between employees, etc.
Reporting of these types of events is adequately addressed by other EALs, or the requirements of 10 CFR &sect; 73.71 or 10 CFR &sect; 50.72.
Emergency plans and implementing procedures are public documents; therefore, EALs should not incorporate Security-sensitive information. This includes information that npay be advantageous to a potential adversary, such as the particulars concerning a specific threat or threat location. Security-sensitive information should be contained in non-public documents such as the Duke Energy Physical Security Plan for ONS (ref. 1).
Escalation of the emergency classification level would be via IC HG1.
ONS Basis Reference(s):
: 1. Duke Energy Physical Security Plan for ONS
: 2. AP/0/A/1700/045 Site Security Threats
: 3. NEI 99-01 HS1
[Document No.]                                Rev. 0                            Page 113 of 240

ATTACHMENT 1 EAL Bases Category:                   H - Hazards Subcategory:                 1 - Security Initiating Condition:       HOSTILE ACTION resulting in loss of physical control of the facility EAL:
HGI.1           General Emergency A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by the Security Shift Supprvision AND EITHER of the following has occurred:
Any of the following safety functions cannot be controlled or maintained
* Reactivity
* Core cooling
* RCS heat removal OR Damage to spent fuel has occurred or is IMMINENT Mode Applicability:
All Definition(s):
HOSTILE ACTION - An act toward ONS 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 ONS. 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).
IMMINENT - The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions PROTECTED AREA - That part of the nuclear station property consisting of the Reactor, Auxiliary, Turbine, and Service Building and grounds, contained within the owner controlled security fence.
Basis:
Indications of damaged spent fuel are provided in AP/1 ,2,3/A/1 700/009 Spent Fuel Damage (ref. 4).
This IC addresses an event       in which a HOSTILE FoRCE has taken physical control of the facility to the extent that the  plant staff can no longer operate equipment necessary to maintain key safety functions. It also    addresses a HOSTILE ACTION leading to a loss of physical control that results in actual   or IMMINENT damage to spent fuel due to 1) damage to a spent I[Document No.]                           I        Rev. 0                         Page 114 of 240I

ATTACHMENT 1 EAL Bases fuel pool cooling system (e.g., pumps, heat exchangers, controls, etc.) or, 2) loss of spent fuel pool integrity such that sufficient water level cannot be maintained.
Timely and accurate communications between the Security Shift Supervision and the Control Room is essential for proper classification of a security-related event (ref. 2, 3).
Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Trainingand Qualification Plan, Safeguards Contingency Plan [and Independent Spent Fuel Storage Installation Security Program].
Emergency plans and implementing procediures are public documents; therefore, EALs should not incorporate Security-sensitive information. This includes information that may be advantageous to a potential adversary, such as the particulars concerning a specific threat or threat location. Security-sensitive information should be contained in non-public documents such as the Duke Energy Physical Security Plan for ONS (ref.1).
ONS Basis Reference(s):
: 1. Duke Energy Physical Security Plan for ONS
: 2. AP/0/A/1 700/045 Site Security Threats
: 3. APIOIAI17001046 Extensive Damage Mitigation
: 4. AP/1 ,2,3/A/1700/009 Spent Fuel Damage
: 5. NEI 99-01 HG1
* [Document No.]                                  Rev. 0                           Page115 of 240

ATTACHMENT 1 EAL Bases Category:                H - Hazards and Other Conditions Affecting Plant Safety Subcategory:              2 - Seismic Event Initiating Condition:    Seismic event greater than OBE levels EAL:
HU2.1          Unusual Event Seismic event > OBE as indicated by EITHER of the following:
* ISA-9/E-1 (SEISMIC TRIGGER) alarm
* 35A-9/E-1 (SEISMIC TRIGGER) alarm Mode Applicability:
All Definition(s):
None Basis:
This EAL is based on a VALID receipt of either of the specified seismic trigger alarms.
For ONS, the Operating Basis Earthquake (OBE) is equivalent to the Design Basis Earthquake (DBE). The design basis earthquake ground acceleration at the site is 0.05g. The maximum hypothetical earthquake ground acceleration is 0.10g and 0.15g for Class 1 structures founded on bedrock and overburden respectively. For ONS, the Operating Basis Earthquake (OBE) is equivalent to the Design Basis Earthquake (DBE). (ref. 1)
If an earthquake of > 0.05 g has occurred on site, all units are required to be shut down to Mode 5 once a plant damage assessment is complete along with the completion of any needed repairs to support the units ability to achieve safe shutdown. (ref. 2)
Earthquake instrumentation is the SMA-3 system consisting of a central recording system, control panel, one TS-3 triaxial seismic trigger package, and two force-balance triaxial accelerometer packages. The seismic trigger and one accelerometer of the SMA-3 system are located in the Unit 1 Tendon Gallery. Also, a second accelerometer is located directly above at elevation 797' +6" in the Oconee 1 Reactor Building. The recorder for the system is located in the Unit 1 Cable Room. Also, a seismic trigger/switch is located in the Unit 1 tendon gallery.
The TS-3 has a preset acceleration threshold of 0.05g which activates the statalarm in Units 1 and 3 control rooms, when design conditions occur. (ref. 3)
To avoid inappropriate emergency classification resulting from spurious actuation of the seismic instrumentation or felt motion not attributable to seismic activity, an offsite agency (USGS, National Earthquake Information Center (NEIC)) can confirm that an earthquake has occurred in the area of the plant. Such confirmation should not, however, preclude a timely emergency declaration based on receipt of the OBE alarm. The NEIC can be contacted by calling (303) 273-8500 (ref. 2). Select option #1 and inform the analyst you wish to confirm recent seismic activity in the vicinity of ONS. If requested, provide the analyst with the S[Document No.]                                 Rev. 0         .Page                    116 of 240,

ATTACHMENT 1 EAL Bases following ONS coordinates:     340 47' 38.2" north latitude, 820 53' 55.4" west longitude (ref. 4).
Alternatively, near real-time seismic activity can be accessed via the NEIC website:
http://earthqaeusgs.gov/eqcenterI This IC addresses a seismic event that results in accelerations at the plant site greater than those specified for an Operating Basis Earthquake (OBE). An earthquake greater than an OBE but less than a Safe Shutdown Earthquake (SSE) should have no significant impact on safety-related systems, structures and components; however, some time may be required for the plant staff to ascertain the actual post-event condition of the plant (e.g., performs walk-downsI and post-event inspections). Given the time necessary to perform walk-downs and inspections, and fully understand any impacts, this event represents a potential degradation of the level of safety of the plant.
Event qverification with external sources should not be necessary during or following an OBE.
Earthquakes of this magnitude should be readily felt by on-site personnel and recognized as a seismic event (e.g., lateral accelerations in excess of 0.05g). The Shift Manager or Emergency Coordinator may seek external verification if deemed appropriate (e.g., a call to the USGS, check internet news sources, etc.); however, the verification action must not preclude a timely emergency declaration.
Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via IC CA6 or SA9.
ONS Basis Reference(s):
: 1. UFSAR Section 3.2.1.3 Seismic Loading Conditions
: 2. APIO/AI17001005 Earthquake
: 3. UFSAR Section 3.7.4 Seismic Instrumentation Program
: 4. UFSAR Section 2.1.1.1 Specification of Location
: 5. NEI 99-01 HU2 l[Document No.]                                 Rev. 0                             Page117 of 240

ATTACHMENT 1 EAL Bases Category:                  H - Hazards and Other Conditions Affecting Plant Safety Subcategory:               3 - Natural or Technological Hazard Initiating Condition:      Hazardous event EAL:
HU3.1          Unusual Event A tornado strike within the PROTECTED AREA Mode Applicability:
All Definition(s):   I PROTECTEDAREA - That part of the nuclear station property consisting of the Reactor, Auxiliary, Turbine, and Service Building and grounds, contained within the owner controlled security fence.
Basis:
Response actions associated with a tornado onsite is provided in AP/0/A/1700/006, Natural Disaster (ref. 1).
If damage is confirmed visually or by other in-plant indications, the event may be escalated to an Alert under EAL CA6.1 or SA9.1.
A tornado striking (touching down) within the PROTECTED AREA warrants declaration of an Unusual Event regardless of the measured wind speed at the meteorological tower. A tornado is defined as a violently rotating column of air in contact with the ground and extending from the base ofa thunderstorm.
This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant.
EAL HU3.1 addresses a tornado striking (touching down) within the PROTECTED AREA.
Escalation of the emergency classification level would be based on ICs in Recognition Categories R, F, S or C.
ONS Basis Reference(s):
: 1. AP/O/A/1 700/006 Natural Disaster
: 2. NEI 99-01 HU3 I[Document No.]                                 Rev. 0                           page 118 of 240

ATTACHMENT 1 EAL Bases Category:                  H - Hazards and Other Conditions Affecting Plant Safety Subcategory:               3 - Natural or Technological Hazard Initiating Condition:       Hazardous event EAL:
HU3.2           Unusual Event Internal room or area FLOODING of a magnitude sufficient to require manual or automatic electrical isolation of a SAFETY SYSTEM component needed for the current1 operating mode Mode Applicability:
All Definition(s):
FLOODING - A condition where water is entering a room or area faster than installed equipment is capable of removal, resulting in a rise of water level within the room or area.
SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2):
Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:
(1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.
Basis:
Areas susceptible to internal flooding are the Turbine Building and Auxiliary Building (ref.1, 2).
Refer to EAL CA6.1 or SA9.1 for internal flooding affecting one or more SAFETY SYSTEM trains.
This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant.
This EAL addresses FLOODING of a building room or area that results in operators isolating power to a SAFETY SYSTEM component due to water level or other wetting concerns.
Classification is also required if the water level or related wetting causes an automatic isolation of a SAFETY SYSTEM component from its power source (e.g., a breaker or relay trip). To warrant classification, operability of the affected component must be required by Technical Specifications for the current operating mode.
Escalation of the emergency classification level would be based on ICs in Recognition Categories R, F, SorC.
ONS Basis Reference(s):
I[Document No.]                                   Rev.O0                            Page 119 of 240

ATTACHMENT 1 EAL Bases
: 1. AP/1,2,3/A/1700/010 Turbine Building Flood
: 2. AP/1-2,3/A/1700/030 Auxiliary Building Flood
: 3. NEI 99-01 HU3 S[Document No.]                            Rev. 0  Page 120 of 240

ATTACHMENT 1 EAL Bases Category: *H                 - Hazards and Other Conditions Affecting Plant Safety Subcategory:               3-  Natural or Technological Hazard Initiating Condition:      Hazardous event EAL:
HU3.3           Unusual Event Movement of personnel within the PROTECTED AREA is IMPEDED due to an offsite event involving hazardous mater~ials (e.g., an offsite chemical spill or toxic gas releas$)
Mode Applicability:
All Definition(s):
IMPEDE(D) - Personnel access to a room or area is hindered to an extent that extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., requiring use of protective equipment, such as SCBAs, that is not routinely employed).
PROTECTED AREA - That part of the nuclear station property consisting of the Reactor, Auxiliary, Turbine, and Service Building and grounds, contained within the owner controlled security fence.
Basis:
As used here, the term "offsite" is meant to be areas external to the ONS PROTECTED AREA.
This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant.
This EAL addresses a hazardous materials event originating at an offsite location and of sufficient magnitude to impede the movement of personnel within the PROTECTED AREA.
Escalation of the emergency classification level would be based on ICs in Recognition Categories R, F, S or C.
ONS Basis Reference(s):
: 1. NEI 99-01 HU3 S[Document No.]                                Rev. 0                            Page 121   of 240I

ATTACHMENT 1 EAL Bases Category:                  H - Hazards and Other Conditions Affecting Plant Safety Subcategory:               3 - Natural or Technological Hazard Initiating Condition:      Hazardous event EAL:
Note 7:  This EAL does not apply to routine traffic impediments such as fog, snow, ice, or vehicle breakdowns or accidents.
Mode Applicability:I All Definition(s):
None Basis:
This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant.
This EAL addresses a hazardous event that causes an on-site impediment to vehicle movement and significant enough to prohibit the plant staff from accessing the site using personal vehicles. Examples of such an event include site FLOODING caused by a hurricane, heavy rains, up-river water releases, dam failure, etc., or an on-site train derailment blocking the access road.
This EAL is not intended apply to routine impediments such as fog, snow, ice, or vehicle breakdowns or accidents, but rather to more significant conditions such as the Hurricane Andrew strike on Turkey Point in 1992, the flooding around the Cooper Station during the Midwest floods of 1993, or the flooding around Ft. Calhoun Station in 2011.
Escalation of the emergency classification level would be based on ICs in Recognition Categories R, F, S or C.
ONS Basis Reference(s):
: 1. NEI 99-01 HU3 I[Document No.]                                       Rev. 0                                Page 122 of 240I

ATTACHMENT I EAL Bases Category:                   H - Hazards and Other Conditions Affecting Plant Safety Subcategory:                4- Fire Initiating Condition:       FIRE potentially degrading the level of safety of the plant EAL:
HU4.I           Unusual Event A FIRE is not extinguished within 15 min. of any of the following FIRE detection indications (Note 1):
* Report from the field (i.e., visual observation)
* Receipt of multiple (more than 1) fire alarms or indications
* Field verification of a single fire alarm n AND The FIRE is located within any Table H-I area Note 1:   The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
Table H-I Fire Areas
* Reactor Building
* Auxiliary Building
* Turbine Building
* Standby Shutdown Facility
* Intake Structure
* Electrical Blockhouse
* Keowee Hydro & associated transformers
* Transformer Yard
* Protected Service Water Building
* Essential Siphon Vacuum Building Mode Applicability:
All Definition(s):
FIRE - Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred but is NOT required if large quantities of smoke and heat are observed.
Basis:
The 15 minute requirement begins with a credible notification that a fire is occurring, or receipt of multiple valid fire detection system alarms or field validation of a single fire alarm. The alarm is to be validated using available Control Room indications or alarms to prove that it is not spurious, or by reports from the field.
I[Document No.]                                       Rev. 0                            Page 123 of 240I

ATTACHMENT I EAL Bases Table H-I Fire Areas are those areas that contain equipment necessary for safe operation and shutdown of the plant (ref. 1, 2).
This IC addresses the magnitude and extent of FIRES that may be indicative of a potential degradation of the level of safety of the plant.
For EAL HU4.1 the intent of the 15-minute duration is to size the FIRE and to discriminate against small FIRES that are readily extinguished (e.g., smoldering waste paper basket). In addition to alarms, other indications of a FIRE could be a drop in fire main pressure, automatic activation of a suppression system, etc.
Upon receipt, operators will take prompt actions to confirm the validity of an initial fire alarm, indication, or report. For EAL assessment purposes, the emergency declaration clock starts at the time that the initial alarm, indication, or report was received, and not the time that a subsequent verification action was performed. Similarly, the fire duration clock also starts at the time of receipt of the initial alarm, indication or report.
Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via IC CA6 or SA9.
ONS Basis Reference(s):
: 1. OSS-0254.00-00-4008 Design Bases Specification for Fire Protection
: 2. AP/1,2,3/A/1700/050 Challenging Plant Fire
: 3. NEI 99-01 HU4 S[Document No.]                                   Rev. 0        [Page                   124 of 240I

ATTACHMENT 1 EAL Bases Category:                  H - Hazards and Other Conditions Affecting Plant Safety Subcategory:               4 - Fire Initiating Condition:      FIRE potentially degrading the level of safety of the plant EAL:
HU4.2          Unusual Event Receipt of a single fire alarm (i.e., no other indications of a FIRE)
AND        I The fire alarm is indicating a FIRE within any Table H-i area AND The existence of a FIRE is not verified within 30 mm. of alarm receipt (Note 1)
Note 1:    The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
Table H-I Fire Areas
* Reactor Building
* Auxiliary Building
* Turbine Building
* Standby Shutdown Facility
* Intake Structure
* Electrical Blockhouse
* Keowee Hydro & associated transformers
* Transformer Yard
* Protected Service Water Building
* Essential Siphon Vacuum Building Mode Applicability:
All Definition(s):
FIRE - Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred but is NOT required if large quantities of smoke and heat are observed.
Basis:
The 30 minute requirement begins upon receipt of a single valid fire detection system alarm.
The alarm is to be validated using available Control Room indications or alarms to prove that it is not spurious, or by reports from the field. Actual field reports must be made within the 30 minute time limit or a classification must be made. If a fire is verified to be occurring by field report, classification shall be made based on EAL HU4.1.
S[Document No.]                                       Re.0Page                                  125 of 240I

ATTACHMENT 1 EAL Bases Control Room indications that may be used to validate a single fire alarm include (ref. 3):
  -     Remote camera system
  -     CRD service structure air temperature
  -     PZR tailpipe temperature
  -     RB dome temperature
    -RBCU        inlet and outlet temperatures
    -RCP       parameters
    -Status      lights of components located inside RB Table F Fire Areas *re those areas that contain equipment necessary fo&safe operation and shutdown of the plant (ref. 1, 2).
The ONS Fire Protection Program is based on 10 CFR 50.48 (a) and (c) requiring compliance with NFPA 805. The NFPA 805 based Fire Protection Program requirements provide are consistent with the NEI 99-01 basis stated below (ref. 1, 4).
This IC addresses the magnitude and extent of FIRES that may be indicative of a potential degradation of the level of safety of the plant.
This EAL addresses receipt of a single fire alarm, and the existence of a FIRE is not verified (i.e., proved or disproved) within 30-minutes of the alarm. Upon receipt, operators will take prompt actions to confirm the validity of a single fire alarm. For EAL assessment purposes, the 30-minute clock starts at the time that the initial alarm was received, and not the time that a subsequent verification action was performed.
A single fire alarm, absent other indication(s) of a FIRE, may be indicative of equipment failure or a spurious activation, and not an actual FIRE. For this reason, additional time is allowed to verify the validity of the alarm. The 30-minute period is a reasonable amount of time to determine if an actual FIRE exists; however, after that time, and absent information to the contrary, it is assumed that an actual FIRE is in progress.
If an actual FIRE is verified by a report from the field, then HU4.1 is immediately applicable, and the emergency must be declared if the FIRE is not extinguished within 15-minutes of the report. If the alarm is verified to be due to an equipment failure or a spurious activation, and this verification occurs within 30-minutes of the receipt of the alarm, then this EAL is not applicable and no emergency declaration is warranted.
Basis-Related Requirements from Appendix R Appendix R to 10 CFR 50, states in part:
Criterion 3 of Appendix A to this part specifies that "Structures, systems, and components important to safety shall be designed and located to minimize, consistent with other safety requirements, the probability and effect of fires and explosions."
When considering the effects of fire, those systems associated with achieving and maintaining safe shutdown conditions assume major importance to safety because damage to them can lead to core damage resulting from loss of coolant through boil-off.
Because fire may affect safe shutdown systems and because the loss of function of systems used to mitigate the consequences of design basis accidents under post-fire conditions does not per se impact public safety, the need to limit fire damage to systems required to achieve and maintain safe shutdown conditions is greater than the need to I[Document No.]                                    Rev. 0                            Page 1.26 of 240I

ATTACHMENT 1 EAL Bases limit fire damage to those systems required to mitigate the consequences of design basis accidents.
In addition, Appendix R to 10 CFR 50, requires, among other considerations, the use of 1-hour fire barriers for the enclosure of cable and equipment and associated non-safety circuits of one redundant train (G.2.c). As used in HU4.2, the 30-minutes to verify a single alarm is well within this worst-case 1-hour time period.
Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via IC CA6 or SA9.
ONS Basis Reference(s):
: 1. OSS-0254.00-00-4008 Design Bases Specification for Fire Protection
: 2. AP/1 ,2,3/A/1700/050 Challenging Plant Fire
: 3. OP/i1,2,31AI61011003          I
: 4. NRC Letter to T. Preston Gillespie (Duke); ONS Units 1, 2, and 3, Issuance of Amendments Regarding Transition to a Risk-Informed, Performance-Based Fire Protection Program in Accordance With 10 CFR 50.48(c); dated December 29, 2010
: 5. NEI 99-01 HU4 S[Document No.]                               Rev. 0                          Page 127 of 240

ATTACHMENT 1 EAL Bases Category:                  H - Hazards and Other Conditions Affecting Plant Safety Subcategory:               4- Fire Initiating Condition:      FIRE potentially degrading the level of safety of the plant EAL:
HU4.3          Unusual Event A FIRE within the PROTECTED AREA not extinguished within 60 min. of the initial report, alarm or indication (Note 1)
Note 1:   The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
Mode Applicability:
All Definition(s):
FIRE - Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred but is NOT required if large quantities of smoke and heat are observed.
PROTECTEDAREA - That part of the nuclear station property consisting of the Reactor, Auxiliary, Turbine, and Service Building and grounds, contained within the owner controlled security fen ce.
Basis:
This IC addresses the magnitude and extent of FIRES that may be indicative of a potential degradation of the level of safety of the plant.
In addition to a FIRE addressed by EAL HU4.1 or HU4.2, a FIRE within the plant PROTECTED AREA not extinguished within 60-minutes may also potentially degrade the level of plant safety.
Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via IC CA6 or SA9.
ONS Basis Reference(s):
: 1. NEI 99-01 HU4 r[Document No.]                                       Rev. 0                            Page 128 of 240

ATTACHMENT 1 EAL Bases Category:                 H - Hazards and Other Conditions Affecting Plant Safety Subcategory:              4 - Fire Initiating Condition:      FIRE potentially degrading the level of safety of the plant EAL:
HU4.4           Unusual Event A FIRE within the PROTECTED AREA that requires firefighting support by an offsite fire response agency to extinguish Mode Applicability:
All Definition(s):
FIRE - Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred but is NOT required if large quantities of smoke and heat are observed.
PROTECTED AREA - That part of the nuclear station property consisting of the Reactor, Auxiliary, Turbine, and Service Building and grounds, contained within the owner controlled security fence.
Basis:
This IC addresses the magnitude and extent of FIRES that may be indicative of a potential degradation of the level of safety of the plant.
If a FIRE within the PLANT PROTECTED AREA is of s~ufficient size to require a response by an offsite firefighting agency (e.g., a local town Fire Department), then the level of plant safety is potentially degraded. The dispatch of an offsite firefighting agency to the site requires an emergency declaration only if it is needed to actively support firefighting efforts because the fire is beyond the capability of the Fire Brigade to extinguish. Declaration is not necessary if the agency resources are placed on stand-by, or supporting post-extinguishment recovery or investigation actions.
Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via IC CA6 or SA9.
ONS Basis Reference(s):
: 1. NEI 99-01 HU4 I[D~cument No.]1                                  Rev. 0                           Page 129 of 240

ATTACHMENT 1 EAL Bases Category:                       H - Hazards and Other Conditions Affecting Plant Safety Subcategory:                    5 - Hazardous Gases Initiating Condition:           Gaseous release IMPEDING access to equipment necessary for normal plant operations, cooldown or shutdown EAL:
HA5.1              Alert Release of a toxic, corrosive, asphyxiant or flanmmable gas into any Table H-2 rooms or areas AND Entry into the room or area is prohibited or IMPEDED (Note 5)*
Note 5:    If the equipment in the listed room or area was already inoperable or out-of-service before the event occurred, then no emergency classification is warranted.
Table H-2           Safe Operation & Shutdown RoomslAreas Room/Area                                        Mode Applicability Turbine Building                                                                         1, 2, 3 Equipment and Cable Rooms                                                                1, 2, 3 Auxiliary Building                                                                    1, 2, 3,4, 5 Reactor Buildings                                                                        3, 4, 5 Mode Applicability:
All Definition(s):
IMPEDE(D) - Personnel access to a room or area is hindered to an extent that extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., requiring use of protective equipment, such as SCBAs, that is not routinely employed).
Basis:
If the equipment in the listed room or area was already inoperable, or out-of-service, before the event, occurred, then no emergency should be declared since the event will have no adverse impact beyond that already allowed by Technical Specifications at the time of the event.
The .list of plant rooms or areas with entry-related mode applicability identified specify those rooms or areas that contain equipment which require a manual/local action as specified in operating procedures used for normal plant operation, cooldown and shutdown. Rooms or areas in which actions of a contingent or emergency nature would be performed (e.g., an action to address an off-normal or emergency condition such as emergency repairs, corrective measures or emergency operations) are not included. In addition, the list specifies the plant mode(s) during which entry would be required for each room or area (ref. 1).
I[Document NO.]                                             Rev. 0                                      Page 130 of 2401

ATTACHMENT 1 EAL Bases This IC addresses an event involving a release of a hazardous gas that precludes or impedes access to equipment necessary to maintain normal plant operation, or required for a normal plant cooldown and shutdown. This condition represents an actual or potential substantial degradation of the level of safety of the plant.
An Alert declaration is warranted if entry into the affected room/area is, or may be, procedurally required during the plant operating mode in effect at the time of the gaseous release. The emergency classification is not contingent upon whether entry is actually necessary at the time of the release.
Evaluation of the IC and EAL do not require atmospheri* sampling; it only requires the Emergency Coordinator's judgment that the gas concentration in the affected room/area is sufficient to preclude or significantly impede procedurally required access. This judgment may be based on a variety of factors including an existing job hazard analysis, report of ill effects on personnel, advice from a subject matter expert or operating experience with the same or similar hazards. Access should be considered as impeded if extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., requiring use of protective equipment, such as SCBAs, that is not routinely employed).
An emergency declaration is not warranted if any of the following conditions apply:
* The plant is in an operating mode different than the mode specified for the affected room/area (i.e., entry is not required during the operating mode in effect at the time of the gaseous release). For example, the plant is in Mode 1 when the gaseous release occurs, and the procedures used for normal operation, cooldown and shutdown do not require entry into the affected room until Mode 4.
   *The gas release is a planned activity that includes compensatory measures which address the temporary inaccessibility of a room or area (e.g., fire suppression system testing).
   *The action for which room/area entry is required is of an administrative or record keeping nature (e.g., normal rounds or routine inspections).
   *The access control measures are of a conservative or precautionary nature, and would not actually prevent or impede a required action.
   *If the equipment in the listed room or area was already inoperable, or out-of-service, before the event occurred, then no emergency should be declared since the event will have no adverse impact beyond that already allowed by Technical Specifications at the time of the event.
An asphyxiant is a gas capable of reducing the level of oxygen in the body to dangerous levels. Most commonly, asphyxiants work by merely displacing air in an enclosed environment.
This reduces the concentration of oxygen below the normal level of around 19%, which can lead to breathing difficulties, unconsciousness or even death.
This EAL does not apply to firefighting activities that automatically or manually activate a fire suppression system in an area..
Escalation of the emergency classification level would be via Recognition Category R, C or F I~s.
I[Document No.]                                  Rev. 0                           Page 131 of 240I

ATTACHMENT 1 EAL Bases ONS Basis Reference(s):
: 1. Attachment 3 Safe Operation & Shutdown Rooms/Areas Tables R-3 & H-2 Bases
: 2. NE! 99-01 HA5 S[Document No.]                          Rev. 0                       Page 132 of 240J

ATTACHMENT 1 EAL Bases Category:                 H - Hazards and Other Conditions Affecting Plant Safety Subcategory:              6 - Control Room Evacuation' Initiating Condition:     Control Room evacuation resulting in transfer of plant control to alternate locations EAL:
HA6.1            Alert An event has resulted in plant control being transferred from the Control Room to the Auxiliary Shutdown Panel or Standby Shutdown Facility Mode Applicability:
All Definition(s):
None Basis:
The Control Room Supervisor (CRS) determines if the Control Room is uninhabitable and requires evacuation. Control Room inhabitability may be caused by fire, dense smoke, noxious fumes, bomb threat in or adjacent to the Control Room, or other life threatening conditions.
AP/1 ,2,3/A/1700/008, Loss of Control Room, provides the instructions for tripping the unit, and maintaining RCS inventory and Hot Shutdown conditions from outside the Control Room due to reasons other than fire (ref. 1).
AP/1 ,2,3/A/1700/050, Challenging Plant Fire, provides the instructions for tripping the unit, and maintaining RCS inventory and Hot Shutdown conditions from outside the Control Room due to a fire (ref. 2).
If normal post-trip conditions cannot be maintained with the Auxiliary Shutdown Panel or there is a challenging fire in an SSF risk area, plant shutdown may be directed from the Standby Shutdown Facility (ref. 3, 4).
Inability to establish plant control from outside the Control Room escalates this event to a Site Area Emergency per EAL HS6.1.
This IC addresses an evacuation of the Control Room that results in transfer of plant control to alternate locations outside the Control Room. The loss of the ability to control the plant from the Control Room is considered to be a potential substantial degradation in the level of plant safety.
Following a Control Room evacuation, control of the plant will be transferred to alternate shutdown locations. The necessity to control a plant shutdown from outside the Control Room, in addition to responding to the event that required the evacuation of the Control Room, will present challenges to plant operators and other on-shift personnel. Activation of the ERO and emergency response facilities will assist in responding to these challenges.
Escalation of the emergency classification level would be via IC HS6.
I[Document No.]                                  Rev. 0                         Page.133 of 240

ATTACHMENT 1 EAL Bases ONS Basis Reference(s):
: 1. AP/1 ,2,3/A/1 700/008 Loss of Control Room
: 2. AP/1 ,2,3/A/1 700/050 Challenging Plant Fire
: 3. AP/0/A/1700/025 Standby Shutdown Facility Emergency Operating Procedure
: 4. AP/0/A/1700/043 Fire Brigade Response Procedure
: 5. NEl 99-01 HA6
[Document No.]                              Rev. 0                    Page 134 of 240I

ATTACHMENT 1 EAL Bases Category:                   H - Hazards and Other Conditions Affecting Plant Safety Subcategory:                6 - Control Room Evacuation
                                                                                                        /
Initiating Condition:       Inability to control a key safety function from outside the Control Room EAL:
HS6.1            Site Area Emergency An event has resulted in plant control being transferred from the Control Room to the Auxiliary Shutdown Panel or Standby Shutdown Facility AND Control of any of the following key safety functions is not re-established within 15 min.
(Note 1):                                                                                                I
* Reactivity (Modes 1, 2 and 3 only)
* Core cooling
* RCS heat removal Note 1:    The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
Mode Applicability:
1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown, 5 - Cold Shutdown, 6 -
Refuel Definition(s):
None Basis:
The Control Room Supervisor (CRS) determines if the Control Room is uninhabitable and requires evacuation. Control Room inhabitability may be caused by fire, dense smoke, noxious fumes, bomb threat in or adjacent to the Control Room, or other life threatening conditions.
AP/1,2,3/A/1700/008, Loss of Control Room, provides the instructions for tripping the unit, and maintaining RCS inventory and Hot Shutdown conditions from outside the Control Room due to reasons other than fire (ref. 1).
AP/1 ,2,3/A/1700/050, Challenging Plant Fire, provides the instructions for tripping the unit, and maintaining ROS inventory and Hot Shutdown conditions from outside the Control Room due to a fire (ref. 2).
If normal post-trip conditions cannot be maintained with the Auxiliary Shutdown Panel or there is a challenging fire in an SSF risk area, plant shutdown may be directed from the Standby Shutdown Facility (ref. 3, 4).
The intent of this EAL is to capture events in which control of the plant cannot be reestablished in a timely manner. The fifteen minute time for transfer starts when the Control Room is evacuated (when CRS leaves the Control Room, not when AP/1 ,2,3/A/1 700/008 or AP/1 ,2,3/A/1700/050 is entered). The time interval is based on how quickly control must be S[Document No.]                                        Rev. 0                            Page 135 of 240I

ATTACHMENT 1 EAL Bases reestablished without core uncovery and/or core damage. The determination of whether or not control is established from outside the Control Room is based on Emergency Coordinator judgment. The Emergency Coordinator is expected to make a reasonable, informed judgment that control of the plant from outside the Control Room cannot be established within the fifteen minute interval.
Once the Control Room is evacuated, the objective is to establish control of important plant equipment and maintain knowledge of important plant parameters in a timely manner. Primary emphasis should be placed on components and instruments that supply protection for and information about safety functions. Typically, these safety functions are reactivity control (ability to shutdown the reactor and maintain it shutdown), RCS inventory (ability to cool the core), and secondary heat removal (ability to maintain a heat sink).
This IC addresses an evacuation of the Control Room that results in transfer of plant control to alternate locations, and the control of a key safety function cannot be reestablished in a timely manner. The failure to gain control of a key safety function following a transfer of plant control to alternate locations is a precursor to a challenge to one or more fission product barriers within a relatively short period of time.
The determination of whether or not "control" is established at the remote safe shutdown location(s) is based on Emergency Coordinator judgment. The Emergency Coordinator is expected to make a reasonable, informed judgment within 15 minutes whether or not the operating staff has control of key safety functions from the remote safe shutdown location(s).
Escalation of the emergency classification level would be via IC FG1 or CG1 ONS Basis Reference(s):
: 1. AP/1 ,2,3/A/11700/008 Loss of Control Room
: 2. AP/1,2,3/A/1700/050 Challenging Plant Fire
: 3. AP/0/A/1700/025 Standby Shutdown Facility Emergency Operating Procedure
: 4. AP/0/A/1700/043 Fire Brigade Response Procedure
: 5. NEI 99-01 HS6
[Document No.]                                 Rev. 0                            Page 136 of 240J

ATTACHMENT 1 EAL Bases Category:                    H - Hazards and Other conditions Affecting Plant Safety Subcategory:                 7 - Emergency Coordinator Judgment Initiating Condition:        Other conditions existing that' in the judgment of the Emergency Coordinator warrant declaration of a UE EAL:
HU7.1                   Unusual Event Other conditions exist which in the judgment of the Errergency Coordinator 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 uless further degradation of SAFETY SYSTEMS occurs.
Mode Applicability:
All Definition(s):
SAFETY SYSTEM -A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2):
Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:
(1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures..
Basis:
The Emergency Coordinator is the designated onsite individual having the responsibility and authority for implementing the ONS Emergency Plan (ref. 1). The Operations Shift Manager (SM) initially acts in the capacity of the Emergency Coordinator and takes actions as outlined in the Emergency Plan implementing procedures (ref. 2). If required by the emergency classification or if deemed appropriate by the Emergency Coordinator, emergency response personnel are notified and instructed to report to their emergency response locations. In this manner, the individual usually in charge of activities in the Control Room is responsible for initiating the necessary emergency response, but plant management is expected to manage the emergency response as soon as available to do so in anticipation of the possible wide-ranging responsibilities associated with managing a major emergency.
This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Coordinator to fall under the emergency classification level description for an Unusual Event.
S[Document No.]                                  Rev. 0                             Page 137 of 240

ATTACHMENT 1 EAL Bases ONS Basis Reference(s):
: 1. ONS Emergency Plan Section A Assignment of Responsibility
: 2. RP/0/A/1 000/001 Emergency Classification
: 3. NEI 99-01 HU7 Rev.'0 I
[Document No.]                            Rev.' 0                 138 of 240 I '
Page 138

ATTACHMENT 1 EAL Bases Category:                  H - Hazards and Other Conditions Affecting Plant Safety Subcategory:               7 - Emergency Coordinator Judgment Initiating Condition:      Other conditions exist that in the judgment of the Emergency Coordinator warrant declaration of an Alert EAL:
HA7.1           Alert Other conditions dxist which, in the judgment of the Emergency Coordiihator, indicate that events are in progress or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant or a security event that involves probable life threatening risk to site personnel or damage to site equipment because of HOSTILE ACTION. Any releases are expected to be limited to small fractions of the EPA Protective Action Guideline exposure levels.
Mode Applicability:
All Definition(s):
HOSTILE ACTION - An act toward ONS 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 ONS. Non-terrorism-based EALs should be used to address such activities (ie.e, this may include violent acts between individuals in the owner controlled area).
Basis:
The Emergency Coordinator is the designated onsite individual having the responsibility and authority for implementing the ONS Emergency Plan (ref. 1). The Operations Shift Manager (SM) initially acts in the capacity of the Emergency Coordinator and takes actions as outlined in the Emergency Plan implementing procedures (ref. 2). If required by the emergency classification or if deemed appropriate by the Emergency Coordinator, emergency response personnel are notified and instructed to report to their emergency response locations. In this manner, the individual usually in charge of activities in the Control Room is responsible for initiating the necessary emergency response, but Plant Management is expected to manage the emergency response as soon as available to do so in anticipation of the possible wide-ranging responsibilities associated with managing a major emergency.
This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Coordinator to fall under the emergency classification level description for an Alert.
S[Document No.]                                Rev. 0                            Page 139 of 240I

ATTACHMENT 1 EAL Bases ONS Basis Reference(s):
: 1. ONS Emergency Plan Section A Assignment of Responsibility
: 2. RP/0/A/1 000/001 Emergency Classification
: 3. NEI 99-01 HA7 S[Document No.]                           Rev. 0              Page 140 of 240

ATTACHMENT 1 EAL Bases Category:                    H - Hazards and Other Conditions Affecting Plant Safety S ubcategory:               7 - Emergency Coordinator Judgment Initiating Condition:      Other conditions existing that in the judgment of the Emergency Coordinator warrant declaration of a Site Area Emergency EAL:
HS7.1          Site Area Emergency Other conditions exist which in the judgment of the Emergency Coordinator indicate that events are in progress or have occurred which involve actual or likely major failures of plant functions needed for protection of the public or HOSTILE ACTION that results in intentional damage or malicious acts, (1) toward site personnel or equipment that could lead to the likely failure of or, (2) that prevent effective access to equipment needed for the protection of the public. Any releases are not expected to result in exposure levels which exceed EPA Protective Action Guideline exposure levels beyond the SITE BOUNDARY Mode Applicability:
All Definition(s):
HOSTILE ACTION - An act toward ONS 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 ONS. 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).
SITE BOUNDARY- That area, including the PROTECTED AREA, in which Duke Energy has the authority to control all activities including exclusion or removal of personnel and property (1 mile radius) from the center of Unit 2.
Basis:
The Emergency Coordinator is the designated onsite individual having the responsibility and authority for implementing the ONS Emergency Plan (ref. 1). The Operations Shift Manager (SM) initially acts in the capacity of the Emergency Coordinator and takes actions as outlined in the Emergency Plan implementing procedures (ref. 2). If required by the emergency classification or if deemed appropriate by the Emergency Coordinator, emergency response personnel are notified and instructed to report to their emergency response locations. In this manner, the individual usually in charge of activities in the Control Room is responsible for initiating the necessary emergency response, but Plant Management is expected to manage the emergency response as soon as available to do so in anticipation of the possible wide-ranging responsibilities associated with managing a major emergency.
S[Document No.]                           .        Rev. 0                         Page 141 of 240,

ATTACHMENT 1 EAL Bases This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Coordinator to fall under the emergency classification level description for a Site Area Emergency.
ONS Basis Reference(s):
: 1. ONS Emergency Plan Section A Assignment of Responsibility
: 2. RP/0/N/1000/001 Emergency Classification
: 3. NEI 99-01 HS7 S[Document No.]                              Rev.' 0                            Page 142 of 240

ATTACHMENT 1 EAL Bases Category:                 H - Hazards and Other Conditions Affecting Plant Safety Subcategory:              7 - Emergency Coordinator Judgment Initiating Condition:     Other conditions exist which in the judgment of the Emergency Coordinator warrant declaration of a General Emergency EAL:
HG7.1                  General Emergency Other conditions exist which in the judgment* of the Emergency Coordinator indicate that events are in progress or have occurred which involve actual Or IMMINENT substantial core degradation or melting with potential for loss of containment integrity or HOSTILE ACTION that results in an actual loss of physical control of the facility. Releases can be reasonably expected to exceed EPA Protective Action Guideline exposure levels offsite for more than the immediate site area.
Mode Applicability:
All Definition(s):
HOSTILE ACTION - An act toward ONS 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 ONS. 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).
IMMINENT - The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions.
Basis:
The Emergency Coordinator is the designated onsite individual having the responsibility and authority for implementing the ONS Emergency Plan (ref. 1). The Operations Shift Manager (SM) initially acts in the capacity of the Emergency Coordinator and takes actions as outlined in the Emergency Plan implementing procedures (ref. 2). If required by the emergency classification or if deemed appropriate by the Emergency Coordinator, emergency response personnel are notified and instructed to report to their emergency response locations. In this manner, the individual usually in charge of activities in the Control Room is responsible for initiating the necessary emergency response, but Plant Management is expected to manage the emergency response as soon as available to do so in anticipation of the possible wide-ranging responsibilities associated with managing a major emergency.
Releases can reasonably be expected to exceed EPA PAG plume exposure levels outside the Site Boundary.
S[Document No.]                                Rev.O0                            Page143 of 240

ATTACHMENT 1 EAL Bases This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Coordinator to fall under the emergency classification level description for a General Emergency.
ONS Basis Reference(s):
: 1. ONS Emergency Plan Section A Assignment of Responsibility
: 2. RP/0/A/1 000/001 Emergency Classification
: 13. NEI 99-01 HG7I I[Document No.]                               Rev. 0                             Page 144 of 240

ATTACHMENT 1 EAL Bases Category S     -  System Malfunction EAL Group: Hot Conditions (RCS temperature       > 210&deg;F); EALs in this category are applicable only in one or more hot operating modes.
Numerous system-related equipment failure events that warrant emergency classification have been identified in this category. They may pose actual or potential threats to plant safety.
The events of this category pertain to the following subcategories:
: 1. Ldss of Essential AC PowerI Loss of essential electrical power can compromise plant safety system operability including decay heat removal and emergency core cooling systems which may be necessary to ensure fission product barrier integrity. This category includes loss of onsite and offsite sources for 4160V AC essential buses.
: 2. Loss of Vital DC Power Loss of emergency electrical power can compromise plant safety system operability including decay heat removal and emergency core cooling systems which may be necessary to ensure fission product barrier integrity. This category includes loss of vital plant 125V DC power sources.
: 3. Loss of Control Room Indications Certain events that degrade plant operator ability to effectively assess plant conditions within the plant warrant emergency classification. Losses of indicators are in this subcategory.
: 4. RCS Activity During normal operation, reactor coolant fission product activity is very low. Small concentrations of fission products in the coolant are primarily from the fission of tramp uranium in the fuel clad or minor perforations in the clad itself. Any significant increase from these base-line levels (2% - 5% clad failures) is indicative of fuel failures and is covered under the Fission Product Barrier Degradation category. However, lesser amounts of clad damage may result in coolant activity exceeding Technical Specification limits.. These fission products will be circulated with the reactor coolant and can be detected by coolant sampling.
: 5. RCS Leakage The reactor vessel provides a volume for the coolant that covers the reactor core. The reactor pressure vessel and associated pressure piping (reactor coolant system) together provide a barrier to limit the release of radioactive material should the reactor fuel clad integrity fail. Excessive RCS leakage greater than Technical Specification limits indicates potential pipe cracks that may propagate to an extent threatening fuel clad, RCS and containment integrity.
: 6. RPS Failure This subcategory includes events related to failure of the Reactor Protective System (RPS) to initiate and complete reactor trips. In the plant licensing basis, postulated failures of the RPS to complete a reactor trip comprise a specific set of analyzed events referred to as I[Document No.]
* Rev. 0                     .      Page 145 of 240I

ATTACHMENT 1 EAL Bases Anticipated Transient Without Scram (ATWS) events. For EAL classification, however, ATWS is intended to mean any trip failure event that does not achieve reactor shutdown. If RPS actuation fails to assure reactor shutdown, positive control of reactivity is at risk and could cause a threat to fuel clad, RCS and containment integrity.
: 7. Loss of Communications Certain events that degrade plant operator ability to effectively communicate with essential personnel within or external to the plant warrant emergency classification.
: 8. Containment Failure Failure of containment isolation capability (under conditions in which the containment is not currently challenged) warrants emergency classification. Failure of containment pressure control capability also warrants emergency classification.
: 9. Hazardous Event Affectina Safety Systems Various natural and technological events that result in degraded plant safety system performance or significant visible damage warrant emergency classification under this subcategory.
J[Document No.]                      J         Rev. 0                          Page 146 of 240

ATTACHMENT 1 EAL Bases Category:                  S   -   System Malfunction Subcategory:                1 - Loss of Essential AC Power Initiating Condition:     Loss of all offsite AC power capability to essential buses for 15 minutes or longer EAL:
SUI.1           Unusual Event Loss of all offsite AC povw'er capability, Table S-i, to essential 4160 V buses Mv~FB-1 and MFB-2 for > 15 mai. (Note 1)
Note 1:   The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or sWill likely be exceeded.
Table S-I     AC Power Sources Offsite:
* Unit Normal Transformer (backcharged)
* Unit Startup Transformer (SWYD)
* Another Unit Startup Transformer (aligned) (SWYD)
* CT5 (Central/energizing Standby Bus)
Emergency:
* Unit Startup Transformer (Keowee)
* Another Unit Startup Transformer (aligned) (Keowee)
* CT4
* CT5 (dedicated line/energizing Standby Bus)
Mode Applicability:
1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):
None Basis:
The 4160 V AC System provides the power requirements for operation and safe shutdown of the plant. 4160 V buses MFB-1 and MFB-2 are the essential buses (ref. 1).
The condition indicated by this EAL is the degradation of all offsite AC power sources such that only onsite AC power capability exists for 1,5 minutes or longer.
Each unit is provided with two physically independent circuits from the switching station. One is the circuit from the 230 kV switching station through the startup transformer, which is designed to be available within a few seconds following a loss of coolant accident. The second circuit is the path from the switchyard through the main step-up transformer, the main generator bus and the unit auxiliary transformer with the generator disconnected from the main bus. The second circuit is currently used during refueling as an additional power feed for the shutdown II[Document    No.]                                     Rev. 0                           Page 147 .of 240I

ATTACHMENT 1 EAL Bases unit(s) from the 230 kV switchyard. Whenever there is inadequate power from the generating units, the 230 kV switching station and the hydro units, power is available to the standby power buses either directly from the 100 kV Central Tie Substation or from Lee Steam Station via Transformer CT5. (ref. 2)
Upon loss of power from the Oconee generating unit and 230 kV switchyard, power is supplied from both Keowee Hydro Station units through two separate and independent routes. One route is an underground feeder to Transformer CT4 which supplies the two redundant Main Feeder Buses (MFB-1 and MFB-2). The other route is an overhead feeder to the 230 KV switching station which supplies each unit startup transformer. (ref. 3)
The Standby Shutdown Facility (SSF) consists of standby systems for use in an extreme emergency and is equipped with a manually started diesel generator that can supply power necessary to maintain hot shutdown of the reactors of each unit in the event of loss of ,power.
However, the SSF cannot supply'power to the essential buses and therefore not credited in this EAL (ref. 3).
The 15-minute interval was selected as a threshold to exclude transient or momentary power losses.
This IC addresses a prolonged loss of offsite power. The loss of offsite power sources renders the plant more vulnerable to a complete loss of power to AC essential buses. This condition represents a potential reduction in the level of safety of the plant.
For emergency classification purposes, "capability" means that an offsite AC power source(s) is available to the essential buses, whether or not the buses are powered from it.
Escalation of the emergency classification level would be via IC SAI.
ONS Basis Reference(s):
: 1. UFSAR Figure 8.1 Single Line Diagram
: 2. UFSAR Section 8.20Offsite Power System
: 3. UFSAR Section 8.30Onsite Power Systems
: 4. NEI 99-01 SU1I'
[Document No.]                                Rev. 0                           Page 148 of 240

ATTACHMENT 1 EAL Bases Category:                   S  -  System Malfunction Subcategory:                1 - Loss of Essential AC Power Initiating Condition:       Loss of all but one AC power source to essential buses for 15 minutes or longer EAL:'
SAI.1           Alert AC power capability, Table S-1, to essential 4160 V buses MFB-1 and MFB-2 reduced to a single power source for &#x17d;- 15 min. (Note 1)
AND Any additional single power source failure will result in loss of all AC power to SAFETY SYSTEMS Note 1:    The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
Table S-I     AC Power Sources Offsite:
* Unit Normal Transformer (backcharged)
* Unit Startup Transformer (SWYD)
* Another Unit Startup Transformer (aligned) (SWYD)
* CT5 (Central/energizing Standby Bus)
Emergency:
* Unit Startup Transformer (Keowee)
* Another Unit Startup Transformer (aligned) (Keowee)
* CT4
* CT5 (dedicated line/energizing Standby Bus)
Mode Applicability:
1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4         - Hot Shutdown Definition(s):
SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10OCFR50.2):
Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:
(1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; Rv                                Pg       49o        *4
[DcmntN.
*I[Document No.]                                        Rev. 0 Page 149 of 240 I

ATTACHMENT 1 EAL Bases (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.
Basis:
For emergency classification purposes, "capability" means that an AC power source is available to and capable of powering the emergency bus(es) within 15 min, whether or not the buses are currently powered from it.
The 4160 V AC System provides the power requirements for operation and safe shutdown of the plant. 4160 V buses MFB-1 and MFB-* are the essential buses (ref. 1).
The condition indicated by this EAL is the degradation of the offsite and onsite power sources such that any additional single failure would result in a loss of all AC power to the essential buses.
Each unit is provided with two physically independent circuits from the switching station. One is the circuit from the 230 kV switching station through the startup transformer, which is designed to be available within a few seconds following a loss of coolant accident. The second circuit is the path from the switchyard through the main step-up transformer, the main generator bus and the unit auxiliary transformer with the generator disconnected from the main bus. The second circuit is currently used during refueling as an additional power feed for the shutdown unit(s) from the 230 kV switchyard. Whenever there is inadequate power from the generating units, the 230 kV switching station and the hydro units, power is available to the standby power buses either directly from the 100 kV Central Tie Substation or from Lee Steam Station via a Transformer CT5. (ref. 2)
Upon loss of power from the Oconee generating unit and 230 kV switchyard, power is supplied from both Keowee Hydro Station units through two separate and independent routes. One route is an underground feeder to Transformer CT4 which supplies the two redundant Main Feeder Buses (MFB-1 and MFB-2). The other route is an overhead feeder to the 230 KV switching station which supplies each unit startup transformer. (ref. 3)
The Standby Shutdown Facility (SSF) consists of standby systems for use in an extreme emergency and is equipped with a manually started diesel generator that can supply power necessary to maintain hot shutdown of the reactors of each unit in the event of loss of power.
However, the SSF cannot supply power to the essential buses and therefore not credited in this EAL (ref. 3).
The 15-minute interval was selected as a threshold to exclude transient or momentary power losses. If the capability of a second source of emergency bus power is not restored within 15 minutes, an Alert is declared under this EAL.
This IC describes a significant degradation of offsite and onsite AC power sources such that any additional single failure would result in a loss of all AC power to SAFETY SYSTEMS. In this condition, the sole AC power source may be powering one, or more than one, train of safety-related equipment. This IC provides an escalation path from IC SUI.
An "AC power source" is a source recognized in APs and EOPs, and capable of supplying required power to an essential bus. Some examples of this condition are presented below.
* A loss of all offsite power with a concurrent failure of all but one essential power source (e.g., CT4, CT5, CT1, 2 ,3 (Keowee)).
((Document No.]                          [        Rev. 0         [Page                    150 of 240I

ATTACHMENT 1 EAL Bases
       *A loss of essential power sources (e.g., CT4, CT5, CT1, CT2, CT3 (Keowee)) with a single train of essential buses being back-fed from an offsite power source.
Escalation of the emergency classification level would be via IC SSI.
ONS Basis Reference(s):
: 1. UFSAR Figure 8.1 Single Line Diagram
: 2. UFSAR Section 8.2 Offsite Power System
: 3. UFSAR Section 8.30Onsite Power Systems I4. NEI 99-01 SA1 I[Document No.]                       IRev.         0                          Page 151 of 240

ATTACHMENT 1 EAL Bases Category:                  S - System Malfunction Subcategory:               1 - Loss of Essential AC Power Initiating Condition:      Loss of all offsite power and all emergency AC power to essential buses for 15 minutes or longer EAL:
SS1.1           Site Area Emergency Loss o*f all offsite and all emergency AC power capability td essential 4160 V buses MFB-1 and MFB-2 for > 15 min. (Note 1)
Note 1:   The Emergency Coordinator should declare the event promptly upon determining that time limit has I been exceeded, or will likely be exceeded.
Mode Applicability:
1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):
None Basis:
For emergency classification purposes, "capability" means that an AC power source is available to and capable of powering the emergency bus(es) within 15 minutes, whether or not the buses are currently powered from it.
The condition indicated by this EAL is the degradation of the offsite and emergency power sources resulting in a loss of all AC power to the emergency buses. 4160 V buses MFB-1 and MFB-2 are the essential buses (ref. 1).
Each unit is provided with two physically independent circuits from the switching station. One is the circuit from the 230 kV switching station through the startup transformer, which is designed to be available within a few seconds following a loss of coolant accident. The second circuit is the path from the switchyard through the main step-up transformer, the main generator bus and the unit auxiliary transformer with the generator disconnected from the main bus. The second circuit is currently used during refueling as an additional power feed for the shutdown unit(s) from the 230 kV switchyard. Whenever there is inadequate power from the generating units, the 230 kV switching station and the hydro units, power is available to the standby power buses either directly from the 100 kV Central Tie Substation or from Lee Steam Station via a Transformer CT5. (ref. 2)
Upon loss of power from the Oconee generating unit and 230 kV switchyard, power is supplied from both Keowee Hydro Station units through two separate and independent routes. One route is an underground feeder to Transformer CT4 which supplies the two redundant Main Feeder Buses (MFB-1 and MFB-2). The other route is an overhead feeder to the 230 KV switching station which supplies each unit startup transformer. (ref. 3)
The Standby Shutdown Facility (SSF) consists of standby systems for use in an extreme emergency and is equipped with a manually started diesel generator that can supply power necessary to maintain hot shutdown of the reactors of each unit in the event of loss of power.
mI[Document   No.]                                   Rev. 0                            Page 152 of 2401

ATTACHMENT 1 EAL Bases However, the SSF cannot supply power to the essential buses and therefore not credited in this EAL. (ref. 3).
The 15-minute interval was selected as a threshold to exclude transient or momentary power losses. The interval begins when both offsite and onsite AC power capability are lost.
This IC addresses a total loss of AC power that compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink.
In addition, fission product barrier monitoring capabilities may be degraded under these conditions. Thfs IC represents a condition that involves actual or"likely major failures of plant functions needed for the protection of the public.
Escalation of the emergency classification level would be via ICs RG1, FG1 or SGI.
ONS Basis Reference(s):
: 1. UFSAR Figure 8.1 Single Line Diagram
: 2. UFSAR Section 8.20Offsite Power System
: 3. UFSAR Section 8.30Onsite Power Systems
: 4. NEI 99-01 SS1 S[Document No.]                                Rev. 0                          page 153 of 240I

ATTACHMENT 1 EAL Bases Category:                  S -System Malfunction S ubcategory:               1 - Loss of Essential AC Power Initiating Condition:       Prolonged loss of all offsite and all emergency AC power to essential buses EAL:
SGI.1           General Emergency Loss of all offsite and all emergency AC power capability to essential 4160 \* buses MEB-1 and MFB-2 AND Failure to power 5SF equipment and PSW unavailable AND EITHER:
* Restoration of at least one essential bus in < 4 hour is not likely (Note 1)
* CETC reading > 1200&deg;F Note 1:   The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
Mode Applicability:
1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):
None Basis:
For emergency classification purposes, "capability" means that an AC power source is available to and capable of powering the essential bus(es), whether or not the buses are currently powered from it.~4160 V buses MFB-1 and MFB-2 are the essential buses (ref. 1).
Each unit is provided with two physically independent circuits from the switching station. One is the circuit from the 230 kV switching station through the startup transformer, which is designed to be available within a few seconds following a loss of coolant accident. The second circuit is the path from the switchyard through the main step-up transformer, the main generator bus and the unit auxiliary transformer with the generator disconnected from the main bus. The second circuit is currently used during refueling as an additional power feed for the shutdown unit(s) from the 230 kV switchyard. Whenever there is inadequate power from the generating units, the 230 kV switching station and the hydro units, power is available to the standby power buses either directly from the 100 kV Central Tie Substation or from Lee Steam Station via Transformer CT5. (ref. 2)
Upon loss of power from the Oconee generating unit and 230 kV switchyard, power is supplied from both Keowee Hydro Station units through two separate and independent routes. One route is an underground feeder to Transformer CT4 which supplies the two redundant Main S[Document No.]                                       Rev. 0                            Page 154 of 240

ATTACHMENT 1 EAL Bases Feeder Buses (MFB-1 and MFB-2). The other route is an overhead feeder to the 230 KV switching station which supplies each unit startup transformer. (ref. 3)
The Standby Shutdown Facility (SSF) consists of standby systems for use in an extreme emergency and is equipped with a manually started diesel generator and Protected Service Water (PSW) power supply that can supply power necessary to maintain hot shutdown of the reactors of each unit in the event of loss of power. Although the SSF requires manual initiation, it is considered in this EAL because it may be capable of powering the SSF load center (ref. 3, 4).
The station blackout coping peribd is four hours (ref. 5).
Core Exit Thermocouple readings of 1200&deg;F are indicative of superheat conditions and inability to adequately remove heat from the core (ref. 6).-
This IC addresses a prolonged toss of all power sources to AC essential buses. A Io'ss of all AC power compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink. A prolonged loss of these buses will lead to a loss of one or more fission product barriers. In addition, fission product barrier monitoring capabilities may be degraded under these conditions.
The EAL should require declaration of a General Emergency prior to meeting the thresholds for IC FGI. This will allow additional time for implementation of offsite protective actions.
Escalation of the emergency classification from Site Area Emergency will occur if it is projected that power cannot be restored to at least one AC essential bus by the end of the analyzed station blackout coping period. Beyond this time, plant responses and event trajectory are subject to greater uncertainty, and there is an increased likelihood of challenges to multiple fission product barriers.
The estimate for restoring at least one essential bus should be based on a realistic appraisal of the situation. Mitigation actions with a low probability of success should not be used as a basis for delaying a classification upgrade. The goal is to maximize the time available to prepare for, and implement, protective actions for the public.
The EAL will also require a General Emergency declaration if the loss of AC power results in parameters that indicate an inability to adequately remove decay heat from the core.
ONS Basis Reference(s):
: 1. UFSAR Figure 8.1 Single Line Diagram
: 2. UFSAR Section 8.20Offsite Power System
: 3. UFSAR Section 8.3 Onsite Power Systems
: 4. UFSAR Section 9.6 Standby Shutdown Facility
: 5. UFSAR Section 8.3.2.2.4 Station Blackout Analysis
: 6. RP/O/A/1000/18 Core Damage Assessment
: 7. NEI 99-01 SGI I[Document No.]                                 Rev. 0                             Page 155 of 240

ATTACHMENT 1 EAL Bases Category:                  S -System Malfunction Subcategory:               1 - Loss of Essential AC Power Initiating Condition:      Loss of all essential AC and vital DC power sources for 15 minutes or longer EAL:
SG1.2           General Emergency Loss of all offsite and all emergency ACipower capability to essential 4160 V buses MEB-1 and MFB-2 for > 15 min.
AND Failure to power SSF equipment and PSW unavailable AND Loss of 125 VDC power based on battery bus voltage indications      <  105 VDC on both vital DC Distribution Centers DCA and DCB for > 15 min. (Note 1) been exceeded, or will likely be exceeded.
Mode Applicability:
1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):
None Basis:
This EAL is indicated by the loss of all offsite and emergency AC power capability to 4160 V essential buses MFB-1 and MFB-2 for greater than 15 minutes in combination with degraded vital DC power voltage. This EAL addresses operating experience from the March 2011 accident at Fukushima Daiichi.
For emergency classification purposes, "capability" means that an AC power source is available to and capable of powering the emergency bus(es) within 15 minutes, whether or not the buses are currently powered from it.
Each unit is provided with two physically independent circuits from the switching station. One is the circuit from the 230 kV switching station through the startup transformer, which is designed to be available within a few seconds following a loss of coolant accident. The second circuit is the path from the switchyard through the main step-up transformer, the main generator bus and the unit auxiliary transformer with the generator disconnected from the main bus. The second circuit is currently used during refueling as an additional power feed for the shutdown unit(s) from the 230 kV switchyard. Whenever there is inadequate power from the generating units, the 230 kV switching station and the hydro units, power is available to the standby power buses either directly from the 100 kV Central Tie Substation or from Lee Steam Station via Transformer CT5. (ref. 2) e..                        ae   5    f20
[DcmnIo]
I [Document No.]                                       Rev. 0 Page 156 of 240 I

ATTACHMENT 1 EAL Bases Upon loss of power from the Oconee generating unit and 230 kV switchyard, power is supplied from both Keowee Hydro Station units through two separate and independent routes. One route is an underground feeder to Transformer CT4 which supplies the two redundant Main Feeder Buses (MFB-1 and MFB-2). The other route is an overhead feeder to the 230 KV switching station which supplies each unit startup transformer. (ref. 3)
The Standby Shutdown Facility (SSF) consists of standby systems for use in an extreme emergency and is equipped with a manually started diesel generator and Protected Service Water (PSW) power supply that can supply power necessary to maintain hot shutdown of the reactors of each unit in the event of loss of powe~r. Although the SSF requires manual initiation, it is considered in this EAL because it may be capable of powering the SSF load center (ref. 3).
For each unit, two independent and physically separated 125 volt DC batteries and DC buses are provided for the vital instrumentation and control power system. (ref. 4, 5). Minimum DC bus voltage is 105 VDC (ref. 6).
This IC addresses a concurrent and prolonged loss of both emergency AC and Vital DC power. A loss of all emergency AC power compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink.
A loss of vital DC power compromises the ability to monitor and control SAFETY SYSTEMS.
A sustained loss of both emergency AC and vital DC power will lead to multiple challenges to fission product barriers.
Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.
The 15-minute emergency declaration clock begins at the point when both EAL thresholds are met.
ONS Basis Reference(s):
: 1. UFSAR Figure 8.1 Single Line Diagram
: 2. UFSAR Section 8.20Offsite Power System
: 3. UFSAR Section 8.30Onsite Power Systems
: 4. UFSAR Figure 8.5 Typical DC and AC Vital Power System - Single Line
: 5. UFSAR Section 8.3.2 DC Power Systems
: 6. EP/*/A/1800/001 Blackout Tab
: 7. NEI 99-01 SG8 S[Document No.]                                Rev. 0                          Page 157 of 240

ATTACHMENT 1 EAL Bases Category:                   S - System Malfunction Subcategory:                2 - Loss of Vital DC Power Initiating Condition:      Loss of all vital DC power for 15 minutes or longer EAL:
Note 1:  The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
Mode Applicability:I 1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):
None Basis:
For each unit, two independent and physically separated 125 volt DC batteries and DC buses are provided for the vital instrumentation and control power system. (ref. 1, 2). Minimum DC bus voltage is 105 VDC (ref. 3).
This IC addresses a loss of vi~tal DC power which compromises the ability to monitor and control SAFETY SYSTEMS. In modes above Cold Shutdown, this condition involves a major failure of plant functions needed for the protection of the public.
Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.
Escalation of the emergency classification level would be via ICs RG1, FG1 or SGI.
ONS Basis Reference(s):
: 1. UFSAR Figure 8.5 Typical DC and AC Vital Power System - Single Line
: 2. UFSAR Section 8.3.2.DC Power Systems
: 3. PI*/IAI1 800/00 1 Blackout Tab
: 4. Technical Specifications 3.8.3 DC Sources             - Operating
: 5. NEI 99-01 SS8 S[Document No.]                                       Rev. 0                           Page 158 of 240

ATTACHMENT 1 EAL Bases Category:                   S - System Malfunction Subcategory:                3 - Loss of Control Room Indications Initiating Condition:        UNPLANNED loss of Control Room indications for 15 minutes or longer EAL:
SU3.1            Unusual Event An UNPLANNED event results in the inability to monitor one or mbre Table S-2 parameters from within the Control Room for --15 min. (Note 1)
Note 1:    The Emergency Coordinator should declare the event promptly upon determining that time limit has been! exceeded, or will likely be exceeded.
Table S-2       Safety System Parameters
* Reactor power
* RCS level
* RCS pressure
* CETC temperature
* Level in at least one S/G
* EFW flow to at least one SIG Mode Applicability:
1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):
UNPLANNED - A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.
Basis:
SAFETY SYSTEM parameters listed in Table S-2 are monitored in the Control Room through a combination of hard control panel indicators as well as computer based information systems.
The SPDS serves as a redundant compensatory indicator which may be utilized in lieu of normal Control Room indicators (ref. 1).
This IC addresses the difficulty associated with monitoring normal plant conditions without the ability to obtain SAFETY SYSTEM parameters from within the Control Room. This condition is a precursor to a more significant event and represents a potential degradation in the level of safety of the plant.
As used in this EAL, an "inability to monitor" means that values for one or more of the listed parameters cannot be determined from within the Control Room. This situation would require a loss of all of the Control Room sources for the given parameter(s). For example, the reactor power level cannot be determined from any analog, digital and recorder source within the Control Room.
S[Document No.]                                         Rev. 0                           Page 159 of 240

ATTACHMENT 1 EAL Bases An event involving a loss of plant indications, annunciators and/or display systems is evaluated in accordance with 10 CER 50.72 (and associated guidance in NUREG-1022) to determine if an NRC event report is required. The event would be reported if it significantly impaired the capability to perform emergency assessments. In particular, emergency assessments necessary to implement abnormal operating procedures, emergency operating procedures, and emergency plan implementing procedures addressing emergency classification, accident assessment, or protective action decision-making.
This EAL is focused on a selected subset of plant parameters associated with the key safety functions of reactivity control, core cooling and RCS heat removal.i The loss of the ability to determine one or more of these parameters from within the Control Room is considered to be more significant than simply a reportable condition. In addition, if all indication sources for one or more of the listed parameters are lost, then the ability to determine the values of other SAFETY SYSTEM parameters may be impacted as well. For example, if the value for reactor vessel level cannot be determined from the indications and recorders on a main control board, the SPDS or the plant computer, the availability of other parameter values may be" compromised as well.
Fifteen minutes was selected as a threshold to exclude transient or momentary losses of indication.
Escalation of the emergency classification level would be via IC SA3.
ONS Basis Reference(s):
: 1. UFSAR Section 7.5 Display Instrumentation
: 2. NEI 99-01 SU2
[Document No.]                                 Rev. 0                             Page 160 of 240

ATTACHMENT 1 EAL Bases Category:                S - System Malfunction Subcategory:              3 - Loss of Control Room Indications Initiating Condition:     UNPLANNED loss of Control Room indications for 15 minutes or longer with a significant transient in progress EAL:
SA3.1          Alert An UNPLANNED eveht results in the inability to monitor one or more Tablb S-2 parameters from within the Control Room for __15 mai. (Note 1)
AND Any significant transient is in progress, Table S-3 Note 1:  The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
Table S-2        Safety System Parameters
* Reactor power
* RCS level
* RCS pressure
* CETC temperature
* Level in at least one S/G
* EFW flow to at least one S/G Table S-3      Significant Transients
* Reactor trip
* Runback > 25% thermal power
* Electrical load rejection   > 25%
electrical load
                                     *ECCS actuation Mode Applicability:
I - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):
UNPLANNED - A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.
S[Document No.]                                      Rev. 0                            Page 1.61 of 240

ATTACHMENT 1 EAL Bases Basis:
SAFETY SYSTEM parameters listed in Table S-2 are monitored in the Control Room through a combination of hard control panel indicators as well as computer based information systems.
The SPDS serves as a redundant compensatory indicator which may be utilized in lieu of normal Control Room indicators (ref. 1).
Significant transients are listed in Table S-3 and include response to automatic or manually initiated functions such as reactor trips, runbacks involving greater than 25% thermal power change, electrical load rejections of greater than 25% full electrical load, reactor power cutbacks or ECCS (SI) injectio'n actuations.
This IC addresses the difficulty associated with monitoring rapidly changing plant conditions during a transient without the ability to obtain SAFETY SYSTEM parameters from within the Control Room. During this condition, the margin to a potential fission product barrier challenge is reduced. It thus represents a potential substantial degradation in the level of safety of the plant.
As used in this EAL, an "inability to monitor" means that values for one or more of the listed parameters cannot be determined from within the Control Room. This situation would require a loss of all of the Control Room sources for the given parameter(s). For example, the reactor power level cannot be determined from any analog, digital and recorder source within the Control Room.
An event involving a loss of plant indications, annunciators and/or display systems is evaluated in accordance with 10 CFR 50.72 (and associated guidance in NUREG-1 022) to determine if an NRC event report is required. The event would be reported if it significantly impaired the capability to perform emergency assessments. In particular, emergency assessments necessary to implement abnormal operating procedures, emergency operating procedures, and emergency plan implementing procedures addressing emergency classification, accident assessment, or protective action decision-making.
This EAL is focused on a selected subset of plant parameters associated with the key safety functions of reactivity control, core cooling and ROS heat removal. The loss of the ability to determine one or more of these parameters from within the Control Room is considered to be more significant than simply a reportable condition. In addition, if all indication sources for one or more of the listed parameters are lost, then the ability to determine the values of other SAFETY SYSTEM parameters may be impacted as well. For example, if the value for reactor vessel level cannot be determined from the indications and recorders on a main control board, the SPDS or the plant computer, the availability of other parameter values may be compromised as well.
Fifteen minutes was selected as a threshold to exclude transient or momentary losses of indication.
Escalation of the emergency classification level would be via ICs FS1 or IC RS1 ONS Basis Reference(s):
: 1. UFSAR Section 7.5 Display Instrumentation
: 2. NEI 99-01 SA2 S[Document No.]                                  Rev. 0                            Page 162 of 24.0

ATTACHMENT 1 EAL Bases Category:                S - System Malfunction Subcategory:             4 - RCS Activity Initiating Condition:    Reactor coolant activity greater than Technical Specification allowable limits EAL:
SU4.1         Unusual Event ROS activity > 50 pCi/gm Dose Equivalent 1-131 for > 48 hr continuous period OR RCS activity > 280 pCi/gm Dose Equivalent Xe-I133 for > 48 hr continuous period Mode Applicability:
I - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):
None Basis:
The specific iodine activity is limited to -<50 pCi/gm Dose Equivalent 1-131 for > 48 hr continuous period. The specific Xe-I133 activity is limited to -<280 pCi/gm Dose Equivalent Xe-133 for > 48 hr continuous period. Entry into Condition C of LCO 3.4.11 meets the intent of this EAL (ref 1).
This IC addresses a reactor coolant activity value that exceeds an allowable limit specified in Technical Specifications. This condition is a precursor to a more significant event and represents a potential degradation of the level of safety of the plant.
Escalation of the emergency classification level would be via ICs FA1 or the Recognition Category R ICs.
ONS Basis Reference(s):
: 1. ONS Technical Specifications LCO 3.4.11 RCS Specific. Activity
: 2. NEI 99-01 SU3 r[Document No.]                                  Rev. 0                         Page  163 of 2401

ATTACHMENT 1 EAL Bases Category:                  S - System Malfunction Suboategory:               5 - RCS Leakage Initiating Condition:      RCS leakage for 15 minutes or longer EAL:
SU5.1          Unusual Event RCS unidentified or pressure boundary leakage > 10 gpm for >- 15 mai.
OR RCS identified leakage >25 gpm for > 15 mai.
OR Leakage from the RCS to a location outside containment > 25 gpm for > 15 min.
(Note 1)
Note 1:   The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
Mode Applicability:
1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):
None Basis:
Manual or computer-based methods of performing an RCS inventory balance are normally used to determine RCS leakage (ref. 1).
Identified leakage includes (ref. 2):
* Leakage such as that from pump seals or valve packing (except reactor coolant pump (RCP) seal water injection or leakoff), that is captured and conducted to collection systems or a sump or collecting tank, or
* Leakage into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be pressure boundary leakage, or
* RCS leakage through a steam generator to the secondary system.
Unidentified leakage is all leakage (except RCP seal water injection or leakoff) that is not identified leakage (ref. 2).
Pressure Boundary leakage is leakage (except SG leakage) through a nonisolable fault in an RCS component body, pipe wall, or vessel wall (ref. 2).
Reactor coolant leakage outside of the containment that is not considered identified or unidentified leakage per Technical Specifications includes leakage via interfacing systems.
Escalation of this EAL to the Alert level is via Category F, Fission Product Barrier Degradation, EAL FAI.1.
S[Document No.]                                       Rev. 0                            Page 164 of 2401

ATTACHMENT I EAL Bases This IC addresses RCS leakage which may be a precursor to a more significant event. In this case, RCS leakage has been detected and operators, following applicable procedures, have been unable to promptly isolate the leak. This condition is considered to be a potential degradation of the level of safety of the plant.
The first and second EAL conditions are focused on a loss of mass from the RCS due to "unidentified leakage", "pressure boundary leakage" or "identified leakage" (as these leakage types are defined in the plant Technical Specifications). The third condition addresses an RCS mass loss caused by an UNISOLABLE leak through an interfacing system. These conditions thu~s apply to leakage into the containment, a secondary-side system (e.g., steam generator tube leakage) or a location outside of containment.
The leak rate values for each condition were selected because they are usually observable with normal Control Room indications. Lesser values ty~pically require time-consuming calculations to determine (e.g., a mass balance calculation). The first condition uses a lower value that reflects the greater significance of unidentified or pressure boundary leakage.
The release of mass from the RCS due to the as-designed/expected operation of a relief valve does not warrant an emergency classification. An emergency classification would be required if a mass loss is caused by a relief valve that is not functioning as designed/expected (e.g., a relief valve sticks open and the line flow cannot be isolated).
The 15-minute threshold duration allows sufficient time for prompt operator actions to isolate the leakage, if possible.                '
Escalation of the emergency classification level would be via ICs of Recognition Category R or F.
ONS Basis Reference(s):
: 1. PT/I1,2,3/A/0600/01 0 Reactor Coolant Leakage
: 2. ONS Technical Specifications Section 1.1 Definitions
: 2. NEI 99-01 SU4 S[Document No.]                                 Rev. 0                           Page 165 of 240

ATTACHMENT 1 EAL Bases Category:                     S - System Malfunction Subcategory:                  6 - RPS Failure Initiating Condition:         Automatic or manual trip fails to shut down the reactor EAL:
SU6.1            Unusual Event An automatic trip did not shut down the reactor as indicated by reactor power > 5% after any RPS setpoint is exceeded AND A subsequent automatic trip or the manual trip pushbutton is successful in shutting down the reactor as indicated by reactor power          < 5% (Note 8)
Note 8:   A manual trip action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron injection strategies.
Mode Applicability:
1 - Power Operation Definition(s):
None Basis:
The first condition of this EAL identifies the need to cease critical reactor operations by actuation of the automatic Reactor Protective System (RPS) trip function. A reactor trip is automatically initiated by the RPS when certain continuously monitored parameters exceed predetermined setpoints (ref. 1).
Following a successful reactor trip, rapid insertion of the control rods occurs. Nuclear power promptly drops to a fraction of the original power level and then decays to a level several decades less with a negative startup rate. The reactor power drop continues until reactor power reaches the point at which the influence of source neutrons on reactor power starts to be observable. A predictable post-trip response from an automatic reactor trip signal should therefore consist of a prompt drop in reactor power as sensed by the nuclear instrumentation and a lowering of power into the source range. For the purpose of emergency classification a successful trip has occurred when there is sufficient rod insertion from the trip of RPS to bring the reactor power below the Power Operation Mode threshold of 5% (ref. 2).
5% rated power is the Power Operation mode threshold. Below 5%, plant response will be similar to that observed during a normal shutdown. Nuclear instrumentation can be used to determine if reactor power is greater than or equal to 5 % power (ref. 1, 2).
For the purposes of emergency classification, a successful manual trip action is that which can be quickly performed from the reactor control console (actuation of the manual trip pushbutton). There is a separate set of switch contacts in series with the output of each reactor trip component. All switch contacts are actuated through a mechanical linkage from a single pushbutton. Reactor shutdown achieved by use of other trip actions such as opening supply S[Document No.]                                       Rev. 0                                 Page 166 of 240I

ATTACHMENT 1 EAL Bases breakers, emergency boration, or manually driving control rods) do not constitute a successful manual trip (ref. 3).
Following any automatic RPS trip signal, insertion of redundant manual trip signals are performed to back up the automatic RPS trip function and ensure reactor shutdown is achieved. Even if the first subsequent manual trip signal inserts all control rods to the full-in position immediately after the initial failure of the automatic trip, the lowest level of classification that must be declared is an Unusual Event.
In the event that the operator identifies a reactor trip is imminent and initiates a successful manual reactor trip Ibefore the automatic RPS trip setpoint is reached, noldeclaration is
*required. The successful manual trip of the reactor before it reaches its automatic trip setpoint or reactor trip signals caused by instrumentation channel failures do not lead to a potential fission product barrier loss. However, if subsequent manual reactor trip actions fail to reduce reactor power belov* 5%, the event escalates to the Alert under EAL SA6.1.
If by procedure, operator actions include the initiation of an immediate manual trip following receipt of an automatic trip signal and there are no clear indications that the automatic trip failed (such as a time delay following indications that a trip setpoint was exceeded), it may be difficult to determine if the reactor was shut down because of automatic trip or manual actions.
If a subsequent review of the trip actuation indications reveals that the automatic trip did not cause the reactor to be shut down, consideration should be given to evaluating the fuel for potential damage, and the reporting requirements of 10CFR50.72 should be considered for the transient event.
This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor trip that results in a reactor shutdown, and either a subsequent operator manual action taken at the reactor control consoles or an automatic trip is successful in shutting down the reactor.
This event is a precursor to a more significant condition and thus represents a potential degradation of the level of safety of the plant.
Following the failure on an automatic reactor trip, operators will promptly initiate manual actions at the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor trip). Ifthese manual actions are successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plant's decay heat removal systems.
If an initial manual reactor trip is unsuccessful, operators will promptly take manual action at another location(s) on the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor trip) using a different switch). Depending upon several factors, the initial or subsequent effort to manually trip the reactor, or a concurrent plant condition, may lead to the generation of an automatic reactor trip signal. If a subsequent manual or automatic trip is successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plant's decay heat removal systems.
A manual action at the reactor control Consoles is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor trip). This action does not include manually driving in control rods or implementation of boron injection strategies. Actions taken at back-panels or other locations within the Control Room, or any location outside the Control Room, are not considered to be "at the reactor control consoles".
r[Document No.]                                     Rev. 0                          Page 167 of 240

ATTACHMENT 1 EAL Bases The plant response to the failure of an automatic or manual reactor trip will vary based upon several factors including the reactor power level prior to the event, availability of the condenser, performance of mitigation equipment and actions, other concurrent plant conditions, etc. If subsequent operator manual actions taken at the reactor control consoles are also unsuccessful in shutting down the reactor, then the emergency classification level will escalate to an Alert via IC SA6. Depending upon the plant response, escalation is also possible via IC FAI. Absent the plant conditions needed to meet either IC SA6 or FA1, an Unusual Event declaration is appropriate for this event.
Should a reactor trip signal b~e generated as a result of plant work (e.g., RPS setp~oint testing),
the following classification guidance should be applied.
* If the signal causes a plant transient that should have included an automatic reactor trip and the RPS fails to automatically shutdown the reactor, then this IC and the EALs are applicable, and shoul~d be evaluated.
* If the signal does not cause a plant transient and the trip failure is determined through other means (e.g., assessment of test results), then this IC and the EALs are not applicable and no classification is warranted.
ONS Basis Reference(s):
: 1. ONS Technical Specifications Section 3.3.1 Reactor Protective System (RPS)
Instrumentation   - Operating
: 2. ONS Technical Specifications Table 1.1-1 Modes
: 3. UFSAR Section 7.2.3.7 Manual Trip
: 4. OSC-02820 Emergency Procedure Setpoints, Setpoint No. 7.1.7.5
: 5. NEI 99-01 SU5 r[Document No.]                                  Rev. 0                             Page 168 of 240

ATTACHMENT 1 EAL Bases
*Category:                     S - System Malfunction Subcategory:                  6 - RPS Failure Initiating Condition:         Automatic or manual trip fails to shut down the reactor EAL:
SU6.2            Unusual Event A manual trip did not shut down the reactor as indicated by reactor power > 5% after any manual trip action was initiated            I AND A subsequent automatic trip or the manual trip pushbutton is successful in shutting down the reactor as indicated by reactor power          < 5% (Note 8)
Note 8:   A manual trip action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving-in control rods or implementation of boron injection strategies.
Mode Applicability:
1 - Power Operation Definition(s):
None Basis:
This EAL addresses a failure of a manually initiated trip in the absence of having exceeded an automatic RPS trip setpoint and a subsequent automatic or manual trip is successful in shutting down the reactor (ref. 1).
Following a successful reactor trip, rapid insertion of the control rods occurs. Nuclear power promptly drops to a fraction of the original power level and then decays to a level several decades less with a negative startup rate. The reactor power drop continues until reactor power reaches the point at which the influence of source neutrons on reactor power starts to be observable. A predictable post-trip response from an a~utomatic reactor trip signal should therefore consist of a prompt drop in reactor power as sensed by the nuclear instrumentation and a lowering of power into the source range. For the purpose of emergency classification a successful trip has occurred when there is sufficient rod insertion from the manual trip to bring the reactor power below the Power Operation Mode threshold level of 5% (ref. 2).
5% rated power is the Power Operation mode threshold. Below 5%, plant response will be similar to that observed during a normal shutdown. Nuclear instrumentation can be used to determine if reactor power is greater than or equal to 5% power (ref. 1, 2).
For the purposes of emergency classification, a successful manual trip action is that which can be quickly performed from the reactor control console (actuation of the manual trip pushbutton). There is a separate set of switch contacts in series with the output of each reactor trip component. All switch contacts are actuated through a mechanical linkage from a single pushbutton. Reactor shutdown achieved by use of other trip actions such as opening supply r[Document No.]                                        Rev. 0                                 Page 169 of 240

ATTACHMENT 1 EAL Bases breakers, emergency boration, or manually driving control rods) do not constitute a successful manual trip (ref. 3).
Following any automatic RPS trip signal, insertion of redundant manual trip signals are performed to back up the automatic RPS trip function and ensure reactor shutdown is achieved. Even if a subsequent automatic trip signal or the first subsequent manual trip signal inserts all control rods to the full-in position immediately after the initial failure of the manual trip, the lowest level of classification that must be declared is an Unusual Event.
If both subsequent automatic and subsequent manual reactor trip actions in the Control Room fail to reduce reactor power     < 5% follbwing a failure of an initial manual trip, the event      I escalates to an Alert under EAL SA6.1.
This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor trip that results in a reactor shutdown, and either a subsequent operator manual action taken at the reactor control consoles or an automatic trip is successful in shutting down the reactor.
This event is a precursor to a more significant condition and thus represents a potential degradation of the level of safety of the plant.
Following the failure on 'an automatic reactor trip, operators will promptly initiate manual actions at the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor trip). If these manual actions are successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plant's decay heat removal systems.
If an initial manual reactor trip is unsuccessful, operators will promptly take manual action at another location(s) on the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor trip) using a different switch). Depending upon several factors, the initial or subsequent effort to manually the reactor, or a concurrent plant condition, may lead to the generation of an automatic reactor trip signal. If a subsequent manual or automatic trip is successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plant's decay heat removal systems.
A manual action at the reactor control consoles is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor trip). This action does not include manually driving in control rods or implementation of boron inject~ion strategies. Actions taken at back-panels or other locations within the Control Room, or any location outside the Control Room, are not considered to be "at the reactor control consoles".
The plant response to the failure of an automatic or manual reactor trip will vary based upon several factors including the reactor power level prior to the event, availability of the condenser, performance of mitigation equipment and actions, other conicurrent plant conditions, etc. If subsequent operator manual actions taken at the reactor control consoles are also unsuccessful in shutting down the reactor, then the emergency classification level will escalate to an Alert via IC SA6. Depending upon the plant response, escalation is also possible via IC FAI. Absent the plant conditions needed to meet either IC SA6 or FA1, an Unusual Event declaration is appropriate for this event.
Should a reactor trip signal be generated as a result of plant work (e.g., RPS setpoint testing),
the following classification guidance should be applied.
I[Document No.]                          I         Rev. 0           IPage                    170 of 240

ATTACHMENT 1 EAL Bases
* Ifthe signal causes a plant transient that should have included an automatic reactor trip and the RPS fails to automatically shutdown the reactor, then this IC and the EALs are applicable, and should be evaluated.
* If the signal does not cause a plant transient and the trip failure is determined through other means (e.g., assessment of test results), then this IC and the EALs are not applicable and no classification is warranted.
ONS Basis Reference(s):
: 1. ONS Technical Specifications Section 3.3.11 Reactor Protective System (RPS)
Instrumentation - Operating
: 2. ONS Technical Specifications Table 1.1-1
: 3. UFSAR Section 7.2.3.7 Manual Trip
: 4. OSC-02820 Emergency Procedure Setpoints, Setpoint No. 7.1.7.5
: 5. NEI 99-01 SU5 I[Document No.]                       r        Rev. 0                             Page 171 of 2401

ATTACHMENT 1 EAL Bases Category:                     S - System Malfunction Subcategory:                  6 - RPS Failure Initiating Condition:         Automatic or manual trip fails to shut down the reactor and subsequent manual actions taken at the reactor control consoles are not successful in shutting down the reactor EAL:
SA6.1            Alert An automatic or manual trip fails to shut down the reactor as indicated by reactor power
  ->5%
AND Manual trip pushbutton is not successful in shutting down the reactor as indicated by reactor power --5% (Note 8)
Note 8:   Amanual trip action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving incontrol rods or implementation of boron injection strategies.
Mode Applicability:
1 - Power Operation Definition(s):
None Basis:
This EAL addresses any automatic or manual reactor trip signal that fails to shut down the reactor followed by a subsequent manual trip that fails to shut down the reactor to an extent the reactor is producing significant power (ref. 1).
Following a successful reactor trip, rapid insertion of the control rods occurs. Nuclear power promptly drops to a fraction of the original power level and then decays to a level several decades less with a negative startup rate. The reactor power drop continues until reactor power reaches the point at which the influence of source neutrons on reactor power starts to be observable. A predictable post-trip response from an automatic reactor trip signal should therefore consist of a prompt drop in reactor power as sensed by the nuclear instrumentation and a lowering of power into the source range. For the purpose of emergency classification a successful trip has occurred when there is sufficient rod insertion from the manual trip to bring the reactor power below 5% (ref. 2).
5% rated power is the Power Operation mode threshold. Below 5%, plant response will be similar to that observed during a normal shutdown. Nuclear instrumentation can be used to determine if reactor power is greater than or equal to 5% power (1, 2).
For the purposes of emergency classification, a successful manual trip action is that which can be quickly performed from the reactor control console (actuation of the manual trip pushbutton). There is a separate set of switch contacts in series with the output of each reactor trip component. All switch contacts are actuated through a mechanical linkage from a single I[Document No.]                                       Rev. 0              IPage                      172 of 240I

ATTACHMENT 1 EAL Bases pushbutton. Reactor shutdown achieved by use of other trip actions such as opening supply breakers, emergency boration, or manually driving control rods) do not constitute a successful manual trip (ref. 3).
Escalation of this event to a Site Area Emergency would be under EAL SS6.1 or Emergency Coordinator judgment.
This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor trip that results in a reactor shutdown, and subsequent operator manual actions taken at the reactor control consoles to shutdown the reactor are also unsuccessful. This condition represents an actual or potential substantial degradation of the' level of safety of the plant. An emergency declaration is r~equired even if the reactor is subsequently shutdown by an action taken away from the reactor control consoles since this event entails a significant failure of the RPS.
A manual action at the reactor control console is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor trip). This action does not include manually driving in control rods or implementation of boron injection strategies. If this action(s) is unsuccessful, operators would immediately pursue additional manual actions at locations away from the reactor control console (e.g., locally opening breakers). Actions taken at back panels or other locations within the Control Room, or any location outside the Control Room, are not considered to be "at the reactor control console".
The plant response to the failure of an automatic or manual reactor trip will vary based upon several factors including the reactor power level prior to the event, availability of the condenser-, performance of mitigation equipment and actions, other concurrent plant conditions, etc. If the failure to shut down the reactor is prolonged enough to cause a challenge to the core cooling or RCS heat removal safety functions, the emergency classification level will escalate to a Site Area Emergency via IC SS6. Depending upon plant responses and symptoms, escalation is also possible via IC ESI. Absent the plant conditions needed to meet either IC SS6 or FS1, an Alert declaration is appropriate for this event.
It is recognized that plant responses or symptoms may also require an Alert declaration in accordance with the Recognition Category F ICs; however, this IC and EAL are included to ensure a timely emergency declaration.
A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.
ONS Basis Reference(s):
: 1. ONS Technical Specifications Section 3.3.1 Reactor Protective System (RPS)
Instrumentation   - Operating
: 2. ONS Technical Specifications Table 1.1-1
: 3. UFSAR Section 7.2.3.7 Manual Trip
: 4. OSC-02820 Emergency Procedure Setpoints, Setpoint No. 7.1.7.5
: 5. NEI 99-01 SA5 S[Document No.]                                  Rev. 0                            Page 173 of 240

ATTACHMENT I EAL Bases Category:       S - System Malfunction Subcategory:              6 - RPS Failure Initiating Condition:     Inability to shut down the reactor causing a challenge to core cooling or RCS heat removal EAL:
SS6.1           Site Area Emergency An automatic or manual trip fails to shut down the reactor as indicated by reactor power
-&#x17d;5%
AND All actions to shut down the reactor are not successful as indicated by reactor power
->5%
AND EITHER:
* CETCs >1200&deg;F on ICCM
* RCS subcooling < 0&deg;F Mode Applicability:
1 - Power Operation Definition(s):
None Basis:
This EAL addresses the following:
      *Any automatic reactor trip signal (ref. 1) followed by a manual trip that fails to shut down the reactor to an extent the reactor is producing energy in excess of the heat load for which the safety systems were designed (ref. 5), and
      *Indications that either core cooling is extremely challenged or heat removal is extremely challenged.
The combination of failure of both front line and backup protection systems to function in response to a plant transient, along with the continued production of heat, poses a direct threat to the Fuel Clad and RCS Barriers.
Reactor shutdown achieved by use of other trip actions such as opening supply breakers, emergency boration, or manually driving control rods are also credited as a successful manual trip provided reactor power can be reduced below 5% before indications of an extreme challenge to either core cooling or heat removal exist (ref. 2, 3).
5% rated power is the Power Operation mode threshold. Below 5%, plant response will be similar to that observed during a normal shutdown. Nuclear instrumentation can be used to determine if reactor power is greater than or equal to 5% power.
[Dcmn o]Rev.                                0                          Page 174 of 240

ATTACHMENT 1 EAL Bases Indication of continuing core cooling degradation is manifested by CETCs are reading greater than 1200&deg;F. This setpoint is used as an indication of an extreme ICC condition and entry into the Oconee Severe Accident Guidelines (OSAG) is initiated for further mitigative actions (ref.
4).
Indication of inability to adequately remove heat from the RCS is manifested by subcooling less than 0&deg;F (ref. 6).
This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor trip that results in a reactor shutdown, all subsequent operator actions to manually shutdown the reactor are unsuccessful, and continued power generation is challenging th* capability to adequately remove heat from the core and/or the RCS. This condition will lead to fuel damage if additional mitigation actions are unsuccessful and thus warrants the declaration of a Site Area Emergency.
In some instances, the emergency classification resulting from this IC/EAL may be higher than that resulting from an assessment of the plant responses and symptoms against the Recognition Category F ICs/EALs. This is appropriate in that the Recognition Category F ICs/EALs do not address the additional threat posed by a failure to shut down the reactor. The inclusion of this IC and EAL ensures the timely declaration of a Site Area Emergency in response to prolonged failure to shutdown the reactor.
A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.
Escalation of the emergency classification level would be via IC RG1 or FGI.
ONS Basis Reference(s):
: 1. ONS Technical Specifications Section 3.3.1 Reactor Protective System (RPS)
Instrumentation   - Operating
: 2. ONS Technical Specifications Table 1.1-1
: 3. UFSAR Section 7.2.3.7 Manual Trip
: 4. OSC-02820 Emergency Procedure Setpoints, Setpoint No. 7.1.1.7
: 5. OSC-02820 Emergency Procedure Setpoints, Setpoint No. 7.1.7.5
: 6. EP/1 ,2,3/A/1800/001 Loss of Subcooling Margin
: 7. NEI 99-01 SS5 S[Document No.]                                 Rev. 0                         Page 175 of 240

ATTACHMENT 1 EAL Bases Category:                S - System Malfunction Subcategory:              7 - Loss of Communications Initiating Condition:     Loss of all onsite or offsite communications capabilities EAL:
SU7.1          Unusual Event Loss of all Table S-4 onsite communication methods OR Loss of all Table S-4 offsite communication methods OR Loss of all Table S-4 NRC communication methods Mode Applicability:
1 - Power Operation, 2 - Startup, 3  - Hot Standby, 4 - Hot Shutdown Definition(s):
None S[Document No.]                                 Rev. 0                           Page 176 of 240

ATTACHMENT 1 EAL Bases Basis:
Onsite, offsite and NRC communications include one or more of the systems listed in Table S-4 (ref. 1).
: 1. Commercial phone service The Commercial phone service does not go through the site telephone system.
: 2. O)NS site phone system The site phone system is generator and ibattery backed with:
* Fiber-Optic to Charlotte GO (65 lines)
* Telephone line to Easley (6 circuits)
* Anderson (4 lines)
* Six Mile (4 lines)
* Site Telephone System: Inward and outward direct dial available from the Control Room, TSC, and OSC
: 3. EOF phone system The emergency communications systems at the Charlotte EOF are designed to ensure the reliable, timely flow of information between all parties having an emergency response role.
: 4. Public Address (Paging) system The paging system provides paging and party line communications between stations located throughout the plant. Inside and outside type wall and desk-mounted stations are used to communicate between roaming personnel and fixed work locations. Plant-wide instructions are issued using the paging feature.
: 5. Onsite radio system The onsite radio system receives emergency backup power from Keowee Hydro Units supporting communications with: Control Room 1&2, 3, Fire Brigade, Chemistry, Safety, Radiation Protection, Maintenance, Medical Emergency Response Team, and Hazardous Materials Response Team.
: 6. DEMNET DEMNET is the primary means of offsite communication. This circuit allows intercommunication among the EOF, TSC, control room, counties, and states. DEMNET operates as an internet based (VolP) communications system with a satellite back-up. Should the internet transfer rate become slow or unavailable, the DEMNET will automatically transfer to satellite mode.
: 7. Offsite radio system The offsite radio system is battery backed supporting communications with: Control Room Units 1&2, TSC, Field Monitoring Teams, EOF, counties and State of South Carolina.
.[Document No.]                                Rev. 0          rPage                177 of 240

ATTACHMENT 1 EAL Bases
: 8. NRC Emergency Telephone System (ETS)
The NRC uses a Duke Energy dedicated telephone line which allows direct telephone communications from the plant to NRC regional and national offices. The Duke Energy communications line provides a link independent of the local public telephone network.
Telephones connected to this network are located in the Oconee Control Rooms, TSC, and EOF and can be used to establish NRC Emergency Notification System (ENS) and Health Physics Network (HPN) capability.
: 9. Satellite Phone\
Satellite Phones can be used for external communications This EAL is the hot condition equivalent of the cold condition EAL CU5.1.
This IC addresses a significant loss of on-site or offsite communications capabilities. While not a direct challenge to plant or personnel safety, this event warrants prompt notifications to OROs and the NRC.
This IC should be assessed only when extraordinary means are being utilized to make communications possible (e~g., use of non-plant, privately owned equipment, relaying of on-site information via individuals or multiple radio transmission points, individuals being sent to offsite locations, etc.).
The first EAL condition addresses a total loss of the communications methods used in support of routine plant operations.
The second EAL condition addresses a total loss of the communications methods used to notify all OROs of an emergency declaration. The OROs referred to here are the State EOC and FEO, Pickens County LEC and EOC, and Oconee County LEC and EOC.
The third EAL addresses a total loss of the communications methods used to notify the NRC of an emergency declaration.
ONS Basis Reference(s):
: 1. ONS Emergency Plan, Section 7.2 Communications Systems
: 2. NEI 99-01 SU6 S[Document No.]                                 Rev. 0           *Page                  178 of 240I

ATTACHMENT 1 EAL Bases Category:                  S - System Malfunction Subcategory:               8 - Containment Failure Initiating Condition:      Failure to isolate containment or loss of containment pressure control.
EAL:
SU8.1          Unusual Event Any penetration is not closed within 15 min. of a VALID ES actuation signal OR Containment pressure > 10 psig with < one full train of containment heat removal system (1 RBS with    >  700 gpm spray flow OR 2 RBCUs) operating per design for
 > 15 min. (Note 1)
Note 1:   The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
Mode Applicability:
1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):
VALID - An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment.
Basis:
Reactor Building isolations are initiated by Engineered Safeguards Actuation Channels 5 and 6 in response to a high reactor building pressure signal (3.0 psig) (ref. 1, 2, 4).
Two engineered safeguards Systems, the Reactor Building Spray System and the Reactor Building Cooling System, are provided to remove heat from the containment atmosphere following an accident. Both the Reactor Building Spray System and the Reactor Building Cooling System, with either at full capacity, are individually capable of maintaining the containment pressure below the design limit following a LOCA or MSLB. (ref. 1, 3)
The Reactor Building Spray (RBS) System consists of two separate trains of equal capacity.
Spray flow greater or equal to 700 gpm satisfies the spray flow design requirement. The Reactor Building pressure setpoint (10 psig) is the pressure at which the Reactor Building Spray equipment should actuate and begin performing its function (ref. 1, 2, 3, 5).
Each of three Reactor Building Cooling Units (RBCUs) consists of a fan, cooling coils, and the required distribution duct work. The Reactor Building atmosphere is circulated past cooling coils by fans and returned to the building. Cooling water for the cooling units is supplied by the Low Pressure Service Water System. The Reactor Building Cooling System provides the design heat removal capacity with two of three coolers operating (ref. 1).
I[Document No.].*                                     Rev. 0                    .       Page 179 of 240I

ATTACHMENT 1 EAL Bases This EAL addresses a failure of one or more containment penetrations to automatically isolate (close) when required by an actuation signal. It also addresses an event that results in high containment pressure with a concurrent failure of containment pressure control systems.
Absent challenges to another fission product barrier, either condition represents potential degradation of the level of safety of the plant.
For the first condition, the containment isolation signal must be generated as the result on an off-normal/accident condition (e.g., a safety injection or high containment pressure); a failure resulting from testing or maintenance does not warrant classification. The determination of containment and penetration status - isolated or not isolated i- should be made in accordance with the appropriate criteria contained in the plant APs and EOPs. The 15-minute criterion is included to allow operators time to manually isolate the required penetrations, if possible.
The second condition addresses a condition where containment pressure is greater than the setpoint at which containment energy (heat) removal systems are designed to automatically actuate, and less than one full train of equipment is capable of operating per design. The 15-minute criterion is included to allow operators time to manually start equipment that may not have automatically started, if possible. The inability to start the required equipment indicates that containment heat removal/depressurization systems (e.g., containment sprays) are either lost or performing in a degraded manner.
This event would escalate to a Site Area Emergency in accordance with IC FS1 if there were a concurrent loss or potential loss of either the Fuel Clad or RCS fission product barriers.
ONS Basis Reference(s):
: 1. UFSAR Section 6.2.2 Containment Heat Removal Systems
: 2. UFSAR Table 7-2 Engineered Safeguards Actuation Conditions
: 3. UFSAR Table 6-25 Minimum Acceptable Combinations of Containment Heat Removal Equipment Performance
: 4. OSC-02820 Emergency Procedure Setpoints, Setpoint No. 7.1 .1.1
: 5. OSC-02820 Emergency Procedure Setpoints, Setpoint No. 7.4.1.2
: 6. NEI 99-01 SU7
[Document No.].                                Rev. 0                           Page 180 of 240

ATTACHMENT 1 EAL Bases Category:                 S - System Malfunction Subcategory:              9 - Hazardous Event Affecting Safety Systems Initiating Condition:     Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode EAL:
SA9.1          Alert The occurrence bf any Table S-5 hazardous event AND EITHER:
       *Event damage has caused indications of degraded performance in at least one train of a SAFETY SYSTEM needed for the current operating mode*
* The event has caused VISIBLE DAMAGE to a SAFETY SYSTEM component or structure needed for the current operating mode
[         Table S-5     Hazardous Events
* Seismic event (earthquake)
* Internal or external FLOODING event
* High winds or tornado strike
* FIRE
* EXPLOSION
* Other events with similar hazard characteristics as determined by the Shift Manager Mode Applicability:
1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):
EXPLOSION - A rapid, violent and catastrophic failure of a piece of equipment due to combustion, chemical reaction or overpressurization. A release of steam (from high energy lines or components) or an electrical component failure (caused by short circuits, grounding, arcing, etc.) should not automatically be considered an explosion. Such events require a post-event inspection to determine if the attributes of an explosion are present.
FIRE - Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred but is NOT required if large quantities of smoke and heat are observed.
FLOODING - A condition where water is entering a room or area faster than installed equipment is capable of removal, resulting in a rise of water level within the room or area.
SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems S[Document No.]                                 Rev. 0                          Page 181 of 240

ATTACHMENT 1 EAL Bases classified as safety-related (as defined in 10OCFR50.2):
Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:
(1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.
VISIBLE DAMAGE - Damage to a component or structure that is readily obse rvable without measurements, testing, or analysis. The visual impact of the damage is sufficient to cause concern regarding the operability or reliability of the affected component or structure.
Basis:
* The significance of seismic events are discussed under EAL HU2.1 (ref. 1).
* Internal FLOODING may be caused by events such as component failures, equipment misalignment, or outage activity mishaps.
   *External flooding at ONS is highly unlikely since the probable maximum flood (PMF) would be contained by the Keowee Reservoir. Plant grade elevation is 796.0 ft MSL. The minimum external access elevation for the Auxiliary, Turbine, and Service Buildings is 796.5 ft MSL which provides a 6 inch water sill. (ref. 2)
   *High winds in excess of design (95 mph) or tornado strikes can cause significant structural damage (ref. 3).
   *Areas containing functions and systems required for safe shutdown of the plant are identified by fire area (ref. 4, 5).
   *An explosion that degrades the performance of a SAFETY SYSTEM train or visibly damages a SAFETY SYSTEM component or structure would be classified under this EAL.
This IC addresses a hazardous event that causes damage to a SAFETY SYSTEM, or a structure containing SAFETY SYSTEM components, needed for the current operating mode.
This condition significantly reduces the margin to a loss or potential loss of a fission product barrier, and therefore represents an actual or potential substantial degradation of the level of safety of the plant.
The first condition addresses damage to a SAFETY SYSTEM train that is in service/operation since indications for it will be readily available. The indications of degraded performance should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train.
The second condition addresses damage to a SAFETY SYSTEM component that is not in service/operation or readily apparent through indications alone, or to a structure containing SAFETY SYSTEM components. Operators will make this determination based on the totality of available event and damage report information. This is intended to be a brief assessment not requiring lengthy analysis or quantification of the damage.
Escala'tion of the emergency classification level would be via IC FS1 or RSI.
I[Document No.]                                   Rev. 0                         Page 182 .of 240I

ATTACHMENT I EAL Bases ONS Basis Reference(s):
I. AP/O/A/1 700/005 Earthquake
: 2. UFSAR Section 3.4.1.1 Flood Protection Measures for Seismic Class 1 Structures
: 3. UFSAR Section 3.3.1.1 Design Wind Velocity
: 4. OSS-0254.00-00-4008 Design Bases Specification for Fire Protection
: 5. AP/1,2,3/A/1700/050 Challenging Plant Fire
: 6. NEI 99-01 SA9
[Document No.]                             Rev. 0                            Page 183 of 240

ATTACHMENT I EAL Bases Category F - Fission Product Barrier Degradation EAL Group: Hot Conditions (RCS temperature        > 200&deg;F); EALs in this category are applicable only in one or more hot operating modes.
EALs in this category represent threats to the defense in depth design concept that precludes the release of highly radioactive fission products to the environment. This concept relies on multiple physical barriers any one of which, if maintained intact, precludes the release of significant amounts of radioactive fission products to the environment. The primary fission product barriers are:
A. Fuel Clad (FC): The Fuel Clad Barrier consists of the cladding material that contains the fuel pellets.
B. Reactor Coolant System (RCS): The RCS Barrier includes the RCS primary side and its connections up to and including the pressurizer safety and relief valves, and other connections up to and including the primary isolation valves.
C. Containment (CMT): The Containment (Reactor Building) Barrier includes the Reactor Building and connections up to and including the outermost containment isolation valves. This barrier also includes the main steam, feedwater, and blowdown line extensions outside the Reactor Building up to and including the outermost secondary side isolation valve. Containment Barrier thresholds are used as criteria for escalation of the ECL from Alert to a Site Area Emergency or a General Emergency.
The EALs in this category require evaluation of the loss and potential loss thresholds listed in the fission product barrier matrix of Table F-I (Attachment 2). "Loss" and "Potential Loss" signify the relative damage and threat of damage to the barrier. "Loss" means the barrier no longer assures containment of radioactive materials. "Potential Loss" means integrity of the barrier is threatened and could be lost if conditions continue to degrade. The number of barriers that are lost or potentially lost and the following criteria determine the appropriate emergency classification level:
Alert:
Any loss or any potential loss of either Fuel Clad or RCS Barrier Site Area Emerqency:
Loss or potential loss of any two barriers General Emergqency:
Loss of any two barriersand loss or potential loss of third barrier The logic used for emergency classification based on fission product barrier monitoring should reflect the following considerations:                                                           .
* The Fuel Clad Barrier and the RCS Barrier are weighted more heavily than the Containment Barrier.
* Unusual Event ICs associated with RCS and Fuel Clad Barriers are addressed under System Malfunction ICs.
* For accident conditions involving a radiological release, evaluation of the fission product barrier thresholds will need to be performed in conjunction with dose assessments to l/[Document      No.]                               Rev. 0.                         Page 184 of 240 I

ATTACHMENT 1 EAL Bases ensure correct and timely escalation of the emergency classification. For example, an evaluation of the fission product barrier thresholds may result in a Site Area Emergency classification while a dose assessment may indicate that an EAL for General Emergency IC RG1 has been exceeded.
     *The fission product barrier thresholds specified within a scheme reflect plant-specific ONS design and operating characteristics.
     *As used in this category, the term RCS leakage encompasses not just those types defined in Technical Specifications but also includes the loss of RCS mass to any S location- inside the primary containment, ar* interfacing system, or outside of the primary containment. The release of liquid or steam mass from the RCS due to the as designed/expected operation of a relief valve is not considered to be RCS leakage.
     *At  the Site Area Emergency level, EAL users should maintain cognizance of how far present conditions are from meeting a threshold that would require a General Emergency declaration. For example, if the Fuel Clad and RCS fission product barriers were both lost, then there should be frequent assessments of containment radioactive inventory and integrity. Alternatively, if both the Fuel Clad and RCS fission product barriers were potentially lost, the Emergency Coordinator would have more assurance that there was no immediate need to escalate to a General Emergency.
I[Document No.]                                Rev. 0                           Page 185 of 240I

ATTACHMENT I EAL Bases Category:                  Fission Product Barrier Degradation Su bcategory:              N/A Initiating Condition:     Any loss or any potential loss of either Fuel Clad or RCS barrier EAL:
FA1.1           Alert Any loss or any potential loss of either Fuel Clad or RCS barrier (Table F-I)
Mode Applicability:
1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):
None Basis:
Fuel Clad, RCS and Containment comprise the fission product barriers. Table F-I (Attachment
: 2) lists the fission product barrier thresholds, bases and references.
At the Alert classification level, Fuel Clad and RCS barriers are weighted more heavily than the Containment barrier. Unlike the Containment barrier, loss or potential loss of either the Fuel Clad or RCS barrier may result in the relocation of radioactive materials or degradation of core cooling capability. Note that the loss or potential loss of Containment barrier in combination with loss or potential loss of either Fuel Clad or RCS barrier results in declaration of a Site Area Emergency under EAL FSI.1 ONS Basis Reference(s):
I. NEI 99-01 FA1 I[Document No;]                                  Rev. 0         IPage                   186of2401

ATTACH MENT 1 EAL Bases Category:                  Fission Product Barrier Degradation Subcategory:             .,N/A Initiating Condition:      Loss or potential loss of any two barriers EAL:
FSI.1           Site Area Emergency Loss or potential loss of any two barriers (Table F-i)
Mode Applicability:
1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):
None Basis:
Fuel Clad, RCS and Containment comprise the fission product barriers. Table F-I (Attachment
: 2) lists the fission product barrier thresholds, bases and references.
At the Site Area Emergency classification level, each barrier is weighted equally. A Site Area Emergency is therefore appropriate for any combination of the following conditions:
* One barrier loss and a second barrier loss (i.e., loss - loss)
* One barrier loss and a second barrier potential loss (i.e., loss - potential loss)
* One barrier potential loss and a second barrier potential loss (i.e., potential loss -
potential loss)
At the Site Area Emergency classification level, the ability to dynamically assess the proximity of present conditions with respect to the threshold for a General Emergency is important. For example, the existence of Fuel Clad and RCS Barrier loss thresholds in addition to offsite dose assessments would require continual assessments of radioactive inventory and Containment integrity in anticipation of reaching a General Emergency classification. Alternatively, if both Fuel Clad and RCS potential loss thresholds existed, the Emergency Coordinator would have greater assurance that escalation to a General Emergency is less imminent.
ONS Basis Reference(s):
: 1. NE199-O1 FSI I[Document No.]                                  Rev. 0                          Page 187 of 240I

ATTACHMENT 1 EAL Bases Category:                 Fission Product Barrier Degradation Subcategory:              N/A Initiating Condition:     Loss of any two barriers and loss or potential loss of third barrier EAL:
FGI.1            General Emergency Loss of any two barriers AND Loss or potential loss of third barrier (Table F-I)
Mode Applicability:
I - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):
None Basis:
Fuel Clad, RCS and Containment comprise the fission product barriers. Table F-I (Attachment
: 2) lists the fission product barrier thresholds, bases and references.
At the General Emergency classification level each barrier is weighted equally. A General Emergency is therefore appropriate for any combination of the following conditions:
* Loss of Fuel Clad, RCS and Containment Barriers
* Loss of Fuel Clad and RCS Barriers with potential loss of Containment Barrier
* Loss of RCS and Containment Barriers with potential loss of Fuel Clad Barrier
* Loss of Fuel Clad and Containment Barriers with potential loss of RCS Barrier ONS Basis Reference(s):
I. NEI 99-01 FG1
[Document No.]                                  Rev. 0                          Page 188 of 240

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Introduction Table F-I lists the threshold conditions that define the Loss and Potential Loss of the three fission product barriers (Fuel Clad, Reactor Coolant System, and Containment). The table is structured so that each of the three barriers occupies adjacent columns. Each fission product barrier column is further divided into two columns; one for Loss thresholds and one for Potential Loss thresholds.
The first column of the table (to the left of the Fuel Clad Loss column) lists the categories (types) of fission Product barrier thresholds. The fission product barrier &#xb6;.ategories are:
A. RCS or SG Tube Leakage B. Inadequate Heat removal C. CMT Radiation / RCS Activity D. CMT Integrity or Bypass E. Emergency Coordinator Judgment Each category occupies a row in Table F-I thus forming a matrix defined by the categories.
The intersection of each row with each Loss/Potential Loss column forms a cell in which one or more fission product barrier thresholds appear. If NEI 99-01 does not define a threshold for a barrier Loss/Potential Loss, the word "None" is entered in the cell.
Thresholds are assigned sequential numbers within each Loss and Potential Loss column beginning with number one. In this manner, a threshold can be identified by its category title and number. For example, the first Fuel Clad Barrier Loss in Category A would be assigned "FC Loss A.I ," the third Containment Barrier Potential Loss in Category C would be assigned "CMT P-Loss C.3," etc.
If a cell in Table F-i contains more than one numbered threshold, each of the numbered thresholds, if exceeded, signifies a Loss or Potential Loss of the barrier. It is not necessary to exceed all of the thresholds in a category before declaring a barrier Loss/Potential Loss.
Subdivision of Table F-i by category facilitates association of plant conditions to the applicable fission product barrier Loss and Potential Loss thresholds. This structure promotes a systematic approach to assessing the classification status of the fission product barriers.
When equipped with knowledge of plant conditions related to the fission product barriers, the EAL-user first scans down the category column of Table F-i, locates the likely category and then reads across the fission product barrier Loss and Potential Loss thresholds in that category to determine if a threshold has been exceeded. If a threshold has not been exceeded, the EAL-user proceeds to the next likely category and continues review of the thresholds in the new category If the EAL-user determines that any threshold has been exceeded, by definition, the barrier is lost or potentially lost - even if multiple thresholds in the same barrier column are exceeded, only that one barrier is lost or potentially lost. The EAL-user must examine each of the three fission product barriers to determine if other barrier thresholds in the category are lost or potentially lost. For example, if containment radiation is sufficiently high, a Loss of the Fuel Clad and RCS Barriers and a Potential Loss of the Containment Barrier can occur. Barrier S[Document No.]                                 Rev. 0          IPage                   189 of 240I

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Losses and Potential Losses are then applied to the algorithms given in EALs FG1 .1, ES1 .1, and FA1 .1 to determine the appropriate emergency classification.
In the remainder of this Attachment, the Fuel Clad Barrier threshold bases appear first, followed by the RCS Barrier and finally the Containment Barrier threshold bases. In each barrier, the bases are given according category Loss followed by category Potential Loss beginning with Category A, then B,..., E.
S[Document No.]                             Rev. 0                           Page 190 of 240I

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Table F-I Fission Product Barrier Threshold Matrix Fuel Clad (FC) Barrier                                              Reactor Coolant System (RCS) Barrier                                                        Containment (CMT) Barrier Category                    LOSS                            Potential Loss                                    Loss                              Potential Loss                                  Loss                          Potential Loss 1.RCS leakage > normal makeup capacity due to EITHER:
S1.                                                                    An automatic or manual ES
* UNISOLABLE RCS leakage At                                                                                            actuation required by EITHER:                &deg; SG tube leakage RCS or                     None                      1. RVLS -<O(Note 9)
* UNISOLABLE RCS                                                             1. Aleaking SG ia FAULTED outside                       None of containment SG Tube                                                                                                leakage                         2. RCS cooldown < 400&deg;F Leakage                                                                                             *SG tube RUPTURE                        at > 100&deg;Flhr OR HPI has operated in the injection mode with no RCPs operating
: 1. CETCs > 700&deg;F                                                                 1. RCS heat removal cannot be B                                                                                                                                       established                                                                1. CETCs >1200&deg;F Indqae 1    E~        20F2.                       RCS beat removal cannot be                                                           AND                                                 None                       AND Heat                                                  esalse                                                  oeRSsboln                                         FRestoration                                                         procedurea not Removal                                                    AND                                                                                                                                                             effective within 15 mam.(Note 1)
RCS subcooting    < 0 "F                                                      2. HPI forced cooling initiated C                                                                                         1. containment radiation:
: 1. 1/213RIA 57/58 > Table F-2                                                           &deg; 1,3 RIA57158 >1.0 R/hr                                                                                          1   12315/vTbeF2clm CMT            column "FC Loss"                                    None
* 2 RIA 57 > 1.6 Rlhr                              None                                    None                      C1 T PotentiA 75 Lossa ble         oum Radiation     2.CooanTativtyt30ntiial              lLossA8>.Olh I RCS       2     oln      ciiy>30pim           E                                                          I   8>10Rh Activity     ________________
: 1. Containment isolation is required                           1. Containment pressure      > 59 psig AND EITHER:
* Containment integrity has     2. Containment hydrogen D                                                                                                                                       been lost based on EmrecDoriao concentration   >  4%
CMT                         None                                    None                                      None                                      None                            judgment Integrity                                                                                                                                                                              *UNISOLABLE pathway from        3. Containment presaure >10 psig Continmet tothewith               < one full train of containment or BypacssCnanetoth                                                                                                                                                                                                        heat removal system (1 RBS with >
envionmet euats700               gpm spray flow OR 2 RBCUs) operating per design for > 15 mai.
: 2. Indications of RCS leakage            (Note 1) outside of Containment E         1. Any condition in the opinion of      1.Any condition in the judgment of     1. An1.dto n h ugeto                          Any condition in the judgment of    1. Any condition in the judgment of  1. Any condition in the judgment of the Emergency Coordinator that           the Emergency Coordinator that       . n odto ntejdmn f                        the Emergency Coordinator that          the Emergency Coordinator that        the Emergency Coordinator that EC     indicates loss of the fuel clad          indicates potential loss of the         the Emergency Coordinator that             indicates potential loss of the RCS     indicates loss of the Containment     indicates potential loss of the Judgment          barrier                                  Fuel Clad Banrier                        indicates loss of the RCS Barrder         Barrier                                 Barrier                               Containment Barrier I[DOcument NO.]                                                                                             rRev.                      0                                                                                      Page 191 of 240

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier:              Fuel Clad Category:             A. RCS or SG Tube Leakage Degradation Threat:   Loss Threshold:
None I[Document No.]                  I        Rev. 0          IPage              192 of 240I

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier:                     Fuel Clad Category:                    A. RCS or SG Tube Leakage Degradation Threat:          Potential Loss Threshold:
: 1. RVLS < 0" (Note 9)I Note 9: RVLS is not valid if EITHER of the following exists:
        - One or more RCPs are running OR
        - LPI pump(s) are running AND taking suction from the LPI drop line Definition(s):
None Basis:
RVLS indicated level < 0" with all RCPs not running and both LPI pumps taking suction from the drop line not running represents reactor vessel level below the bottom of the RCS hotleg (without instrument uncertainty considered). This is the lowest measurable reactor vessel level and is used in lieu of actual reactor vessel level indication of level at or below top of active fuel.
This reading indicates a reduction in reactor vessel water level sufficient to allow the onset of heat-induced cladding damage.
ONS Basis Reference(s):
: 1. OSC-2820 Emergency Procedure Setpoints, Setpoint No. 7.1.6.5
: 2. NEI 99-01 RCS or SG Tube Leakage Potential Loss 1I.A
[Document No.]                                      Rev. 0                       Page 193 of 240

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier:                Fuel Clad Category:               B. Inadequate Heat Removal Degradation Threat:     Loss Threshold:
1l. CETCs > 1200&deg;F Definition(s):
iNone Basis:
CETCs > 1200&deg;F indicates extreme ICC conditions that may result in at least 516&deg;F of superheat.
This reading indicates temperatures within the core are sufficient to cause significant superheating of reactor coolant.
ONS Basis Reference(s):
: 1. OSC-2820 Emergency Procedure Setpoints, Setpoint No. 7.1 .1.7
: 2. NEI 99-01 Inadequate Heat Removal Loss 2.A S[Document No.]                             Rev. 0          {Page                    194 of 240

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier:                Fuel Clad Category:               B. Inadequate Heat Removal Degradation Threat:     Potential Loss Threshold:
: 1. CET3s > 700&deg;F Definition(s):
None                                                          i Basis:
CETCs > 7000&deg;F indicates conditions that may result in at least -1 6&deg;F of superheat and that may indicate core uncovery.
This reading indicates a reduction in reactor vessel water level sufficient to allow the onset of heat-induced cladding damage.
ONS Basis Reference(s):
: 1. OSC-2820 Emergency Procedure Setpoints, Setpoint No. 7.1.1.6
: 2. NEI 99-01 Inadequate Heat Removal Potential Loss 2.A I[Document No.] "                            Rev. 0                             Page 195 of 240

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier:                  Fuel Clad Category:                 B. Inadequate Heat Removal Degradation Threat:       Potential Loss Threshold:
: 2. RCS heat removal cannot be established AND RCS subcooling < 00 F Definition(s):
None Basis:
In combination with RCS Potential Loss B.1, meeting this threshold results in a Site Area Emergency.
The combination of these conditions indicates the ultimate heat sink function is under extreme challenge (i.e., superheated). This threshold addresses loss of functions required for hot shutdown with the reactor at pressure and temperature and thus a potential loss of the Fuel Clad Barrier (ref. 1).
This condition indicates an extreme challenge to the ability to remove RCS heat using the steam generators (i.e., loss of an effective secondary-side heat sink). This condition represents a potential loss of the Fuel Clad Barrier. In accordance with EOPs, there may be unusual accident conditions during which operators intentionally reduce the heat removal capability of the steam generators; during these conditions, classification using threshold is not warranted.
ONS Basis Reference(s):
: 1. OSC-2820 Emergency Procedure.Setpoints, Setpoint No. 7.1.3.1
: 2. NEI 99-01 Inadequate Heat Removal Potential Loss 2.B S[Document No.]                                Rev. 0                          Page 196 of 240

                                          *ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier:                   Fuel Clad Category:                  C. CMT Radiation / RCS Activity Degradation Threat:        Loss Threshold:
Table F-2       Containment Radiation - R/hr (1/2/3RIA 57/58,)
TieAfe SDFC             Loss          CMT Potential Loss (Hrs)           RIA 57      RIA 58      RIA 57      RIA 58 0O-<O0.5           300          140        1500        700 0.5 - <2.0           80          40         400         195 2.0O- <8.0           32           15         160         75
                        > 8.0             10            5          50         25 Definition(s):
None Basis:
The specified containment radiation monitor readings (ref. 1) indicate the release of reactor coolant, with elevated activity indicative of fuel damage, into the Containment. The readings are derived assuming the instantaneous release and dispersal of the reactor coolant noble gas and iodine inventory associated with approximately 4% fuel cladding failure 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. This value is higher than that specified for RCS barrier Loss #3.
Monitors used for this fission product barrier loss threshold are the Containment High Range Radiation Monitors RIA 57 and RIA 58 (ref. 1).
The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the containment, assuming that reactor coolant activity equals 300 pCi/gm dose equivalent I-1 31. Reactor coolant activity above this level is greater than that expected for iodine spikes and corresponds to an approximate range of 2% to 5% fuel clad damage. Since this condition indicates that a significant amount of fuel clad damage has occurred, it represents a loss of the Fuel Clad Barrier.
I[Document No.]                                 Rev. 0                           Page 197 of 240

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases The radiation monitor reading in this threshold is higher than that specified for RCS Barrier Loss threshold C.1 since it indicates a loss of both the Fuel Clad Barrier and the RCS Barrier.
Note that a combination of the two monitor readings appropriately escalates the ECL to a Site Area Emergency.
ONS Basis Reference(s):
: 1. OSC-5283 oNs Core Damage Assessment Guidelines, Rev. 2, 2/27/12
: 2. NEI 99-01 CMT Radiation / RCS Activity FC Loss 3.A I [Document    No.]                           *Rev.00                            Page 198 off 240 II4
[DI   metNo]Pae19

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier:                  Fuel Clad Category:                 C. CMT Radiation / RCS Activity Degradation Threat:       Loss Threshold:
S2. Coolant atvy> 300 p~/IDEl Definition(s):
None Basis:
Elevated reactor coolant activity represents a potential degradation in the level of safety of the plant and a potential precursor of more serious problems. The threshold Dose Equivalent 1-131 (DEl) concentration is well above that expected for iodine spikes and corresponds to about 2%
to 5% fuel clad damage. When reactor coolant activity reaches this level the Fuel Clad Barrier is considered lost (ref. 1).
This threshold indicates that RCS radioactivity concentration is greater than 300 pCi/gm dose equivalent I-131. Reactor coolant activity above this level is greater than that expected for iodine spikes and corresponds to an approximate range of 2% to 5% fuel clad damage. Since this condition indicates that a significant amount of fuel clad damage has occurred, it represents a loss of the Fuel Clad Barrier.
It is recognized that sample collection and analysis of reactor coolant with highly elevated activity levels could require several hours to complete. Nonetheless, a sample-related threshold is included as a backup to other indications.
There is no Potential Loss threshold associated with RCS Activity / Containment Radiation.
ONS Basis Reference(s):
: 1. NEI 99-01 CMT Radiation / RCS Activity Fuel Clad Loss 3.B l[Document No.]                      I         Rev. 0
* IPage                  199 of 240

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier:              Fuel Clad Category:             C. CMT Radiation / RCS Activity Degradation Threat:   Potential Loss Threshold:
SNoneI
*[Document No.]                           Rev. 0 *Page                       200 of 240

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier:              Fuel Clad Category:             D. CMT Integrity or Bypass Degradation Threat:   Loss Threshold:
None S[Document No.]                           Rev. 0                          Page 201 of 240

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier:              Fuel Clad Category:             D. CMT Integrity or Bypass Degradation Threat:   Potential Loss Threshold:
None S[Document No.]                  [Rev.         0                          Page 202 of 240

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier:                   Fuel Clad Category:                  E. Emergency Coordinator Judgment Degradation Threat:        Loss Threshold:
: 1. Any condition in the judgment of the Emergency Coordinator that indicates loss of the Fuel Clad Barrier Definition(s):   I None Basis:
The Emergency Coordinator judgment threshold addresses any other factors relevant to determining if the Fuel Clad Barrier is lost. Such a determination should include imminent barrier degradation, barrier monitoring capability and dominant accident sequences.
* Imminent barrier degradation exists if the degradation will likely occur within relatively short period of time based on a projection of current safety system performance. The term "imminent" refers to recognition of the inability to reach safety function acceptance criteria before completion of all checks.
* Barrier monitoring capability is decreased if there is a loss or lack of reliable indicators.
This assessment should include instrumentation operability concerns, readings from portable instrumentation and consideration of offsite monitoring results.
* Dominant ac~cident sequences lead to degradation of all fission product barriers and likely entry to the EOPs. The Emergency Coordinator should be mindful of the Loss of AC power (Station Blackout) and ATWS EALs to assure timely emergency classification declarations.
This threshold addresses any other factors that are to be used by the Emergency Coordinator in determining whether the Fuel Clad barrier is lost ONS Basis Reference(s):
: 1. NEI 99-01 Emergency Director Judgment Fuel Clad Loss 6.A
[Document No.]                       I        Rev. 0                             Page 203 of 240

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier:                   Fuel Clad Category:                  E. Emergency Coordinator Judgment Degradation Threat:        Potential Loss Threshold:
: 1. Any condition in the1 judgment of the Emergency Coordinator that indicates potential loss of the Fuel Clad Barrier Basis:
The Emergency Coordinator judgment threshold addresses any other factors relevant to determining if the Fuel Clad Barrier is potentially lost. Such a determination should include imminent barrier degradation, barrier monitoring capability and dominant accident sequences.
* Imminent barrier deqradation exists if the degradation will likely occur within relatively short period of time based on a projection of current safety system performance. The term "imminent", refers to recognition of the inability to reach safety function acceptance criteria before completion of all checks.
* Barrier monitorinq capability is decreased if there is a loss or lack of reliable indicators.
This assessment should include instrumentation operability concerns, readings from portable instrumentation and consideration of offsite monitoring results.
* Dominant accident sequences lead to degradation of all fission product barriers and likely entry to the EOPs. The Emergency Coordinator should be mindful of the Loss of AC power (Station Blackout) and ATWS EALs to assure timely emergency classification declarations.
This threshold addresses any other factors that are to be used by the Emergency Coordinator in determining whether the Fuel Clad barrier is potentially lost. The Emergency Coordinator should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored.
ONS Basis Reference(s):
: 1. NEI 99-01 Emergency Director Judgment Potential Fuel Clad Loss 6.A I[Document No.]                                Rev. 0             IPage                  204* of 240r

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier:                   Reactor Coolant System Category:                  A. RCS or SG Tube Leakage Degradation Threat:       Loss Threshold:
: 1. An automatic or manual ES actuation required by EITHER:
* UNISOLABLE RCS leakage
* SG tube RUPTURE Definition(s):
UNISOLABLE - An open or breached system line that cannot be isolated, remotely or locally.
RUPTURE - The condition of a steam generator in which primary-to-secondary leakage is of sufficient magnitude to require a safety injection.
Basis:
A RCS leak should be considered UNISOLABLE if the leak cannot be isolated within 15 minutes This threshold is based on an UNISOLABLE ROS leak of sufficient size to require an automatic or manual actuation of the Emergency Core Cooling System (ECCS). This condition clearly represents a loss of the RCS Barrier.
This threshold is applicable to unidentified and pressure boundary leakage, as well as identified leakage. It is also applicable to UNISOLABLE RCS leakage through an interfacing system. The mass loss may be into any location      - inside containment, to the secondary-side (i.e., steam generator tube leakage) or outside of containment.
A steam generator with primary-to-secondary leakage of sufficient magnitude to require a
.safety injection is considered to be RUPTURED. If a. RUPTURED steam generator is also FAULTED outside of containment, the declaration escalates to a Site Area Emergency since the Containment Barrier Loss threshold I .A will also be met.
ONS Basis Reference(s):
: 1. UFSAR Section 7.3 Engineered Safeguards Protective System
: 2. NEI 99-0 1 RCS or SG Tube Leakage Reactor Coolant System Loss 1.A S[Document No.]                               Rev. 0                           Page 205 of 240

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier:                  Reactor Coolant System Category:                 A. RCS or SG Tube Leakage Degradation Threat:       Potential Loss Threshold:
: 1. RCS leakage > normal makeup capacity due to EITHER:
* UNISOLABLE RCS leakage/
* SG tube leakage Definition(s):
UNISOLABLE - An open or breached system line that cannot be isolated, remotely or locally.
Basis:
A RCS leak should be considered UNISOLABLE if the leak cannot be isolated within 15 min.
This threshold is based on the inability to maintain liquid inventory within the RCS by normal operation of the High Pressure Injection System (HPI). The HPI includes three pumps. (ref. 1)
Any one HPI pump runout flow rate is 475 gpm (ref. 2).
This threshold is based on an UNISOLABLE RCS leak that results in the inability to maintain pressurizer level within specified limits by operation of a normally used charging (makeup) pump, but an ES actuation has not occurred. The threshold is met when an operating procedure, or operating crew supervision, directs that a HPI (makeup) pump be placed in service to restore and maintain pressurizer level.
This threshold is applicable to unidentified and pressure boundary leakage, as well as identified leakage. It is also applicable to UNISOLABLE RCS leakage through an interfacing system. The mass loss may be into any location       - inside containment, to the secondary-side (i.e., steam generator tube leakage) or outside of containment.
If a leaking steam generator is also FAULTED outside of containment, the declaration escalates to a Site Area Emergency since the Containment Barrier Loss threshold 1 .A will also be met.
ONS Basis Reference(s):
: 1. UFSAR Section 9.3.2 High Pressure Injection System
: 2. OSC-2820 Emergency Procedure Setpoints, Setpoint No. 7.3.1.2
: 3. NEI 99-01 RCS or SG Tube Leakage Reactor Coolant System Potential Loss 1.A Rev. 0                           Page 206 of 240
[Document
[Document No.]                                Rev. 0 Page 206 of 240 I

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier:                  Reactor Coolant System Category:                A. RCS or SG Tube Leakage Degradation Threat:       Potential Loss Threshold:
: 2. RCS cooldown to    < 400&deg; F at > 1000&deg;F/hr OR HPI has operated in the injection mode with no RCPs operating Definition(s):
None Basis:
4000 F is the temperature below which a cooldown greater than 100&deg;F/hr requires implementation of Pressurized Thermal Shock (PTS) guidance (rule 8) (ref. 1, 2).
HPI operating in the injection mode with no RCPs operating also invokes Rule 8 (ref. 3).
This condition indicates an extreme challenge to the integrity of the RCS pressure boundary due to pressurized thermal shock - a transient that causes rapid RCS cooldown while the ROS is in Mode 3 orjhigher (i.e., hot and pressurized).
ONS Basis Reference(s):
: 1. OSC-2820 Emergency Procedure Setpoints, Setpoint No. 7.1.2.7
: 2. OSC-2820 Emergency Procedure Setpoints, Setpoint No. 7.1.8.7
: 3. EP/*/AI1800/001 Rule 8 Pressurized Thermal Shock (PTS)
: 4. NEI 99-0 1 RCS or SG Tube Leakage Reactor Coolant System Potential Loss I .B S[Document No.]                                Rev. 0                        Page 207 of 240

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier:               Reactor Coolant System Category:              B. Inadequate Heat Removal Degradation Threat:    Loss Threshold:
None S[Document No.]                          'Rev. 0                          Page 208 of 240I

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier:                  Reactor Coolant System Category:                 B. Inadequate Heat Removal Degradation Threat:       Potential Loss Threshold:
: 1. ROS heat removal cannot be established
          *AND RCS subcooling < 0&deg;F Definition(s):
None Basis:
In combination with FC Potential Loss B.1, meeting this threshold results in a Site Area Emergency.
The combination of these conditions indicates the ultimate heat sink function is under extreme challenge (i.e., superheated). This threshold addresses loss of functions required for hot shutdown with the reactor at pressure and temperature and thus a potential loss of the RCS Barrier.
This condition indicates an extreme challenge to the ability to remove RCS heat using the steam generators (i.e., loss of an effective secondary-side heat sink). This condition represents a potential loss of the RCS Barrier. In accordance with EOPs, there may be unusual accident conditions during which operators intentionally reduce the heat removal capability of the steam generators; during these conditions, classification using threshold is not warranted.
Meeting this threshold results in a Site Area Emergency because this threshold is identical to Fuel Clad Barrier Potential Loss threshold B.2; both will be met. This condition warrants a Site Area Emergency declaration because inadequate RCS heat removal may r~esult in fuel heat-up
 .sufficient to damage the cladding and increase RCS pressure to the point where mass will be lost from the system.
ONS Basis Reference(s):
: 1. OSC-2820 Emergency Procedure Setpoints, Setpoint No. 7.1.3.6
: 2. NEI 99-01 Inadequate Heat Removal RCS Loss 2.B I[Document No.]                      1         Rev. 0         IPage                  209 of 240i

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier:                Reactor Coolant System Category:               B. Inadequate Heat Removal Degradation Threat:     Potential Loss Threshold:
J2. PIforc~ed coolingintae Definition(s):
None Basis:
HPI Forced Cooling (Rule 4) is used when the SGs are not capable of heat removal and RCS pressure is greater than 2300 psig. A Pressurizer PORV is opened to relieve pressure until HPI cools the reactor (feed and bleed). (ref. 1)
ONS Basis Reference(s):
: 1. OSC-2820 Emergency Procedure Setpoints, Setpoint No. 7.1.4.16
: 2. NEI 99-01 Other Indications Potential Loss 5.A S[Document No.1                               Rev. 0                       .Page 210 of 240

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier:                  Reactor Coolant System Category:                 C. CMT Radiation!/RCS Activity Degradation Threat:       Loss Threshold:
: 1. Containment radiation:
* 1,3 RIA57/58 >1.0OR/hr
* 2R1A57>1.6R/hr
* 2RIA58>1.0R/hr Definition(s):
N/A Basis:
Containment radiation monitor readings greater than the specified values (ref. 1) indicate the release of reactor coolant to the Containment. The readings assume the instantaneous release and dispersal of the reactor coolant noble gas and iodine inventory associated with normal operating concentrations (i.e., within Technical Specifications) into the Containment atmosphere. Because of the very high fuel clad integrity, only small amounts of noble gases would be dissolved in the primary coolant.
Monitors used for this fission product barrier loss threshold are the Containment High Range Radiation Monitors RIA-57 and RIA-58. The difference in the threshold values is due to the relative strength of the detector check source which affects the background readings for the detector (the source for 2RIA-57 is stronger than that for the other detectors). (ref. 1)
The radiation monitor reading corresponds, to an instantaneous release of all reactor coolant mass into the containment, assuming that reactor coolant activity equals Technical Specification allowable limits. This value is lower than that specified for Fuel Clad Barrier Loss threshold C.1 since it indicates a loss of the RCS Barrier only.
There is no Potential Loss threshold associated with RCS Activity / Containment Radiation.
ONS Basis Reference(s):
: 1. OSC-4244 ONS High Range Containment Monitor Correlation Factors for RIA-57 and RIA-58
: 2. NEI 99-01 CMT Radiation / RCS Activity RCS Loss 3.A S[Document No.]                               Rev. 0                           Page 211 of 240I

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier:              Reactor Coolant System Category:             B. CMT Radiation!/RCS Activity Degradation Threat:   Potential Loss Threshold:
None
[Document No.]                           Rev. 0                          Page 212 of 240

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier:              Reactor Coolant System Category:             D. CMT Integrity or Bypass Degradation Threat:   Loss Threshold:
None I[Document No.]                 I         Rev.. 0         IPage              213 of 240

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier:              Reactor Coolant System Category:             D. CMT Integrity or Bypass Degradation Threat: Potential Loss Threshold:
NoneI S[Document No.]                           Rev. 0                          Page 214 of 240

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier:                  Reactor Coolant System Category:                 E. Emergency Coordinator Judgment Degradation Threat:       Loss Threshold:
Definition(s):
None Basis:
The Emergency Coordinator judgment threshold addresses any other factors relevant to determining if the RCS Barrier is lost. Sucha determination should include imminent barrier degradation, barrier monitoring capability and dominant accident sequences.
* Imminent barrier de qradation exists if the degradation will likely occur within relatively short period of time based on a projection of current safety system performance. The term "imminent" refers to recognition of the inability to reach safety function acceptance criteria before completion of all checks.
* Barrier monitorinq capability is decreased if there is a loss or lack of reliable indicators.
This assessment should include instrumentation operability concerns, readings from portable instrumentation and consideration of offsite monitoring results.
* Dominant accident sequences lead to degradation of all fission product barriers and likely entry to the EOPs. The Emergency Coordinator should be mindful of the Loss of AC power (Station Blackout) and ATWS EALs to assure timely emergency classification declarations.
This threshold addresses any other factors that may be used by the Emergency Coordinator in determining whether the RCS Barrier is lost.
ONS Basis Reference(s):
: 1. NEI 99-01 Emergency Director Judgment RCS Loss 6.A IDocument No.]                               Rev. 0                             Page 215 of 240

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier:                  Reactor Coolant System Category:                 E. Emergency Coordinator Judgment Degradation Threat:       Potential Loss Threshold:
: 1. Any corjdition in the judgment of the Emergency Coordinator that indicates potential loss of the RCS Barrier Definition(s):
None Basis:
The Emergency Coordinator judgment threshold addresses any other factors relevant to determining if the RCS Barrier is potentially lost. Such a determination should include imminent barrier degradation, barrier monitoring capability and dominant accident sequences.
* Imminent barrier degradation exists if the degradation will likely occur within relatively short period of time based on a projection of current safety system performance. The term "imminent" refers to recognition of the inability to reach safety function acceptance criteria before completion of all checks.
* Barrier monitorinq capability is decreased if there is a loss or lack of reliable indicators.
This assessment should include instrumentation operability concerns, readings from portable instrumentation and consideration of offsite monitoring results.
* Dominant accident sequences lead to degradation of all fission product barriers and likely entry to the EOPs. The Emergency Coordinator should be mindful of the Loss of AC power (Station Blackout) and ATWS EALs to assure timely emergency classification declarations.
This threshold addresses any other factors that may be used by the Emergency Coordinator in determining whether the RCS Barrier is potentially lost. The Emergency Director should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored.
ONS Basis Reference(s):
: 1. NEI 99-01 Emergency Director Judgment RCS Potential Loss 6,A
[Document No.]                      I        Rev. 0            IPage                  216 of 240

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier:                   Containment Category:                  A. RCS or SG Tube Leakage Degradation Threat:        Loss Threshold:
: 1. A leaking SG is FAULTED outside of containment Definition(s):
FAULTED - The term applied to a steam generator that has a steam leak on the secondary side of sufficient size to cause an uncontrolled drop in steam generator pressure or the steam generator to become completely depressurized.
Basis:
This threshold addresses a leaking Steam Generator (SG) that is also FAULTED outside of containment. The condition of the SG leakage, is determined in accordance with the thresholds for RCS Barrier Potential Loss A.1 and Loss A.1, respectively. This condition represents a bypass of the containment barrier.
FAULTED is a defined term within the NEI 99-01 methodology; this determination is not necessarily dependent upon entry into, or diagnostic steps within, an EOP. For example, if the pressure in a steam generator is decreasing uncontrollably (part of the FAULTED definition) and the FAULTED steam generator isolation procedure is not entered because EOP user rules are dictating implementation of another procedure to address a higher priority condition, the steam generator is still considered FAULTED for emergency classification purposes.
The FAULTED criterion establishes an appropriate lower bound on the size of a steam release that may require an emergency classification. Steam releases of this size are readily observable with normal Control Room indications. The lower bound for this aspect of the containment barrier is analogous to the lower bound criteria specified in IC SU4 for the fuel clad barrier (i.e., RCS activity values) and IC SU5 for the RCS barrier (i.e., RCS leak rate values).
This threshold also applies to prolonged steam releases necessitated by operational considerations such as the forced steaming of a leaking steam generator directly to atmosphere to cooldown the plant. These type of condition will result in a significant and sustained release of radioactive steam to the environment (and are thus similar to a FAULTED condition). The inability to isolate the steam flow without an adverse effect on plant cooldown meets the intent of a loss of containment.
Steam releases associated with the expected operation of a SG Atmospheric Dump Valve(s) do not meet the intent of this threshold. Such releases may occur intermittently for a short period of time following a reactor trip as operators process through emergency operating procedures to bring the plant to a stable condition and prepare to initiate a plant cooldown.
Steam releases associated with the unexpected operation of a valve (e.g., a stuck-open safety valve) do meet this threshold.
I[Document No.]                                Rev. 0                           Page 217 of 240I

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Following an SG tube leak, there may be minor radiological releases through a secondary-side system component (e.g., air ejectors, gland seal exhausters, valve packing, steam traps, terry turbine exhaust, etc.). These types of releases do not constitute a loss or potential loss of containment but should be evaluated using the Recognition Category R ICs.
The ECLs resulting from primary-to-secondary leakage, with or without a steam release from the FAULTED SG, are summarized below.
Affected SG is FAULTED Outside of Containment?-
P-to-S Leak Rate                        Yes                       No Less than or equal to 25 gpm             No classification       No classification Greater than 25 gpm                 Unusual Event per SU5.1  Unusual Event per SU5.1 Greater than normal makeup pump capacity (RCS Barrier          Site Area Emergency per PotentialLoss)                                FSI .1             Alert per FAI.1 Requires an automatic or manual Site Area Emergency per ECCS (ES) actuation (RCS                                         'Alert per FAI.1 FSI1.1 BarrierLoss)
There is no Potential Loss threshold associated with RCS or SG Tube Leakage.
ONS Basis Reference(s):
: 1. NEI 99-01 RCS or SG Tube Leakage Containment Loss I .A S[Document No.]                        IRev.            0                            Page 218 of 240

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier:              Containment Category:             A. RCS or SG Tube Leakage Degradation Threat:   Potential Loss Threshold:
Nonei I[Document No.]                           Rev. 0                          Page 219 of 240[

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier:              Containment Category:             B. Inadequate Heat Removal Degradation Threat:   Loss Threshold:
None I[Document No.]                           Rev. 0          IPage 220 of 240I

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier:                   Containment Category:                  B. Inadequate Heat Removal Degradation Threat:        Potential Loss Threshold:
: 1. CETCs >12000 F AND Restoration procedures not effective within 15 min. (Note 1)
Note 1:   The Emergency Coordinator should declare qthe event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
Definition(s):
None Basis:
Core Exit Thermocouples (CETCs) are a component of Inadequate Core Cooling Instrumentation and provide an indirect indication of fuel clad temperature by measuring the temperature of the reactor coolant that leaves the core region. Although clad rupture due to high temperature is not expected for CETC readings less than the threshold, temperatures of this magnitude signal significant superheating of the reactor coolant and core uncover (ref. 1).
The restoration procedures are those emergency operating procedures that address the recovery of the RCS and core heat removal acceptance criteria. The procedure is considered effective if the temperature is decreasing or if the vessel water level is increasing (ref. 1 ).The 15 minute threshold starts when operator action begins taking procedurally directed functional recovery actions.
If CETC readings are greater than I1,200&deg;F, Fuel Clad barrier is also lost.
This threshold addresses any other factors that may be used by the Emergency Coordinator in determining whether the Containment Barrier is potentially lost. The Emergency Director should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored.
ONS Basis Reference(s):
: 1. EP/1 ,2,3/A/1 800/00 1 Inadequate Core Cooling
: 2. NEI 99-01 Inadequate Heat Removal Containment Potential Loss 2.A EDocument No.]                                      Rev. 0          IPage                      221 of 240I

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier:              Containment Category:             C. CMT Radiation/RCS Activity Degradation Threat:   Loss Threshold:
None I[Document No.]                           Rev. 0          IPage              222 of 240I

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier:                     Containment Category:                    C. CMT Radiation/RCS Activity Degradation Threat:          Potential Loss Threshold:
I1. 1/2/3RIA 57/58    >  Table F-2 column "CTPtnilLoss"i Table F-2       Containment Radiation    - R/hr (112/3RIA 57/58)
TieAfe SDFC            Loss         CMT Potential Loss (Hrs)            RIA 57     RIA 58     RIA 57       RIA 58 0- <O.5             300         140       1500          700 0.5 - <2.0             80         40         400          195 2.0O- <8.0             32          15        160           75
                         > 8.0               10          5         50           25 Definition(s):
None Basis:
Containment radiation monitor readings greater than the values shown (ref. 1) indicate significant fuel damage well in excess of that required for loss of the RCS Barrier and the Fuel Clad Barrier.
The specified containment radiation monitor readings (ref. 1) indicate the release of reactor coolant, with significant fuel damage well in excess of that required for loss of the RCS Barrier and the Fuel Clad Barrier, into the Containment. The readings are derived assuming the instantaneous release and dispersal of the reactor coolant noble gas and iodine inventory associated with 20% clad failure into the Containment atmosphere.
Containment radiation readings at or above the Containment Barrier Potential Loss threshold signify a loss of two fission product barriers and Potential Loss of a third, indicating the need to upgrade the emergency classification to a General Emergency.
Monitors used for this fission product barrier loss threshold are the Containment High Range Radiation Monitors RIA-57 and RIA-58 (ref. 1).
I[Document No.]                                  Rev. 0                            Page 223 of 240

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the containment, assuming that 20% of the fuel cladding has failed. This level of fuel clad failure is well above that used to determine the analogous Fuel Clad Barrier Loss and RCS Barrier Loss thresholds.
NUREG-1 22*8, Source Estimations During Incident Response to Severe Nuclear Power Plant Accidents, indicates the fuel clad failure must be greater than approximately 20% in order for there to be a major release of radioactivity requiring offsite protective actions. For this condition to exist, there must already have been a loss of the ROS Barrier and the Fuel Clad Barrier. It is therefore prudent to treat this condition as a potential loss pf containment which would then escalate the ECL to a General Emergency.
ONS Basis Reference(s):
: 1. OSC-5283 ONS Core Damage Assessment Guidelines, Rev. 2, 2/27/12
: 2. NEI 99-01 CMT Radiation / RCS Activity Containment Potential Loss 3.A S[Document No.]                                  Rev. 0                            Page 224 of 240

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier:                   Containment Category:                  D. CMT Integrity or Bypass Degradation Threat:        Loss Threshold:
: 1. Containment isolation is required AND EITHER:
* Containment integrity has been lost based on Emergency Coordinator' judgment
* UNISOLABLE pathway from Containment to the environment exists Definition(s):
UNISOLABLE - An open or breached system line that cannot be isolated, remotely or locally.
Basis:
The pathway should be considered UNISOLABLE if the Containment cannot be isolated within 15 min.
Reactor Building Essential and Non-essential Isolation occurs on an Engineered Safeguards signal of 3 psig (ref. 1).
These thresholds address a situation where containment isolation is required and one of two conditions exists as discussed below. Users are reminded that there may be accident and release conditions that simultaneously meet both bulleted thresholds.
First Threshold   -  Containment integrity has been lost, i.e., the actual containment atmospheric leak rate likely exceeds that associated with allowable leakage (or sometimes referred to as design leakage). Following the release of RCS mass into containment, containment pressure will fluctuate based on a variety of factors; a loss of containment integrity condition may (or may not) be accompanied by a noticeable drop .in containment pressure. Recognizing the inherent difficulties in determining a containment leak rate during accident conditions, it is expected that the Emergency Coordinator will assess this threshold using judgment, and with due consideration given to current plant conditions, and available operational and radiological data (e.g., containment pressure, readings on radiation monitors outside containment, operating status of containment pressure control equipment, etc.).
Refer to the middle piping run of Figure 1. Two simplified examples are provided. One is leakage from a penetration and the other is leakage from an in-service system valve.
Depending upon radiation monitor locations and sensitivities, the leakage could be detected by any of the four monitors depicted in the figure.
Another example would be a loss or potential loss of the RCS barrier, and the simultaneous occurrence of two FAULTED locations on a steam generator where one fault is located inside containment (e.g., on a steam or feedwater line) and the other outside of containment. In this case, the associated steam line provides a pathway for the containment atmosphere to escape to an area outside the containment.
S[Document No.]                                 Rev. 0                            Page 225 of 240

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Following the leakage of RCS mass into containment and a rise in containment pressure, there may be minor radiological releases associated with allowable (design) containment leakage through various penetrations or system components. These releases do not constitute a loss or potential loss of containment but should be evaluated using the Recognition Category R l~s.
Second Threshold     -  Conditions are such that there is an UNISOLABLE pathway for the migration of radioactive material from the containment atmosphere to the environment. As used here, the term "environment" includes the atmosphere of a room or area, outside the containment, that may, in turn, com~municate with the outside-the-plant atmosphere (e.g.I, through discharge of a ventilation system or atmospheric leakage). Depending upon a variety of factors, this condition may or may not be accompanied by a noticeable drop in containment pressure.
Refer to the top piping run of Figure 1. In this simplified example, the inboard and outbdard isolation valves remained open after a containment isolation was required (i.e., containment isolation was not successful). There is now an UNISOLABLE pathway from the containment to the environment.
The existence of a filter is not considered in the threshold assessment. Filters do not remove fission product noble gases. In addition, a filter could become ineffective due to iodine and/or particulate loading beyond design limits (i.e., retention ability has been exceeded) or water saturation from steam/high humidity in the release stream.
Leakage between two interfacing liquid systems, by itself, does not meet this threshold.
*Refer to the bottom piping run of Figure 1. In this simplified example, leakage in an RCP seal cooler is allowing radioactive material to enter the Auxiliary Building. The radioactivity would be detected by the Process Monitor. If there is no leakage from the closed water cooling system to the Auxiliary Building, then no threshold has been met. If the pump developed a leak that allowed steam/water to enter the Auxiliary Building, then second threshold would be met. Depending upon radiation monitor locations and sensitivities, this leakage could be detected by any of the four monitors depicted in the figure and cause the first threshold to be met as well.
Fol~lowing the leakage of RCS mass into containment and a rise in containment pressure, there may be minor radiological releases associated with allowable containment leakage through various penetrations or system components. Minor releases may also occur if a containment isolation valve(s) fails to close but the containment atmosphere escapes to an enclosed system. These releases do not constitute a loss or potential loss of containment but should be evaluated using the Recognition Category R ICs.
The status of the containment barrier during an event involving steam generator tube leakage is assessed using Loss Threshold A.1.
ONS Basis Reference(s):
: 1. UFSAR Section 6.2.3 Containment Isolation System
: 2. NEI 99-01 CMT Integrity or Bypass Containment Loss 4.A
[Document No.]                                 Rev. 0                          Page 226 of 240

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier:                  Containment Category:                 D. CMT Integrity or Bypass Degradation Threat:       Loss Threshold:
: 2. Indications of RCS leakage outside of q3ontainment                                     1 Definition(s):
None Basis:
Containment sump, temperature, pressure and/or radiation levels will increase if reactor coolant mass is leaking into the containment. If these parameters have not increased, then the reactor coolant mass may be leaking outside of containment (i.e., a containment bypass sequence). Increases in sump, temperature, pressure, flow and/or radiation level readings outside of the containment may indicate that the RCS mass is being lost outside of containment.
Unexpected elevated readings and alarms on radiation monitors with detectors outside containment should be corroborated with other available indications to confirm that the source is a loss of RCS mass outside of containment. If the fuel clad barrier has not been lost, radiation monitor readings outside of containment may not increase significantly; however, other unexpected changes in sump levels, area temperatures or pressures, flow rates, etc.
should be sufficient to determine if RCS mass is being lost outside of the containment.
Refer to the middle piping run of Figure 1. In this simplified example, a leak has occurred at a reducer on a pipe carrying reactor coolant in the Auxiliary Building. Depending upon radiation monitor locations and sensitivities, the leakage could be detected by any of the four monitors depicted in the figure and cause threshold D.1 to be met as well.
To ensure proper escalation of the emergency classification, the RCS leakage outside of containment must be related to the mass loss that is causing the RCS Loss and/or Potential Loss threshold A.1 to be met.
ONS Basis Reference(s):
: 1. NEI 99-01 CMT Integrity or Bypass Containment Loss
*.[Document No.]                              Rev. 0                           Page 227 of 240

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Figure 1: Containment Integrity or Bypass Examples Seal Cooling
{[Document No.]                             Rev. 0                           Page 228 of 240I

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier:                 Containment Category:                D. CMT Integrity or Bypass Degradation Threat:      Potential Loss Threshold:
: 1. Containment pressure > 59 psig Definition(s):
None Basis:
The Reactor Building is designed for an internal pressure of 59 psig (ref. 1).
If containment pressure exceeds the design pressure, there exists a potential to lose the Containment Barrier. To reach this level, there must be an inadequate core cooling condition for an extended period of time; therefore, the RCS and Fuel Clad barriers would already be lost. Thus, this threshold is a discriminator between a Site Area Emergency and General Emergency since there is now a potential to lose the third barrier.
ONS Basis Reference(s):
: 1. UFSAR Section 6.2.1 Containment Functional Design
: 2. NEI 99-01 CMT Integrity or Bypass Containment Potential Loss 4.A S[Document No.]                               Rev. 0                           Page 229 of 240I

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier:                   Containment Category:                  D. CMT Integrity or Bypass Degradation Threat:        Potential Loss Threshold:
: 2. Containment hydrogen concentration > 4%
Definition(s):
None Basis:
Following a design basis accident, hydrogen gas may be generated inside the containment by reactions such as zirconium metal with water, corrosion of materials of construction and radiolysis of aqueous solution in the core and sump.
The 4% hydrogen concentration threshold is generally considered the lower limit for hydrogen deflagrations. ONS is equipped with a Containment Hydrogen Monitoring System (CHMS) that provides continuous indication of hydrogen concentration in the containment atmosphere. The measurement capability is provided over the range of 0% to 10%. A continuous indication of the hydrogen concentration is not required in the control room at all times during normal operation. If continuous indication of the hydrogen concentration is not available at all times, continuous indication and recording shall be functioning within 90 minutes of the initiation of the safety injection. (ref. 1, 2)
The existence of an explosive mixture means, at a minimum, that the containment atmospheric hydrogen concentration is sufficient to support a hydrogen burn (i.e., at the lower deflagration limit). A hydrogen burn will raise containment pressure and could result in collateral equipment damage leading to a-loss of containment integrity. It therefore represents a potential loss of the Containment Barrier.
ONS Basis Reference(s):
: 1. UFSAR Section 9.3.7 Containment Hydrogen Monitoring System
: 2. UFSAR Section 15.16.3 Evaluation of Hydrogen Concentrations
: 3. NEI 99-01 CMT Integrity or Bypass Containment Potential Loss 4.B I[Document No.]                        I        Rev. 0        IPage                  230 of 240

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier:                    Containment Category:                   D. CMT Integrity or Bypass Degradation Threat:         Potential Loss Threshold:
: 3. Containment pressure > 10 psig with < one full train of containment heat removal Isystem (1 RBS with > 700 gpm spray flow OR 2 R'BCUs) operating per design for >-
15 min. (Note 1)
Note 1: The Emergency Coordinator should declare the event promptly upon 'determining that time limit has S been exceeded, or will likely be exceeded.
Definition(s):
None Basis:
Two engineered safeguards systems, the Reactor Building Spray System and the Reactor Building Cooling System, are provided to remove heat from the containment atmosphere following an accident. Both the Reactor Building Spray System and the Reactor Building Cooling System, with either at full capacity, are individually capable of maintaining the containment pressure below the design limit following a LOCA or MS LB. (ref. 1, 3)
        *The Reactor Building Spray (RBS) System consists of two separate trains of equal capacity. Spray flow greater or equal to 700 gpm satisfies the spray flow design requirement. The Reactor Building pressure setpoint (10 psig) is the pressure at which the Reactor Building Spray equipment should actuate and begin performing its function (ref. 1,2, 3, 4).
     ,,   Each of three Reactor Building Cooling Units (RBCUs) consists of a fan, cooling coils, and the required distribution duct work. The Reactor Building atmosphere is circulated past cooling coils by fans and returned to the building. Cooling water-for the cooling units is supplied by the Low Pressure Service Water System. The Reactor Building Cooling System provides the design heat removal capacity with two of three coolers operating (ref. 1).
This threshold describes a condition where containment pressure is greater than the setpoint at which containment energy (heat) removal systems are designed to automatically actuate, and less than one full train of equipment is capable of operating per design. The 15-minute criterion is included to allow operators time to manually start equipment that may not have automatically started, if possible. This threshold represents a potential loss of containment in that containment heat removal/depressurization systems (e.g., containment sprays, ice condenser fans, etc., but not including containment venting strategies) are either lost or performing in a degraded manner.
S[Document. No.]                                     Rev. 0                          Page 231 of 240

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases ONS Basis Reference(s):
: 1. UFSAR Section 6.2.2 Containment Heat Removal Systems
: 2. UFSAR Table 7-2 Engineered Safeguards Actuation Conditions
: 3. UFSAR Table 6-25 Minimum Acceptable Combinations of Containment Heat Removal Equipment Performance
: 4. OSC-02820 Emergency Procedure Setpoints, Setpoint No. 7.4.1.2
: 5. NEI 99-01 CMT Integrity or Bypass Containment Potential Loss 4.C S[Document No.]                            Rev. 0                         Page 232 of 240

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier:                   Containment Category:                  E. Emergency Coordinator Judgment Degradation Threat:        Loss Threshold:
: 1. Any condition in the judgment of the Emergency Coordinator that indicates loss of the Containment B~rrier Definition(s):
None Basis:
The Emergency Coordinator judgment threshold addresses any other factors relevant to determining if the Containment Bar'rier is lost. Such a determination should include imminent barrier degradation, barrier monitoring capability and dominant accident sequences.
* Imminent barrier degradation exists if the degradation will likely occur within relatively short period of time based on a projection of current safety system performance. The term "imminent" refers to recognition of the inability to reach safety function acceptance criteria before completion of all checks.
* Barrier monitoring capability is decreased if there is a loss or lack of reliable indicators.
This assessment should include instrumentation operability concerns, readings from portable instrumentation and consideration of offsite monitoring results.
* Dominant accident seqiuences lead to degradation of all fission product barriers and likely entry to the EOPs. The Emergency Coordinator should be mindful of the Loss of AC power (Station Blackout) and ATWS EALs to assure timely emergency classification declarations.
This threshold addresses any other factors that may be used by the Emergency Coordinator in determining whether the Containment Barrier is lost.
ONS Basis Reference(s):
: 1. NEI 99-01 Emergency Director Judgment PC Loss 6.A I[Document No.]                                 Rev. 0          IPage                  23of 240I

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier:                  Containment Category:                 E. Emergency Coordinator Judgment Degradation Threat:       Potential Loss Threshold:
1.I nny condition in the judgment of the Emergency Coor~dinator that indicates potential loss of the Containment Barrier Definition(s):
None Basis:
The Emergency Coordinator judgment threshold addresses any other factors relevant to determining if the Containment Barrier is potentially lost. Such a determination should include imminent barrier degradation, barrier monitoring capability and dominant accident sequences.
* Imminent barrier decqradation exists if the degradation will likely occur within relatively short period of time based on a projection of current safety system performance. The term "imminent" refers to recognition of the inability to reach safety function acceptance criteria before completion of all checks.
* Barrier monitorincq capability is decreased if there is a loss or lack of reliable indicators.
This assessment should include instrumentation operability concerns, readings from portable instrumentation and consideration of offsite monitoring results.
* Dominant accident sequences lead to degradation of all fission product barriers and likely entry to the EOPs. The Emergency Coordinator should be mindful Of the Loss of AC power (Station Blackout) and ATWS EALs to assure timely emergency classification declarations.
This threshold addresses any other factors that may be used by the Emergency Coordinator in determining whether the Containment Barrier is lost.
ONS Basis Reference(s):
: 1. NEI 99-01 Emergency Director Judgment PC Potential Loss 6.A S[Document No.]                               Rev. 0                            Page 234 of 240

ATTACHMENT 3 Safe Operation & Shutdown Rooms/Areas Tables R-2 & H-2 Bases

===Background===
NEI 99-01 Revision 6 l~s AA3 and HA5 prescribe declaration of an Alert based on impeded access to rooms or areas (due to either area radiation levels or hazardous gas concentrations) where equipment necessary for normal plant operations, cooldown or shutdown is located.
These areas are intended to be plant operating mode dependent. Specifically the Developers Notes for AA3 and HA5 states:
The "site-qpecificlist of plant rooms or areas with entry-relatedmnode applicabilityidentified" should specify those rooms or areas that contain equipment which require a manual/local action as specified in operatingproceduresused for normal plant operation, cooldown and shutdown. Do not include rooms or areas in which actions of a contingent or emergency nature would be performed (e.g., an action to address an off-normal or emergency condition such as emergency repairs, corrective measures or emergency operations). In addition, the list should specify the plant mode(s) during which entry would be requiredfor each room or area.
The list should not include rooms or areas for which entry is required solely to perform actions of an administrativeor recordkeeping nature (e.g., normal rounds or routine inspections).
Further, as specified in IC HA5:
The list need not include the Contfrol Room if adequate engineered safety/design features are in place to preclude a Control Room evacuation due to the release of a hazardousgas.
Such features may include, but are not limited to, capability to draw air from multiple air intakes at different and separate locations, inner and outer atmospheric boundaries, or the capabilityto acquire and maintain positive pressure within the Control Room envelope.
[Document No.]                               Rev. 0                            Page 235 of 240r

ATTACHMENT 3 Safe Operation & Shutdown Rooms/Areas Tables R-2 & H-2 Bases ONS Table R-2 and H-2 Bases NEI 99-01 Rev 06 addresses elevated radiation levels and hazardous gases in certain plant rooms/areas sufficient to preclude or impede personnel from performing actions necessary to maintain normal plant operation, or to perform a normal plant shutdown and cool down.
Power Operation was reviewed to determine if any actions are "necessary" to maintain power operations. Over reasonable periods (several days), there are some actions outside the Control Room that are required to be performed to maintain normal operations. The following table lists the locations into which an operator may be dispatched in order perform a normal plant operation, shutc(lown and cool down.I The review was completed using the following procedures as the controlling documents:
* OP/*/A/1102/010 (Controlling Procedure for Unit Shutdown)
* OP/*IAj1 106/001' (Turbine Generator)
* OP/*/A/1106/015 (EHC System)
* OP/*/A~/1 103/004A (RCS Boration)
* OP/*/A/1104/027 (Bleed Transfer Pump Recirculation)
* PTI*IA106001001 B (Surveillance to go to Mode 3)
* OPP *Ill1021010 (Unit SD Mode I to Mode 3)
* IPI*/A/0200104,7 (LTOP Calibration)
* OP/*/AI1 103/006 (RCP Operations)
* OP/*IA.I1 104/012 (CCW Pump Operations)
* CPI11AI20021014 (RCS Sampling)
* OPI/*IN1104/049 (LTOP Operation)
* OP/11AI1104/001 (Core Flood Operations)
* OP/OIAI1 104/048 (TBS Operations)
* OPI*NIA 104/004 (Low Pressure Injection System)
* OPI*IAI1 103/008 (RCS Crud Burst)
Travel paths to the locations where the equipment is operated were considered as part of the determination of affected rooms. ONS Reactor and Auxiliary Building design consist of mostly single entry rooms located off of a common hallway, therefore access to the hallway is required to access a given room. Some equipment is located within the hallway itself.
Room                  Mode          Procedure      Enclosure       Steps TB                                     I      OP/1/A/1102/010       4.1    Unit SD TB                                     1      OP/1/A/1106/001       4.2    TG TB                                     1      OP/llAI1106/014       4.3    MSRH TB                                     I      OP/1lA/1106/015       4.2    EHC A-2 LOST Hatch area                  1,2,3    OP/llAI1103/004 A     4.1    RCS Boration A-i hallway 8 S/col 65                1,2,3    OP/1lAI1103/004 A     4.2    RCS Boration A-i hallway 8' S/ col 65              1,2,3    OP/I/AI1103/004 A     4.3    RCS Boration Unit I BTP Rm                        1,2,3    OP/llAI1103/004 A     4.3    RCS Boration S[Document No.]                                     Rev. 0                          Page 236 of 240

ATTACHMENT 3 Safe Operation & Shutdown Rooms/Areas Tables R-2 & H-2 Bases Unit 1 BTP Rm                    1,2,3    OP/IA1A1103/OO4A       4.4    RCS     Boration A-i hallway 8' S/cal 65          1,2,3    OP/1/A/1103/004 A      4.4    RCS     Boration A-2-Unit 2 LDST Hatch area       1,2,3    OP/1/A/1103/004A       4.4    RCS     Boration Unit 2 BTP Rm                    1,2,3    OP/i/Ni1103/004 A      4.4    RCS     Boration A-I-hallway N of Cal 82          1,2,3    OP/i/A/1103/004 A      4.4    RCS     Boration A-I hallway 5' S/col 67          1,2,3    OPIll/N1103/004 A      4.4    RCS     Boration A-I hallway cal 82               1,2,3    OP/1/A/1103/004 A      4.4    RCS     Boratian A-2 Unit 3 LDST Hatch area       1,2,3    OPI/i/Ni103/004 A      4.5    RCS     Boration Unit 3 BTP Rm                    1,2,3    OP/1/A/1103/004 A      4.5    RCS     Boratian I A-I hallway 10' S/cal 96)        1,2,3    OPI/i/N1103/004 A      4.5    RCS     Boratian A-I hallway 5' S/cal 67          1,2,3    OP/il/N1103/004 A      4.5    RCS Baratian LPI Caoler Rm 1'W/North daor     1,2,3    OP/il/Ni103/004 A      4.5    RCS Boratian A-i- BAMT Rm                       1      OP/i/Ni1103/004 A      4.6    RCS Boration A-i Unit 1 & 2 BAMT Rm             1      OP/i/Ni1103/004 A      4.6    RCS Boration Rm 111                             I      OP1/i/Ni103/004 A      4.6    RCS Boration A-2 LDST Hatch area                1      OP/i/Ni1103/004 A !. 4.6    RCS Boratian CTT Rm                             I      OP/I/A/1103/004 A      4.6    RCS     Boration A-2-1i&2 Chem. Add Panel           1      OP/I/A/1103/004 A      4.6    RCS     Boration A-I-Cal Q70                        1      OP/i/Ni1103/004 A      4.7    RCS     Boration A-2-LDST Hatch area                I      OP/i/Ni1103/004 A      4.7    RCS     Boration A-i-Unit I CBAST Rm                1      OP/I/NI1103/004 A      4.7    RCS Boration Unit I BTP Rm                      1      OP/1/AI1104/027       4.19    BTP Recirc Unit I BTP Rm                      1      OP/1/A/1104/027       4.20. BTP Recirc 1      OP/IllA1102/010        4.2    Unit SD 1      PT/1/AN0600/001       13.2    Surv. Mode3 Unit l-BTP Rm                     1,2     OP1/A/N1103/004        4.5    Makeup Unit l-BTP Rm                     1,2     OP/ll/I1103/004        4.6    Makeup 1,2,3    OP/l/A/1102/010        4.3      SD Mode I to 3 3      OP/I/A/1102/010        4.4 RB 779', Cable Roam, 1UB2          3      IP/11/N0200/047               LTOP Calibration RB 779'                            3      IP/11AI0200/047               LTOP Calibration 1UB2                               3      IP/11A!02001047               LTOP Calibration 1AT7                               3      IP/1 /A0200/047               LTOP Calibration 1MTC-4                             3      IPI1/N0200/047                LTOP Calibration 1AT5                               3      IP/l/A/0200/047               LTOP Calibration LPI Coaler Room                    3      OP/I/A/1103/006       4.12 3      OP/1/A/1102/010        4.7   ________
OP/i/NI1104/012 A      4.2         CCW Pump 3  OP/I/A/1102/0I0        4.7 Unit 1 Primary Sample Hood                CP/1/A/2002/014        4.2 AB SAMPLE RM.308                          CP/11AN2002/014        4.2     _______
A-4-402 PZR Heaters                3      OP/I/Ni1102/010        4.7
[Document No.]                                   Rev. 0                           Page 237 of 240

ATTACHMENT 3 Safe Operation & Shutdown  Rooms/Areas Tables R-2 & H-2 Bases Equip Rrm XO/XP                    3    OP/1/A/1102/010          4.7 Equip Rm XO/XP                     3    OP/1/A/1104/049          4.2 A-4-402 PZR Heaters                3    OP/iI/A/i1104/049        4.2 3    OP/1/A/1104/001         4.14 A-4-409                            3    OP/1/A/1104/001          4.4 A-3-308                            3    OP/1/A/1104/001          4.4 A-2 Hallway                        3    OP/I/AI1104/001          4.4 R-1G-W                             3    OP/1/A/1104/001          4.4 R-1 -around "A" CFT                3  IOP/I/A/1104/001           4.4 R-B above Emer Sump                3    OP/i/Ni1104/001          4.4 R-B above RBNS                     3    OP/1/A/1104/001          4.4 _______
R-l-around "B" CFT                 3    OP/1/A/1104/001          4.4 R-B-20' above LD CIr RM            3    OP/i/Ni1104/001          4.4 Equip Rm XO/XP                     3    OP/1/N/1104/001         4.14 A-4-W Pent Rm                      3    OP/I/Ni1104/001         4.14 A-4-E Pent                         3    OP/i/Ni1104/049          4.2 R-3G East Side                     3    OP/i/Ni1104/049          4.2 A-4-402                            3    OP/i/Ni1104/049          4.2 LTOP Alignment A-2-Col. P-63)                     3    OP/i/Ni1104/049          4.2 LTOP Alignment T-3-Equip Rm)                      3    OP/i/Ni1104/049          4.2 LTOP Alignment 3    OPI1/A/1102/010          4.7 3    OP/11N1102/010          4.15 A-2-Unit 1 BAMT, in hallway        3    OP/1/A/1104/002         4.17 Turbine Building                   3    OP/1/A/1106/002 A       4.14 Turbine Building                   3    OP/0/AI1104/048          4.4         Step 3.7 OP/i/Ni1102/010          4.7 ________
Next actions---LPI LPI System Start-up (CR & SSF-CR)                                3    OP/1/A/1104/004          4.2 LPI Fill & S/U Valve lineup for AB 1st Floor                       3    OP/1/A/1104/004          4.5 LPI Breaker line up AB Pent. Rooms                     3    OP/1/AN1102/010          4.1 S/D Secondary Steam TB-3 &CR                           3    OP/I/A/1102/010              SD Align FDW clean-TB All Levels                      3    OP/i/A/1102/010          4.1 up AB-2                             4 &5   OP/I/A/1102/010         4.11 RCS H2 Sampling RB, AB-i, 2 & 3rd                  5    OP/I/A/1i03/008    _____RCS          Crud Burst I[Document No.]                                Rev. 0                            Page 238 of 240

ATTACHMENT 3 Safe Operation & Shutdown Rooms/Areas Tables R-2 & H-2 Bases Unit Shutdown Room List       Mode Turbine Building                      1,2,3 A-I hallway 8' S/ cal 65              1,2,3 A-i-hallway N of CoJ 82               1,2,3 A-i hallway 5' Stool 67               i1,2,3 A-i hallway col 82                    i1,2,3 A-I hallway 10' Stool 96)             1,2,3 A-l- BAMT Rm                             1 A-1iUniti1& 2BAMT Rm                     1 A-i-Col Q70                              i A-2 LOST Hatch area                   1,2,3 A-2-Unit 2 LOST Hatch area            1,2,3 A-2 Unit 3 LDST Hatch area            1,2,3 A-2-1&2 Chem. Add Panel                  1 A-2-CoI. P-63                            3 A-2-Unit i BAMT, in hallway              3 A-2 Hallway                              3 A-3-308                                  3 A-4-402                                  3 A-4-409                                  3 A-4-W Pent Rm                            3 A-4-E Pent                               3 Unit I BTP Rm                         1,2,3 Unit 2 BTP Rm                         1,2,3 Unit 3 BTP Rm                         1,2,3 Ui LPI Cooler Rm                      1,2,3 RB 779', Cable Room, iUB2                3 RB 779'                                  3 R-1G-W                                   3 R-il-around "A" CFT                      3 R-i-around "B"CFT                        3 R-B above Emer. Sum p                    3 R-B above RBNS                           3 R-B-20' above LD Cooler RM               3 R-3G East Side                           3 I1UB2                                    3 I1AT7                                   3 1MTC-4                                  3 I1AT5                                   3 Unit I Primary Sample Hood               3 AB SAMPLE RM.308                         3 RB, AB                                4&5
[Document No.]                              Rev. 0                 Page 239 of 240

ATTACHMENT 3 Safe Operation & Shutdown Rooms/Areas Tables R-2 & H-2 Bases Table R-2 & H-2 Results Table R-2 & H-2    Safe Operation & Shutdown RoomslAreas Room/Area                       Mode Applicability Turbine Building                                               1,2, 3 Equipment and Cable Rooms                                      1,2, 3 Auxiliary Building                                         1, 2, 3, 4, 5 Reactor Buildings                                            '3, 4, 5 S[Document No.]                           Rev. 0                        Page 240 of 240}}