ML11159A144

Response to Request for Additional Information Regarding Emergency Action Level Scheme Change

ML11159A144
Person / Time
Site:	Columbia
Issue date:	06/01/2011
From:	Sawatzke B Energy Northwest
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
G02-11-100
Download: ML11159A144 (282)

Text

t NORTHW ESTBradley J. Sawatzke ENERGY Columbia Generating Station P.O. Box 968, PE08 Richland, WA 99352-0968 Ph. 509.377.4300 1 F. 509.377.4150 bjsawatzke @energy-northwest.com June 1, 2011 G02-11-100 10 CFR 50, Appendix E IV.B(1)

U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001

Subject:

COLUMBIA GENERATING STATION, DOCKET NO. 50-397 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION REGARDING EMERGENCY ACTION LEVEL SCHEME CHANGE

References:

1) Letter G02-10-121 dated August 25, 2010, DK Atkinson (Energy Northwest) to NRC, "Emergency Action Level Scheme Change to NEI 99-01, Rev. 5 Format'

2) Email dated April 26, 2011, from DA Johnson (NRC) to LL Williams (Energy Northwest), "Draft Request for Additional Information Columbia Generating Station Emergency Action Level Scheme Change to NEI 99-01, Revision 5"

Dear Sir or Madam:

By Reference 1, Energy Northwest requested approval of an Emergency Action Level (EAL) scheme change to adopt the NEI 99-01, Revision 5 format. Via Reference 2, the NRC requested additional information related to the Energy Northwest submittal.

The Energy Northwest response to the Reference 2 request for additional information (RAI) is provided in the enclosure to this letter. The enclosure contains the following attachment, which supersedes the original Attachment 2 submitted as part of Reference 1:

• Attachment- Proposed PPM 13.1.1A, Classifying the Emergency - Technical Bases There are no new commitments contained in this response. Ifyou have any questions or require additional information, please contact LL Williams, (Acting) Licensing Supervisor, at (509) 377-8148.

RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION REGARDING EMERGENCY ACTION LEVEL SCHEME CHANGE Page 2 of 2 Respectfully, Vice President, Nuclear Generation & Chief Nuclear Officer cc: NRC Region IVAdministrator NRC NRR Project Manager NRC Senior Resident Inspector/988C RN Sherman - BPA/1 399 WA Horin - Winston & Strawn

RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION REGARDING EMERGENCY ACTION LEVEL SCHEME CHANGE Page 1 of 5 Response to Request for Additional Information Request for Additional Information (RAI):

By letter dated August 25, 2010, Energy Northwest - Columbia Generating Station (CGS),

(Agencywide Documents Access and Management System (ADAMS) Accession No. ML102430334) requested prior approval for a revised emergency action level (EAL) scheme for Columbia Generating Station.

CGS's letter stated that the current EAL scheme is based on generic development guidance from NUMARC/NESP-007, "Methodology for Development of Emergency Action Levels,"

Revision 2, dated January 1992, (ADAMS Accession No. ML041120174). Since 1992, numerous enhancements and clarification efforts have been made to the generic EAL development guidance resulting in the most latest document, Nuclear Energy Institute (NEI) 99-01, Revision 5, "Methodology for Development of Emergency Action Levels," (ADAMS Accession No. ML080450149), which was found to be acceptable for use as generic EAL development guidance by the Nuclear Regulatory Commission (NRC) staff by letter dated February 22, 2008 (ADAMS Accession No. ML080430535).

The proposed EAL scheme was developed using the generic development guidance from NEI 99-01, Revision 5 with numerous differences and deviations based upon design criteria applicable to CGS as well as licensee preferences for terminology, format, and other licensee desired modifications to the generic EAL scheme provided in NEI 99-01 Revision 5.

Attached are the draft requests for additional information (RAIs) to facilitate the technical review being conducted by the Operating Reactor Licensing and Outreach Branch staff.

Timely and accurate response to these draft RAIs is requested.

NRC Request:

1. Section 2.0, "Discussion." This section does not contain information on the treatment of multiple events and classification upgrading as outlined in the endorsed guidance.

Please provide justification for this inconsistency, or revise accordingly.

Energy Northwest Response:

Section 2.0 of the attachment to this enclosure (Proposed PPM 13.1.1A, Classifying the Emergency - Technical Bases) has been revised to add a new section 2.17 'Treatment of Multiple Events and Classification Level Upgrading" consistent with generic Section 3.10 of NEI 99-01.

RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION REGARDING EMERGENCY ACTION LEVEL SCHEME CHANGE Page 2 of 5 NRC Request:

2. Step 2.13 'Validation of Indications, Reports and Conditions," step 2.13.2, states, in part, that "Ifa meter is broke, compensatory measures should be established to monitor the affected parameter." Please explain why this implies an option versus being a requirement as stated in 10 CFR 50.54(q), or revise to state "...shall...."

Energy Northwest Response:

Step 2.13.2 was revised to replace the word "should" with "shall". See the attachment to this enclosure.

NRC Request:

3. The use of the following abbreviations is considered unadvisable from a human factors perspective as their use could cause significant confusion. Please explain why the staff should consider allowing these abbreviations to be implemented, or revise to use generally accepted nomenclature.

a. GE: Numerous EALs use this as an abbreviation for "greater than or equal to."

Please explain, in detail, why the use of "':" or spelling out the words will not suffice, or revise accordingly wherever used.

b. GT: Numerous EALs use this as an abbreviation for "greater than." Please explain, in detail, why the use of ">" or spelling out the words will not suffice, or revise accordingly wherever used.

c. LE: Numerous EALs use this as an abbreviation for "less than or equal to."

Please explain, in detail, why the use of "-' or spelling out the words will not suffice, or revise accordingly wherever used.

d. LT: Numerous EALs use this as an abbreviation for "less than." Please explain, in detail, why the use of "<" or spelling out the words will not suffice, or revise accordingly wherever used.

Energy Northwest Response:

Energy Northwest requests the NRC staff allow the above abbreviations to be implemented for CGS. The use of the acronyms GE, GT, LE and LT have long been established as the CGS approved designators for the terms 'greater than or equal to,' 'greater than,' 'less than or equal to,' and 'less than,' respectively, in accordance with the Energy Northwest Glossary. These writing criteria for expressing relational operators are used throughout the CGS procedure network including station operating procedures (SOPs), abnormal operating procedures (AOPs), emergency operating procedures (EOPs), the Emergency Plan and existing EALs. To take an exception to this requirement only for the implementation of the

RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION REGARDING EMERGENCY ACTION LEVEL SCHEME CHANGE Page 3 of 5 EALs would pose a much greater human factors concern than any potential confusion on the part of an individual not familiar with this approved standard.

NRC Request:

4. EAL RS1.1 and RG1.1: Please explain why the CGS basis information contains information about the Radwaste Building Exhaust (WEA-RIS-14(A)) being applicable when Table R-1 shows these to not be applicable to these EALs.

Energy Northwest Response:

The specified paragraph was unintentionally included in the bases discussion. This issue has been entered into the Energy Northwest corrective action program. The specified paragraph has been deleted from the bases of RS1.1 and RG1.1. See the attachment to this enclosure.

NRC Request:

5. EAL RG1.3: Please explain why the note related to EAL declaration timing was not incorporated in this EAL, or revise accordingly.

Energy Northwest Response:

The note related to EAL declaration timing has been incorporated into EALs RS1.3 and RG1.3. See the attachment to this enclosure.

NRC Request:

6. EAL RA3.1: Please explain, in more detail, why the Central Alarm Station (CAS) is not applicable to this EAL or revise accordingly.

Energy Northwest Response:

The secondary alarm station (SAS) is contained entirely within the Control Room envelope, provides a redundant capability to the CAS, and is continuously manned. Therefore, CAS is not required for continuous occupancy to maintain plant safety functions. With alarm station functions fulfilled by SAS, the CGS security force might not stay in the "cloud" in CAS but, rather, may establish a larger perimeter to accommodate site security requirements. In order to clarify the intent of this EAL, EAL RA3.1 has been expanded to state, "Dose rates GT 15 mR/hr within the Control Room envelope (D21 -S004 CH-1 9)." See the attachment to this enclosure.

NRC Request:

7. EAL CU5.1 and SU6.1: Please explain how the State/County Notification (CRASH)

System will notify the NRC of a declared event at CGS or revise accordingly.

RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION REGARDING EMERGENCY ACTION LEVEL SCHEME CHANGE Page 4 of 5 Energv Northwest Response:

The inclusion of the CRASH System was incorrect. This issue has been entered into the Energy Northwest corrective action program. The CRASH System has been deleted from Communications Tables S-2 and C-4 and associated bases discussions. See the attachment to this enclosure.

NRC Request:

8. Fission Barrier Matrix: Please explain why no "other" site specific indicators have not been developed for this matrix, particularly when the current EAL scheme (based upon NUMARC/NESP-007) has several indicators not carried over into this proposed matrix considering that the matrix from NEI 99-01 Revision 5 is very similar.

Energy Northwest Response:

The process used to generate the fission product barrier thresholds included collection of site specific input data with consideration for other potential fission product barrier indicators not specifically identified within the generic NEI 99-01 Rev. 5 guidance. Multiple drafts and site technical inter-disciplinary reviews were conducted. A formal verification and validation was performed using actual operating crews and a formal review by the Plant Operations Committee was completed; all of which failed to identify any other plant design specific applicable thresholds for inclusion in the Table F-1 criteria. For example, abnormally low suppression pool water level was considered a potential loss of primary containment (PC) but was determined to be adequately addressed by the family of curves in the heat capacity temperature limit (HCTL) in the CGS EOPs. Elevated drywell temperature and wetwell pressure above pressure suppression pressure (PSP) were considered as a potential loss of PC but these conditions would likely not occur without either a station blackout (SBO),

which is addressed by System Malfunction EALs, or a loss of coolant accident (LOCA) condition, which is adequately addressed by other loss/potential loss fission product barrier thresholds.

The existing CGS NUMARC/NESP-007 based EALs do include a number of Containment barrier potential loss thresholds that do not correspond to NEI 99-01 Rev. 5 equivalent thresholds (e.g., loss of pressure suppression function, wetwell pressure exceeds PSP).

These thresholds were evaluated for possible incorporation into the NEI 99-01 Rev. 5 based fission product barrier matrix. However, they were deemed to either not meet the threshold definition of barrier potential loss as intended by the generic NEI 99-01 scheme or are considered redundant to existing threshold criteria. Additionally, they were deemed not to be conditions that are unique, plant-specific design-driven considerations. These conditions apply to all boiling water reactors (BWRs) and, if worthy of thresholds requiring emergency classification, would have been included in the generic scheme.

RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION REGARDING EMERGENCY ACTION LEVEL SCHEME CHANGE Page 5 of 5 NRC Request:

9. The staff approved License Amendment No. 218 for CGS in response to the licensee's application dated April 28, 2010. However, the staff finds no evidence that this amendment has been carried over into this request to revise the CGS EAL scheme. Please explain, in detail, why this license amendment has not been carried over into this EAL scheme change request, or revise accordingly.

Energy Northwest Response:

License Amendment No. 218 was approved November 3, 2010, which was after CGS requested revision of the EAL scheme. It had been the plan of Energy Northwest to incorporate License Amendment No. 218 into the Final NEI 99-01 Rev. 5 EAL scheme. This has now been accomplished. The attachment to this enclosure has been revised to reflect the approved 618-11 area EALs as outlined in License Amendment No.218 as new EALs HU7.1 and HA7.1.

NRC Request:

10. EAL EUI.1: The EAL states, "Multi Purpose Container (MPC) CONFINEMENT BOUNDARY breach". As worded this EAL would require an actual breach of the boundary versus damage to the confinement boundary for EAL declaration. Please provide justification for this inconsistency, or revise accordingly.

Energy Northwest Response:

EAL EU1.1 has been revised to read: "Damage to a loaded Multi Purpose Container (MPC)

CONFINEMENT BOUNDARY." See the attachment to this enclosure.

RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION REGARDING EMERGENCY ACTION LEVEL SCHEME CHANGE Attachment Proposed PPM 13.1.1 A, Classifying the Emergency - Technical Bases 274 Pages

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 1 of 274 ENERGY NORTHWEST COLUMBIA GENERATING STATION PLANT PROCEDURES MANUAL 4.

13.1.1A Effective Date:

DIC # 1308.1 PCN # (Ifapplicable) N/A QPR: David B Holmes 8687 First MI Last Name Ext. #

Sponsor: Timothy J Powell 4161 First Ml Last Name Ext. #

All review and approval signatures are documented on the Procedure Revision Form Synopsis Procedure is revised to incorporate the changes as described by NEI 99-01 R5, Methodology For Development of Emergency Action Levels. This is a major change to the previous revision, so no revision bars were used.

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 2 of 274 TABLE OF CONTENTS Page 1.0 PURPOSE ............................................................................................................................... 3 2.0 DISCUSSION .......................................................................................................................... 3 2.1 Background ............................................................................................................................. 3 2.2 Fission Product Barriers ................................................................................................... 4 2.3 Emergency Classification Based on Fission Product Barrier Degradation ............................ 4 2.4 EAL Relationship to EO Ps .................................................................................................... 5 2.5 Sym ptom-Based vs. Event-Based Approach ....................................................................... 5 2.6 EAL Organization ..................................................................................................................... 5 2.7 Technical Bases Inform ation ................................................................................................ 7 2.8 Mode Applicability .................................................................................................................... 8 2.9 NEI 99-01 Basis: ...................................................................................................................... 8 2.10 CGS Basis: .............................................................................................................................. 8 2.11 CGS Basis Reference(s): .................................................................................................... 8 2.12 Operating Mode Applicability ............................................................................................... 8 2.13 Validation of Indications, Reports and Conditions ................................................................. 9 2.14 Planned vs. Unplanned Events ........................................................................................... 9 2.15 Classifying Transient Events .............................................................................................. 10 2.16 Im m inent EAL Thresholds .................................................................................................. 10

3.0 REFERENCES

...................................................................................................................... 11 3.1 Developm ental ....................................................................................................................... 11 3.2 Im plem enting ......................................................................................................................... 11 3.3 Com m itm ents ........................................................................................................................ 11 4.0 DEFINITIONS ........................................................................................................................ 11 5.0 ABBREVIATIO NS/ACRONYMS ......................................................................................... 15 6.0 CGS-TO-NEI 99-01 EAL CROSSREFERENCE ................................................................. 18 7.0 ATTACHM ENTS .................................................................................................................... 22 7.1 Attachm ent 1 - EAL Bases .................................................................................................. 22 Category R Abnormal Rad Release / Rad Effluent ......................................................... 23 Category C Cold Shutdown / Refueling System Malfunction ........................................... 63 Category H Hazards & Other Conditions Affecting Plant Safety ....................................... 115 Category S System Malfunction ....................................................................................... 168 Category E ISFSI ............................................................................................................. 216 Category F Fission Product Barrier Degradation .............................................................. 218 7.2 Attachment 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases ........................ 218

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 3 of 274 1.0 PURPOSE This document provides an explanation and rationale for each Emergency Action Level (EAL) included in the EAL Upgrade Project for Columbia Generating Station (CGS). It should be used to facilitate review of the CGS EALs, provide historical documentation for future reference and serve as a training aid. Decision-makers responsible for implementation of PPM 13.1.1, Classifying the Emergency, may use this document as a technical reference in support of EAL interpretation.

The expectation is that emergency classifications are to be made as soon as conditions are present and recognizable for the classification, but within 15 minutes or less in all cases of conditions present. Use of this document for assistance is not intended to delay the emergency classification.

This document is controlled pursuant to 10 CFR 50.54(q).

2.0 DISCUSSION 2.1 Background 2.1.1 EALs are the plant-specific indications, conditions or instrument readings that are utilized to classify emergency conditions defined in the CGS Emergency Plan.

2.1.2 In 1992, the NRC endorsed NUMARC/NESP-007 "Methodology for Development of Emergency Action Levels" as an alternative to NUREG-0654 EAL guidance.

2.1.3 NEI 99-01 (NUMARC/NESP-007) Revision 5 represents the most recently accepted methodology. 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 (ISFSls).

• Simplifying the fission product barrier EAL threshold for a Site Area Emergency.

2.1.4 Using NEI 99-01 Rev. 5, CGS conducted an EAL implementation upgrade project that produced the EALs discussed herein.

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 4 of 274 2.2 Fission Product Barriers Many of the EALs derived from the NEI methodology are fission product barrier based. That is, the conditions that define the EALs are based upon loss 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. "Loss" means the barrier no longer assures containment of radioactive materials; "potential loss" infers an increased probability of barrier loss and decreased certainty of maintaining the barrier.

2.2.1 The primary fission product barriers are:

a. Fuel Clad (FC): Zircalloy tubes which house the ceramic uranium oxide pellets along with the end plugs which are welded into each end of the fuel rods comprise the FC barrier.

b. Reactor Coolant System (RCS): The reactor vessel shell, vessel head, CRD housings, vessel nozzles and penetrations, and all primary systems directly connected to the RPV up to the outermost primary containment isolation valve comprise the RCS barrier.

c. Containment (PC): The drywell, the suppression chamber/pool, their respective interconnecting paths, and other connections up to and including the outermost containment isolation valves comprise the PC barrier.

2.3 Emergency Classification Based on Fission Product Barrier Degradation The following criteria are the bases for event classification related to fission product barrier loss or potential loss:

2.3.1 Unusual Event:

Any loss or any potential loss of Containment

2.3.2 Alert

Any loss or any potential loss of either Fuel Clad or RCS 2.3.3 Site Area Emergency:

Loss or potential loss of any two barriers 2.3.4 General Emergency:

Loss of any two barriers and loss or potential loss of third barrier

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft IMinor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 5 of 274 2.4 EAL Relationship to EOPs Where possible, the EALs have been made consistent with and utilize the conditions defined in the CGS Emergency Operating Procedures (EOPs). While the symptoms that drive operator actions specified in the EOPs are not indicative of all possible conditions which warrant emergency classification, they define the symptoms, independent of initiating events, for which reactor plant safety and/or fission product barrier integrity are threatened. When these symptoms are clearly representative of one of the NEI Initiating Conditions, they have been utilized as an EAL. This permits rapid classification of emergency situations based on plant conditions without the need for additional evaluation or event diagnosis. Although some of the EALs presented here are based on conditions defined in the EOPs, classification of emergencies using these EALs is not dependent upon EOP entry or execution. The EALs can be utilized independently or in conjunction with the EOPs.

2.5 Symptom-Based vs. Event-Based Approach To the extent possible, the EALs are symptom-based. That is, the action level threshold is defined by values of key plant operating parameters that identify emergency or potential emergency conditions. This approach is appropriate because it allows the full scope of variations in the types of events to be classified as emergencies. However, a purely symptom-based approach is not sufficient to address all events for which emergency classification is appropriate. Particular events to which no predetermined symptoms can be ascribed have also been utilized as EALs since they may be indicative of potentially more serious conditions not yet fully realized.

2.6 EAL Organization 2.6.1 The CGS EAL scheme includes the following features:

a. Division of the EAL set into three broad groups:

1) EALs applicable under all plant operating modes - This group would be reviewed by the EAL-user any time emergency classification is considered.

2) EALs applicable only under hot operating modes - This group would only be reviewed by the EAL-user when the plant is in Hot Shutdown, Startup or Power Operations mode.

3) EALs applicable only under cold operating modes - This group would only be reviewed by the EAL-user when the plant is in Cold Shutdown, Refuel or Defueled mode.

2.6.2 The purpose of the groups is to avoid review of hot condition EALs when the plant is in a cold condition and avoid review of cold condition EALs when the plant is in a hot condition. This approach significantly minimizes the total number of EALs that must be reviewed by the EAL-user for a given plant condition, reduces EAL-user reading burden and, thereby, speeds identification of the EAL that applies to the emergency.

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 6 of 274 2.6.3 Within each of the above three groups, assignment of EALs to categories/subcategories - Category and subcategory titles are selected to represent conditions that are operationally significant to the EAL-user.

Subcategories are used as necessary to further divide the EALs of a category into logical sets of possible emergency classification thresholds. The CGS EAL categories/subcategories and their relationship to NEI Recognition Categories are listed below.

EAL Groups, Categories and Subcategories EAL Group/Category EAL Subcategory Any Operating Mode:

R - Abnormal Rad Release / Rad 1 - Offsite Rad Conditions Effluent 2 - Onsite Rad Conditions/Fuel Pool Events 3 - CR Rad H - Hazards & Other Conditions 1- Natural & Destructive Phenomena Affecting Plant Safety 2- Fire or Explosion 3- Hazardous Gas 4- Security 5- Control Room Evacuation 6- Judgment 7- 618-11 Area E- ISFSI None Hot Conditions:

S - System Malfunction 1 - Loss of AC Power 2 - Loss of DC Power 3 - ATWS / Criticality 4 - Inability to Reach or Maintain Shutdown Conditions 5 - Instrumentation 6 - Communication 7 - Fuel Clad Degradation 8 - RCS Leakage F - Fission Product Barrier None Degradation Cold Conditions:

C - Cold Shutdown / Refuel System 1- Loss of AC Power Malfunction 2- Loss of DC Power 3- RPV Level 4- RCS Temperature 5- Communications

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 7 of 274 6 - Inadvertent Criticality 2.6.4 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 Sections 2.7 and 2.8, and Attachments 1 and 2 of this document for such information.

2.7 Technical Bases Information 2.7.1 EAL technical bases are provided in Attachment 1 for each EAL according to EAL group (Any, Hot, Cold), EAL category (R, C, E, H, S 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:

a. Category Letter & Title

b. Subcategory Number & Title

c. Initiating Condition (IC) 2.7.2 Site-specific description of the generic IC given in NEI 99-01.

a. EAL Identifier (enclosed in rectangle)

1) 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:

a) First character (letter): Corresponds to the EAL category as described above (R, C, E, H, S or F) b) Second character (letter): The emergency classification (G, S, A or U) c) 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).

d) 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).

2) Classification (enclosed in rectangle):

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 8 of 274 Unusual Event (U), Alert (A), Site Area Emergency (S) or General Emergency (G)

3) EAL (enclosed in rectangle)

Exact wording of the EAL as it appears in the EAL classification matrix 2.8 Mode Applicability One or more of the following plant operating conditions comprise the mode to which each EAL is applicable: 1 - Power Operations, 2 - Startup, 3 - Hot Shutdown, 4 - Cold Shutdown, 5 -

Refuel, D - Defueled, N/A - Not Applicable or All. (See Section 2.12 for operating mode definitions.)

2.9 NEI 99-01 Basis:

The basis discussion applicable to the EAL taken from NEI 99-01.

2.10 CGS Basis:

Description of the site-specific rationale for the EAL 2.11 CGS Basis Reference(s):

Site-specific source documentation from which the EAL is derived 2.12 Operating Mode Applicability 2.12.1 Power Operations Reactor mode switch is in RUN 2.12.2 Startup The mode switch is in STARTUP/HOT STANDBY or REFUEL with all reactor vessel head closure bolts fully tensioned 2.12.3 Hot Shutdown The mode switch is in SHUTDOWN, with all reactor vessel head closure bolts fully tensioned, and reactor coolant temperature is GT 200°F 2.12.4 Cold Shutdown The mode switch is in SHUTDOWN, all reactor vessel head closure bolts are fully tensioned, and reactor coolant temperature is LE 200OF 2.12.5 Refuel

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 9 of 274 The mode switch is in REFUEL or SHUTDOWN and one or more reactor vessel head closure bolts less than fully tensioned 2.12.6 Defueled All reactor fuel removed from RPV. (Full core off load during refueling or extended outage).

2.12.7 The plant operating mode that exists at the time that the event occurs (prior to any protective system or operator action is 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.

2.12.8 For events that occur in Cold Shutdown or Refuel, escalation is via EALs that have Cold Shutdown or Refuel for mode applicability, even if Hot Shutdown (or a higher mode) is entered during any subsequent heat-up. In particular, the fission product barrier EALs are applicable only to events that initiate in Hot Shutdown or higher.

2.13 Validation of Indications, Reports and Conditions 2.13.1 All emergency classifications shall be based upon valid indications, reports or conditions. 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.

2.13.2 Broken or unavailable instrumentation does not necessarily mean an EAL is met.

The actual plant condition referenced in the EAL must be present. If a meter is broke, compensatory measures shall be established to monitor the affected parameter. Generally, instrumentation out of service for maintenance does not need compensatory measures established unless out of service for an extended period of time; the time required to complete PMs and calibrations is not considered to be extended periods of time. Other work such as a system outage or replacement of a broken instrument would require compensatory measures.

2.14 Planned vs. Unplanned Events 2.14.1 Planned evolutions involve preplanning to address the limitations imposed by the condition, the performance of required surveillance testing, and the implementation of specific controls prior to knowingly entering the condition in accordance with the specific requirements of the Columbia Technical Specifications. Activities which cause the site to operate beyond that allowed by the Columbia Technical Specifications, planned or unplanned, may result in an EAL threshold being met or exceeded. Planned evolutions to test, manipulate, repair, perform maintenance or modifications to systems and equipment that result in an EAL value being met or

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 10 of 274 exceeded are not subject to classification and activation requirements as long as the evolution proceeds as planned and is within the operational limitations imposed by the specific operating license. However, these conditions may be subject to the reporting requirements of 10 CFR 50.72.

2.15 Classifying Transient Events 2.15.1 For some events, the condition may be corrected before a declaration has been made. The key consideration in this situation is to determine whether or not further plant damage occurred while the corrective actions were being taken. In some situations, this can be readily determined, in other situations, further analyses (e.g.,

coolant radiochemistry sampling) may be necessary. Classify the event as indicated and terminate the emergency once assessment shows that there were no consequences from the event and other termination criteria are met.

2.15.2 Guidance for classifying transient events addresses the period of time of event recognition and classification (15 minutes). However, in cases when EAL declaration criteria may be met momentarily during the normal expected response of the plant, declaration requirements should not be considered to be met when the conditions are a part of the designed plant response, or result from appropriate Operator actions.

2.15.3 There may be cases in which a plant condition that exceeded an EAL was not recognized at the time of occurrence but is identified well after the condition has occurred (e.g., as a result of routine log or record review), and the condition no longer exists. In these cases, an emergency should not be declared.

2.15.4 Reporting requirements of 10 CFR 50.72 are applicable and the guidance of NUREG-1022 should be applied.

2.16 Imminent EAL Thresholds 2.16.1 Although the majority of the EALs provide very specific thresholds, the Emergency Director must remain alert to events or conditions that lead to the conclusion that exceeding the EAL threshold is imminent. If, in the judgment of the Emergency Director, an imminent situation is at hand, the classification should be made as if the threshold has been exceeded. While this is particularly prudent at the higher emergency classes (the early classification may permit more effective implementation of protective measures), it is nonetheless applicable to all emergency classification levels.

2.16.2 Additionally, for those EAL thresholds with a time criteria, in the absence of information to the contrary, assume that the applicable time criteria has already been exceeded if the length of time which the EAL threshold conditions have existed is unknown.

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CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 11 of 274 2.17 Treatment of Multiple Events and Classification Level Upgrading 2.17.1 When multiple simultaneous events occur, the emergency classification level is based on the highest EAL reached. For example, two Alerts remain in the Alert category. Or, an Alert and a Site Area Emergency is a Site Area Emergency.

2.17.2 Although the majority of the EALs provide very specific thresholds, the Emergency Director must remain alert to events or conditions that lead to the conclusion that exceeding the EAL is imminent. If, in the judgment of the Emergency Director, an imminent situation is at hand, the classification should be made as if the threshold has been exceeded. While this is particularly prudent at the higher emergency classification levels (as the early classification may provide for more effective implementation of protective measures), it is nonetheless applicable to all emergency classification levels.

3.0 REFERENCES

3.1 NEI 99-01 Final Draft Revision 5, Methodology for Development of Emergency Action Levels, Final Draft, February 2008 (ADAMS Accession Number of ML080450149) 3.2 NRC Regulatory Issue Summary (RIS) 2003-18, Supplement 2, Use of Nuclear Energy Institute (NEI) 99-01, Methodology for Development of Emergency Action Levels Revision 4, Dated January 2003 (December 12, 2005) 3.3 SWP-PRO-03, Procedure Writer's Manual 3.4 PPM 13.1.1, Classifying the Emergency 3.5 EAL Comparison Matrix 3.6 PPM 13.1.1B, EAL Hot Matrix 3.7 PPM 13.1.1C, EAL Cold Matrix 4.0 DEFINITIONS 4.1 Affecting Safe Shutdown - Event in progress has adversely affected functions that are necessary to bring the plant to and maintain it in the applicable hot or cold shutdown condition.

Plant condition applicability is determined by Technical Specification LCOs in effect.

4.1.1 Example 1: Event causes damage that results in entry into an LCO that requires the plant to be placed in hot shutdown. Hot shutdown is achievable, but cold shutdown is not. This event is not "Affecting Safe Shutdown."

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CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 12 of 274 4.1.2 Example 2: Event causes damage that results in entry into an LCO that requires the plant to be placed in cold shutdown. Hot shutdown is achievable, but cold shutdown is not. This event is "Affecting Safe Shutdown."

4.2 Aircraft - Includes both small and large aircraft. Examples of aircraft include general aviation Cessna, Piper and Lear types of private planes; large passenger or freight planes; as well as police, medical and media helicopters. A large aircraft is referred to as an airliner.

4.3 Airliner - An aircraft with the potential for causing significant damage to the plant 4.4 Alert - Events are in progress or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant or a security event that involves probable life threatening risk to site personnel or damage to site equipment because of hostile action. Any releases are expected to be limited to small fractions of the EPA PAG exposure levels.

4.5 Bomb - Refers to an explosive device suspected of having sufficient force to damage plant systems or structures.

4.6 Civil Disturbance - A group of persons violently protesting station operations or activities at the site.

4.7 Confinement Boundary - Is the barrier(s) between areas containing radioactive substances and the environment. As related to the CGS ISFSI, Confinement Boundary is defined as the Multi Purpose Container (MPC).

4.8 Containment Closure - The site specific procedurally defined actions taken to secure containment (primary or secondary) and its associated structures, systems, and components as a functional barrier to fission product release under existing plant conditions.

4.9 Credible - As defined in the Physical Security Plan, Credible is information received from a source determined to be reliable (such as Law Enforcement, government agency, etc.) or has been verified to be true.

4.10 Emergency Action Level (EAL) - A pre determined, site specific, observable threshold for a plant IC that places the plant in a given emergency classification level. An EAL can be: an instrument reading; an equipment status indicator; a measurable parameter (on-site or off-site); a discrete, observable event; results of analyses; entry into specific emergency operating procedures; or another phenomenon which, if it occurs, indicates entry into a particular emergency classification level.

4.11 Explosion - A rapid, violent, unconfined combustion, or catastrophic failure of pressurized/energized equipment that imparts energy of sufficient force to potentially damage permanent structures, systems, or components.

4.12 Extortion - Is an attempt to cause an action at the station by threat of force.

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CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 13 of 274 4.13 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.

4.14 General Emergency - Events are in progress or have occurred which involve actual or imminent substantial core degradation or melting with potential for loss of containment integrity or hostile action that results in an actual loss of physical control of the facility. Releases can be reasonably expected to exceed EPA PAG exposure levels off-site for more than the immediate site area.

4.15 Hostaae - Person(s) held as leverage against the station to ensure that demands will be met by the station.

4.16 Hostile Action - An act toward CGS or its personnel that includes the use of violent force to destroy equipment, take hostages, and/or intimidate Energy Northwest 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 CGS. Non-terrorism-based EALs should be used to address such activities, (e.g., violent acts between individuals in the Security Defined Owner Controlled Area) 4.17 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.

4.18 Imminent - Mitigation actions have been ineffective, additional actions are not expected to be successful, and trended information indicates that the event or condition will occur. Where imminent timeframes are specified, they shall apply.

4.19 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.

4.20 Initiating Condition (IC) - One of a predetermined subset of nuclear power plant conditions where either the potential exists for a radiological emergency, or such an emergency has occurred.

4.21 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.

4.22 Normal Plant Operations - Activities at the plant site associated with routine testing, maintenance, or equipment operations, in accordance with normal operating or administrative procedures. Entry into abnormal or emergency operating procedures, or deviation from normal security or radiological controls posture, is a departure from Normal Plant Operations.

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CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 14 of 274 4.23 Projectile - An object directed toward a NPP that could cause concern for its continued operability, reliability, or personnel safety.

4.24 Protected Area - An area which normally encompasses all controlled areas within the security protected area fence as depicted in CGS Graphics Plant Drawing 902118-P.

4.25 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).

4.26 Sabotage - Deliberate damage, misalignment, or mis-operation of plant equipment with the intent to render the equipment inoperable. Equipment found tampered with or damaged due to malicious mischief may not meet the definition of Sabotage until this determination is made by security supervision.

4.27 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.

4.28 Security Defined Owner Control Area - An area determined to be of strategic or economic value for the protection of the Columbia Generating Station, surrounding the protected area barrier and controlled by the Columbia Generating Station Security Force, it provides a "land mass barrier" that serves as an obstacle to intruders.

4.29 Significant Transient - An unplanned event involving any of the following:

• Runback GT 25% thermal power

• Electrical load rejection GT 25% full electrical load

• Reactor scram

• ECCS injection

• Thermal power oscillations GT 10%

4.30 Site Area Emergency - Events are in progress or have occurred which involve actual or likely major failures of plant functions needed for protection of the public or hostile action that results in intentional damage or malicious acts; 1) toward site personnel or equipment that could lead to the likely failure of or; 2) that prevent effective access to, equipment needed for the protection of the public. Any releases are not expected to result in exposure levels which exceed EPA PAG exposure levels beyond the site boundary.

4.31 Strike Action - A work stoppage within the Protected Area by a body of workers to enforce compliance with demands made on CGS. The strike action must threaten to interrupt normal plant operations.

4.32 Unisolable - A breach or leak that cannot be promptly isolated.

4.33 Unplanned - A parameter change or an event that is not the result of an intended evolution and requires corrective or mitigative actions.

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CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 15 of 274 4.34 Unusual Event - Events are in progress or have occurred which indicate a potential degradation of the level of safety of the plant or indicate a security threat to facility protection has been initiated. No releases of radioactive material requiring off-site response or monitoring are expected unless further degradation of safety systems occurs.

4.35 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.

4.36 Visible Damage - Damage to equipment or structure that is readily observable without measurements, testing, or analysis. Damage is sufficient to cause concern regarding the continued operability or reliability of affected safety structure, system, or component. Example damage includes: deformation due to heat or impact, denting, penetration, rupture, cracking, paint blistering. Surface blemishes (e.g., paint chipping, scratches) should not be included.

4.37 Vital Area - Typically any site specific areas, normally within the Protected Area, that contains equipment, systems, components, or material, the failure, destruction, or release of which could directly or indirectly endanger the public health and safety by exposure to radiation.

5.0 ABBREVIATIONS/ACRONYMS AC Alternating Current APRM Average Power Range Meter ATWS Anticipated Transient Without Scram BWR Boiling Water Reactor CDE Committed Dose Equivalent CFR Code of Federal Regulations cpm counts per minute cps counts per second DC Direct Current EAL Emergency Action Level ECCS Emergency Core Cooling System ECL Emergency Classification Level EOF Emergency Operations Facility EOP Emergency Operating Procedure EPA Environmental Protection Agency

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CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 16 of 274 EPG Emergency Procedure Guideline EPIP Emergency Plan Implementing Procedure ESF Engineered Safety Feature FAA Federal Aviation Administration FBI Federal Bureau of Investigation FEMA Federal Emergency Management Agency FSAR Final Safety Analysis Report GDS Graphic Display System GE General Emergency, Greater than or Equal to gm Gram GT Greater Than HPCS High Pressure Core Spray HOO NRC Headquarters Operations Officer IC Initiating Condition IDLH Immediately Dangerous to Life and Health IPEEE Individual Plant Examination of External Events (Generic Letter 88-20)

ISFSI Independent Spent Fuel Storage Installation K*f Effective Neutron Multiplication Factor LCO Limiting Condition of Operation LE Less than or Equal to LER Licensee Event Report LFL Lower Flammability Limit LOCA Loss of Coolant Accident LPCS Low Pressure Core Spray LT Less Than LWR Light Water Reactor pCi Micro Curie MSCRWL Minimum Steam Cooling RPV Water Level MSCP Minimum Steam Cooling Pressure MSIV Main Steam Isolation Valve

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CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 17 of 274 MSL Main Steam Line mR milliRoentgen MW Megawatt NEI Nuclear Energy Institute NESP National Environmental Studies Project NORAD North American Aerospace Defense Command NPP Nuclear Power Plant NRC Nuclear Regulatory Commission NSSS Nuclear Steam Supply System NOUE Notification Of Unusual Event OBE Operating Basis Earthquake OCA Owner Controlled Area ODCM Off-site Dose Calculation Manual PPM Plant Procedure Manual PMU Panel Meter Unit PRA/PSA Probabilistic Risk Assessment / Probabilistic Safety Assessment PRM Process Radiation Monitor PWR Pressurized Water Reactor PSIG Pounds per Square Inch Gauge R Roentgen RCC Reactor Building Closed Cooling RCIC Reactor Core Isolation Cooling RCS Reactor Coolant System Rem Roentgen Equivalent Man RHR Residual Heat Removal RPS Reactor Protection System RPV Reactor Pressure Vessel RWCU Reactor Water Cleanup SGT Stand-By Gas Treatment SBO Station Blackout

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CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 18 of 274 SDOCA Security Defined Owner Controlled Area SDSP Shutdown Safety Plan SLC Standby Liquid Control SPDS Safety Parameter Display System SRO Senior Reactor Operator SSC Structure, System or Component SW Service Water TEA Turbine Exhaust Air TEDE Total Effective Dose Equivalent TAF Top of Active Fuel TSC Technical Support Center TSW Plant Service Water WEA Waste Exhaust Air 6.0 CGS-TO-NEI 99-01 EAL CROSSREFERENCE This cross-reference is provided to facilitate association and location of a Columbia Generating Station EAL within the NEI 99-01 IC/EAL identification scheme. Further information regarding the development of the Columbia Generating Station EALs based on the NEI guidance can be found in the EAL Comparison Matrix.

CGS NEI 99-01 EAL IC Example EAL RU1.1 AU1 1 RU1.2 AU1 2 RU1.3 AU1 3 RU2.1 AU2 1 RU2.2 AU2 2 RA1.1 AA1 1 RA1.2 AA1 2 RA1.3 AA1 3

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CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 19 of 274 CGS NEI 99-01 EAL IC Example EAL RA2.1 AA2 2 RA2.2 AA2 1 RA3.1 AA3 1 RS1.1 AS1 1 RS1.2 AS1 2 RS1.3 AS1 4 RG1.1 AG1 1 RG1.2 AG1 2 RG1.3 AG1 4 CU1.1 CU3 1 CU2.1 CU7 1 CU3.1 CUl 1 CU3.2 CU2 1 CU3.3 CU2 2 CU4.1 CU4 1 CU4.2 CU4 2 CU5.1 CU6 1,2 CU6.1 CU8 1 CA1.1 CA3 1 CA3.1 CA1 1,2 CA4.1 CA4 1,2 CS3.1 CS1 1 CS3.2 CS1 2 CS3.3 CS1 3

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CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 20 of 274 CGS NEI 99-01 EAL IC Example EAL CG3.1 CG1 1 CG3.2 CG1 2 EU1.1 E-HU1 1 FU1.1 FUl 1 FA1.1 FA1 1 FS1.1 FS1 1 FG1.1 FG1 1 HU1.1 HU1 1 HU1.2 HU1 2 HU1.3 HU1 4 HU1.4 HU1 3 HU1.5 HU1 5 HU2.1 HU2 1 HU2.2 HU2 2 HU3.1 HU3 1 HU3.2 HU3 2 HU4.1 HU4 1,2,3 HU6.1 HU5 1 HU7.1 N/A N/A HA1.1 HA1 1 HA1.2 HA1 2 HA1.3 HA1 4 HA1.4 HA1 3 HA1.5 HA1 5

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CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 21 of 274 CGS NEI 99-01 EAL IC Example EAL HA2.1 HA2 1 HA3.1 HA3 1 HA4.1 HA4 1,2 HA5.1 HA5 1 HA6.1 HA6 1 HA7.1 N/A N/A HS4.1 HS4 1 HS5.1 HS2 1 HS6.1 HS3 1 HG4.1 HG1 1,2 HG6.1 HG2 1 SU1.1 SUl 1 SU3.1 SU8 1 SU4.1 SU2 1 SU5.1 SU3 1 SU6.1 SU6 1,2 SU7.1 SU4 1 SU7.2 SU4 2 SU8.1 SU5 1,2 SA1.1 SA5 1 SA3.1 SA2 1 SA5.1 SA4 1 SS1.1 SS1 1 SS2.1 SS3 1

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CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 22 of 274 CGS NEI 99-01 EAL IC Example EAL SS3.1 SS2 1 SS5.1 SS6 1 SG1.1 SG1 1 SG3.1 SG2 1 7.0 ATTACHMENTS 7.1 EAL Bases 7.2 Fission Product Barrier Loss / Potential Loss Matrix and Basis

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CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 23 of 274 EAL BASES CATEGORY R INTRODUCTION 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 emergency classification.

At lower levels, abnormal radioactivity releases may be indicative of a failure of containment systems or precursors to more significant releases. At higher release rates, offsite radiological conditions may result which require offsite protective actions. Elevated area radiation levels in plant may also be indicative of the failure of containment systems or preclude access to plant vital equipment necessary to ensure plant safety.

Events of this category pertain to the following subcategories:

1. Offsite Rad Conditions Direct indication of effluent radiation monitoring systems provides a rapid assessment mechanism to determine releases in excess of classifiable limits. Projected offsite doses, actual off site field measurements or measured release rates via sampling indicate doses or dose rates above classifiable limits.

2. Onsite Rad Conditions/Fuel Pool Events Sustained general area radiation levels in excess of specified in the EALs indicating loss of control of radioactive materials or those levels which may preclude access to vital plant areas also warrant emergency classification.

3. CR Rad Sustained general area radiation levels in areas requiring continuous occupancy (Main Control Room D21 -S004 CH-19) in excess of 15 mR/hr also warrant emergency classification.

Attachment 7.1, EAL Bases

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CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 24 of 274 Category: R - Abnormal Rad Release/Rad Effluent Subcategory: 1 - Offsite Rad Conditions Initiating Condition: Any release of gaseous or liquid radioactivity to the environment greater than 2 times the ODCM for 60 minutes or longer EAL:

RU1.1 Unusual Event Any VALID gaseous monitor reading GT Table R-1 column "UE" for GE 60 min. (Note 2)

NOTE 2: The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the release duration has exceeded, or will likely exceed, the applicable time. In the absence of data to the contrary, assume that the release duration has exceeded the applicable time if an ongoing release is detected and the release start time is unknown.

Tale11Effluent Monitor Clasfcto Thesolds Release Point Monitor GE SAE Alert UE PRM-RE-1 B (I) - 1.81 E+05 cps 1.25E+04 cps 8.62E+02 cps Reactor Building Exhaust PRM-RE-1C (H) 4.22E+01 cps 0

T Turbine Building Exhaust TEA-RIS-13 (I) ............... 3.56E+04 cpm 3.33E+03 cpm TEA-RIS-13A (H) 8.11E+00 PMU 8.11E-01 PMU WEA-RIS-14 (I) -------- 5.55E+04 cpm Radwaste Building Exhaust WEA-RIS-14A (H) - 1.16E+01 PMU Radwaste Effluent FDR-RIS-606 ------- 200 X Hi-Hi alarm 2 X Hi-Hi alarm 0"

TSW Effluent TSW-RIS-5 2.00E-03 pCi/cc 2.OOE-05 pCi/cc Service Water Process A SW-RIS-604 ------- 2.00E+04 cps 2.00E+02 cps Service Water Process B SW-RIS-605 2.00E+04 cps 2.OOE+02 cps Mode Applicability:

All Attachment 7.1, EAL Bases

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CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 25 of 274 NEI 99-01 Basis:

The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the condition will likely exceed the applicable time.

This IC addresses a potential decrease in the level of safety of the plant as indicated by a radiological release that exceeds regulatory commitments for an extended period of time.

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

The ODCM multiples are specified in RU1 and RA1 only to distinguish between non-emergency conditions, and from each other. While these multiples obviously correspond to an off-site dose or dose rate, the emphasis in classifying these events is the degradation in the level of safety of the plant, not the magnitude of the associated dose or dose rate.

This EAL addresses radioactivity releases, that for whatever reason, cause effluent radiation monitor readings to exceed the threshold identified in the IC.

This EAL is intended for sites that have established effluent monitoring on non-routine release pathways for which a discharge permit would not normally be prepared.

CGS Basis:

The column "UE" gaseous release values in Table R-1 represent two times the appropriate ODCM release rate limits associated with the specified monitors (Ref. 2).

Instrumentation that may be used to assess this EAL is listed below (Ref. 1):

Reactor Building Elevated Discharge Monitor PRM-RE-1 This monitoring system measures radioactivity in the reactor building elevated release duct from the:

a) Offgas effluent, b) The gland seal system, c) Mechanical vacuum pump, d) The Standby Gas Treatment System, and e) The exhaust air from the Reactor Building Ventilation System.

Attachment 7.1, EAL Bases

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CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 26 of 274 The purpose of this monitoring subsystem is to characterize the radioactivity being released through the reactor building elevated discharge vent and to provide alarms in the event of abnormal operation. The monitor has no control function but an alarm is provided in the Control Room if the setpoint is exceeded.

Turbine Building Ventilation Exhaust Monitor TEA-RIS-1 3 This monitoring system measures radioactivity in the turbine building exhaust. The monitor detects fission and activation products from the steam which may leak from the turbine or other equipment in the building. The purpose of this monitoring subsystem is to characterize the radioactivity being released through the turbine building discharge vent and to provide alarms in the event of abnormal operation. The monitor is read and recorded in the main Control Room. An alarm is provided in the Control Room if the setpoint is exceeded.

Radwaste Building Ventilation Exhaust Monitor WEA-RIS-14 This monitoring system measures radioactivity in the radwaste building ventilation air exhaust.

Radioactivity originates from radwaste tank vents, processing equipment, sampling hoods, as well as from process treatment systems within the building. The purpose of this monitoring subsystem is to characterize the radioactivity being released through radwaste building discharge vent and to provide alarms in the event of abnormal operation. The monitor is read and recorded in the main Control Room. An alarm is provided in the Control Room if the setpoint is exceeded.

CGS Basis Reference(s):

1. CGS Offsite Dose Calculation Manual (ODCM)

2. Calculation NE-02-09-12 Attachment 7.1, EAL Bases

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CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 27 of 274 Category: R - Abnormal Rad Release / Rad Effluent Subcategory: 1 - Offsite Rad Conditions Initiating Condition: Any release of gaseous or liquid radioactivity to the environment greater than 2 times the ODCM for 60 minutes or longer EAL:

RU1.2 Unusual Event Any VALID liquid monitor reading GT Table R-1 column "UE" for GE 60 min. (Note 2)

NOTE 2: The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the release duration has exceeded, or will likely exceed, the applicable time. In the absence of data to the contrary, assume that the release duration has exceeded the applicable time if an ongoing release is detected and the release start time is unknown.

Table R-1 Effluent Monitor Classification Thresholds Release Point Monitor GE SAE Alert UE PRM-RE-1B (I) 1.81 E+05 cps 1.25E+04 cps 8.62E+02 cps Reactor Building Exhaust PRM-RE-lC (H) 4.22E+01 cps -

W 0

0 on2TrieBidnOxas Turbine Building Exhaust TEA-RIS-13 (I) ..... 3.56E+04 cpm 3.33E+i03 cpm TEA-RIS-13A (H) 8.11E+00 PMU 8.11E-01 PMU WEA-RIS-14 (I) - 5.55E+04 cpm Radwaste Building Exhaust WEA-RIS-14A (H) 1.16E+01 PMU Radwaste Effluent FDR-RIS-606 200 X Hi-Hi alarm 2 X Hi-Hi alarm

'S TSW Effluent TSW-RIS-5 2.OOE-03 pCi/cc 2.OOE-05 pCi/cc Service Water Process A SW-RIS-604 2.OOE+04 cps 2.OOE+02 cps Service Water Process B SW-RIS-605 2.OOE+04 cps 2.OOE+02 cps Mode Applicability:

All Attachment 7.1, EAL Bases

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CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 28 of 274 NEI 99-01 Basis:

The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the condition will likely exceed the applicable time.

This IC addresses a potential decrease in the level of safety of the plant as indicated by a radiological release that exceeds regulatory commitments for an extended period of time.

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

The ODCM multiples are specified in RU1 and RA1 only to distinguish between non-emergency conditions, and from each other. While these multiples obviously correspond to an off-site dose or dose rate, the emphasis in classifying these events is the degradation in the level of safety of the plant, not the magnitude of the associated dose or dose rate.

This EAL addresses radioactivity releases, that for whatever reason, cause effluent radiation monitor readings to exceed the threshold identified in the IC.

This EAL is intended for sites that have established effluent monitoring on non-routine release pathways for which a discharge permit would not normally be prepared (TSW & Service Water) as well as planned batch releases for which a radioactivity discharge permit is prepared (Radwaste Effluent).

CGS Basis:

The column "UE" liquid release values in Table R-1 represent two times the appropriate ODCM release rate limits associated with the specified monitors (Ref. 2).

Instrumentation that may be used to assess this EAL is listed below (Ref. 1):

• Radwaste Effluent FDR-RIS-606 This monitor measures the radioactivity in the liquid effluent prior to its entering the cooling tower blowdown line. All radioactive liquid effluent passes through this four inch line. The detector has seven decades of range and has a sensitivity of 1OE-6 pCi/cc for Cs-137. This monitor will automatically close a valve stopping the effluent release if its setpoint is exceeded.

The alarm/trip setpoint of the liquid Radwaste Effluent Monitor shall be set to ensure the limits of ODCM Section 6.2.1 are not exceeded for each batch of radioactive liquid effluent released.

• TSW Effluent TSW-RIS-5 The TSW system provides for the removal of heat from balance of plant auxiliary equipment. It is normally a non-radioactive system but it has the potential to become contaminated because Attachment 7.1, EAL Bases

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CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 29 of 274 it provides cooling to some reactor auxiliary systems. This radioactivity monitor is provided to detect that off-normal condition by monitoring for radioactivity in the TSW return header to the Circulating Water System.

Service Water A Process SW-RIS-604 & Service Water B Process SW-RIS-605 The Standby Service Water System provides cooling to the reactor during normal shutdown conditions. It also provides cooling of the reactor during emergency conditions. The SW system has the potential to become contaminated because of its interface with the heat exchangers that cool the reactor. This radiation monitor is provided to detect any increase in radioactivity in the system.

The TSW and Service Water Effluent Radioactivity Monitor High alarm setpoints are set in accordance with the ODCM and ensures continuous liquid releases do not exceed ODCM Section 6.2.1 limits.

CGS Basis Reference(s):

1. CGS ODCM

2. Calculation NE-02-09-12 Attachment 7.1, EAL Bases

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CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 30 of 274 Cateqory: R - Abnormal Rad Release / Rad Effluent Subcateaory: 1 - Offsite Rad Conditions Initiating Condition: Any release of gaseous or liquid radioactivity to the environment greater than 2 times the ODCM for 60 minutes or longer EAL:

RU1.3 Unusual Event Confirmed sample analyses for gaseous or liquid releases indicate concentrations or release rates GT 2 x ODCM limits for GE 60 min. (Note 2)

NOTE 2: The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the release duration has exceeded, or will likely exceed, the applicable time.

In the absence of data to the contrary, assume that the release duration has exceeded the applicable time if an ongoing release is detected and the release start time is unknown.

Mode Applicability:

All NEI 99-01 Basis:

The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the condition will likely exceed the applicable time.

This IC addresses a potential decrease in the level of safety of the plant as indicated by a radiological release that exceeds regulatory commitments for an extended period of time.

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

The ODCM multiples are specified in RU1 and RA1 only to distinguish between non-emergency conditions, and from each other. While these multiples obviously correspond to an off-site dose or dose rate, the emphasis in classifying these events is the degradation in the level of safety of the plant, not the magnitude of the associated dose or dose rate.

Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 31 of 274 This EAL includes any release for which a radioactivity discharge permit was not prepared, or a release that exceeds the conditions (e.g., minimum dilution flow, maximum discharge flow, alarm setpoints, etc.) on the applicable permit.

This EAL addresses uncontrolled releases that are detected by sample analyses, particularly on unmonitored pathways, e.g., spills of radioactive liquids into storm drains, heat exchanger leakage in river water systems, etc.

CGS Basis:

Releases in excess of two times the site Offsite Dose Calculation Manual (ODCM) (Ref. 1) instantaneous limits that continue for 60 minutes or longer represent an uncontrolled situation and hence, a potential degradation in the level of safety. The final integrated dose (which is very low in the Unusual Event emergency class) is not the primary concern here; it is the degradation in plant control implied by the fact that the release was not isolated within 60 minutes.

CGS Basis Reference(s):

1. CGS Offsite Dose Calculation Manual Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 32 of 274 Category: R - Abnormal Rad Release / Rad Effluent Subcateqory: 1 - Offsite Rad Conditions Initiatinq Condition: Any release of gaseous or liquid radioactivity to the environment greater than 200 times the ODCM for 15 minutes or longer EAL:

RA1.1 Alert Any VALID gaseous monitor reading GT Table R-1 column "Alert" for GE 15 min. (Note 2)

NOTE 2: The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the release duration has exceeded, or will likely exceed, the applicable time. In the absence of data to the contrary, assume that the release duration has exceeded the applicable time if an ongoing release is detected and the release start time is unknown.

Table.R-1. EffluentMonitor Classification Thresholds Release Point Monitor GE SAE Alert UE PRM-RE-1 B (I) ------- 1.81 E+05 cps 1.25E+04 cps 8.62E+02 cps Reactor Building Exhaust PRM-RE-lC (H) 4.22E+01 cps O TEA-RIS-13 (I) ------- 3.56E+04 cpm 3.33E+03 cpm to Turbine Building Exhaust TEA-RIS-13A (H) 8.11E+00 PMU 8.11E-01 PMU ...............

WEA-RIS-14 (I) 5.55E+04 cpm Radwaste Building Exhaust WEA-RIS-14A (H) 1.16E+01 PMU Radwaste Effluent FDR-RIS-606 ------- ------- 200 X Hi-Hi alarm 2 X Hi-Hi alarm

.3 TSW Effluent TSW-RIS-5 -------- 2.OOE-03 pCi/cc 2.OOE-05 iCi/cc

,-J Service Water Process A SW-RIS-604 2.00E+04 cps 2.OOE+02 cps Service Water Process B SW-RIS-605 2.00E+04 cps 2.OOE+02 cps Mode Applicability:

All Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 33 of 274 NEI 99-01 Basis:

The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the condition will likely exceed the applicable time.

This IC addresses an actual or substantial potential decrease in the level of safety of the plant as indicated by a radiological release that exceeds regulatory commitments for an extended period of time.

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

The ODCM multiples are specified in RU1 and RA1 only to distinguish between non-emergency conditions, and from each other. While these multiples obviously correspond to an off-site dose or dose rate, the emphasis in classifying these events is the degradation in the level of safety of the plant, not the magnitude of the associated dose or dose rate.

This EAL is intended for sites that have established effluent monitoring on non-routine release pathways for which a discharge permit would not normally be prepared.

CGS Basis:

The column "Alert" gaseous release values in Table R-1 represent the geometric average between the UE and SAE values associated with the specified monitors (Ref. 2).

Instrumentation that may be used to assess this EAL is listed below (Ref. 1):

Reactor Building Elevated Discharge Monitor PRM-RE-1 This monitoring system measures radioactivity in the reactor building elevated release duct from the:

a) Offgas effluent, b) The gland seal system, c) Mechanical vacuum pump, d) The Standby Gas Treatment System, and e) The exhaust air from the Reactor Building Ventilation System.

The purpose of this monitoring subsystem is to characterize the radioactivity being released through the reactor building elevated discharge vent and to provide alarms in the event of Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 34 of 274 abnormal operation. The monitor has no control function but an alarm is provided in the Control Room if the setpoint is exceeded.

Turbine Building Ventilation Exhaust Monitor TEA-RIS-13 This monitoring system measures radioactivity in the turbine building exhaust. The monitor detects fission and activation products from the steam which may leak from the turbine or other equipment in the building. The purpose of this monitoring subsystem is to characterize the radioactivity being released through the turbine building discharge vent and to provide alarms in the event of abnormal operation. The monitor is read and recorded in the main Control Room. An alarm is provided in the Control Room if the setpoint is exceeded.

Radwaste Building Ventilation Exhaust Monitor WEA-RIS-1 4 This monitoring system measures radioactivity in the radwaste building ventilation air exhaust.

Radioactivity originates from radwaste tank vents, processing equipment, sampling hoods, as well as from process treatment systems within the building. The purpose of this monitoring subsystem is to characterize the radioactivity being released through radwaste building discharge vent and to provide alarms in the event of abnormal operation. The monitor is read and recorded in the main Control Room. An alarm is provided in the Control Room if the setpoint is exceeded.

CGS Basis Reference(s):

1. CGS Offsite Dose Calculation Manual

2. Calculation NE-02-09-12 Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 35 of 274 Category: R - Abnormal Rad Release / Rad Effluent Subcategory: 1 - Offsite Rad Conditions Initiating Condition: Any release of gaseous or liquid radioactivity to the environment greater than 200 times the ODCM for 15 minutes or longer EAL:

RA1.2 Alert Any VALID liquid monitor reading GT Table R-1 column "Alert" for GE 15 min. (Note 2)

NOTE 2: The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the release duration has exceeded, or will likely exceed, the applicable time. In the absence of data to the contrary, assume that the release duration has exceeded the applicable time if an ongoing release is detected and the release start time is unknown.

Table R-1 Effluent Monitor Classification Thresholds Release Point Monitor GE SAE Alert UE PRM-RE-1B (I) 1.81 E+05 cps 1.25E+04 cps 8.62E+02 cps Reactor Building Exhaust PRM-RE-1C (H) 4.22E+01 cps --------

0 TEA-RIS-13 (1) --------- 3.56E+04 cpm 3.33E+03 cpm U0 Turbine Building Exhaust W TEA-RIS-13A (H) 8.11E+00 PMU 8.11E-01 PMU ........

WEA-RIS-14 (I) - 5.55E+04 cpm Radwaste Building Exhaust WEA-RIS-14A(H) 1.16E+01 PMU Radwaste Effluent FDR-RIS-606 200 X Hi-Hi alarm 2 X Hi-Hi alarm

- TSW Effluent TSW-RIS-5 2.OOE-03 pCi/cc 2.OOE-05 IACi/cc Service Water Process A SW-RIS-604 -- 2.00E+04 cps 2.OOE+02 cps Service Water Process B SW-RIS-605 2.00E+04 cps 2.OOE+02 cps Mode Applicability:

All Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 36 of 274 NEI 99-01 Basis:

The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the condition will likely exceed the applicable time.

This IC addresses an actual or substantial potential decrease in the level of safety of the plant-as indicated by a radiological release that exceeds regulatory commitments for an extended period of time.

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

The ODCM multiples are specified in RU1 and RA1 only to distinguish between non-emergency conditions, and from each other. While these multiples obviously correspond to an off-site dose or dose rate, the emphasis in classifying these events is the degradation in the level of safety of the plant, not the magnitude of the associated dose or dose rate.

This EAL includes any release for which a radioactivity discharge permit was not prepared, or a release that exceeds the conditions (e.g., minimum dilution flow, maximum discharge flow, alarm setpoints, etc.) on the applicable permit.

This EAL is intended for sites that have established effluent monitoring on non-routine release pathways for which a discharge permit would not normally be prepared (TSW & Service Water) as well as planned batch releases for which a radioactivity discharge permit is prepared (Radwaste Effluent).

CGS Basis:

The column "Alert" liquid release values in Table R-1 represent two hundred times the appropriate ODCM release rate limits associated with the specified monitors (Ref. 2).

Instrumentation that may be used to assess this EAL is listed below (Ref. 1):

Radwaste Effluent FDR-RIS-606 This monitor measures the radioactivity in the liquid effluent prior to its entering the cooling tower blowdown line. All radioactive liquid effluent passes through this four inch line. The detector has seven decades of range and has a sensitivity of 1OE-6 pCi/cc for Cs-137. This monitor will automatically close a valve stopping the effluent release if its setpoint is exceeded.

The alarm/trip setpoint of the liquid Radwaste Effluent Monitor shall be set to ensure the limits of ODCM Section 6.2.1 are not exceeded for each batch of radioactive liquid effluent released.

Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 37 of 274 TSW Effluent TSW-RIS-5 The TSW system provides for the removal of heat from balance of plant auxiliary equipment. It is normally a non-radioactive system but it has the potential to become contaminated because it provides cooling to some reactor auxiliary systems. This radioactivity monitor is provided to detect that off-normal condition by monitoring for radioactivity in the TSW return header to the Circulating Water System.

Service Water A Process SW-RIS-604 & Service Water B Process SW-RIS-605 The Standby Service Water System provides cooling to the reactor during normal shutdown conditions. It also provides cooling of the reactor during emergency conditions. The SW system has the potential to become contaminated because of its interface with the heat exchangers that cool the reactor. This radiation monitor is provided to detect any increase in radioactivity in the system.

The TSW and Service Water Effluent Radioactivity Monitor High alarm setpoints are set in accordance with the ODCM and ensures continuous liquid releases do not exceed ODCM Section 6.2.1 limits.

CGS Basis Reference(s):

1. CGS Offsite Dose Calculation Manual

2. Calculation NE-02-09-12 Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 38 of 274 Category: R - Abnormal Rad Release / Rad Effluent Subcategory: 1 - Offsite Rad Conditions Initiating Condition: Any release of gaseous or liquid radioactivity to the environment greater than 200 times the ODCM for 15 minutes or longer EAL:

RA1.3 Alert Confirmed sample analyses for gaseous or liquid releases indicate concentrations or release rates GT 200 x ODCM limits for GE 15 min. (Note 2)

NOTE 2: The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the release duration has exceeded, or will likely exceed, the applicable time. In the absence of data to the contrary, assume that the release duration has exceeded the applicable time if an ongoing release is detected and the release start time is unknown.

Mode Applicability:

All NEI 99-01 Basis:

The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the condition will likely exceed the applicable time.

This IC addresses an actual or substantial potential decrease in the level of safety of the plant as indicated by a radiological release that exceeds regulatory commitments for an extended period of time.

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

The ODCM multiples are specified in RU1 and RA1 only to distinguish between non-emergency conditions, and from each other. While these multiples obviously correspond to an off-site dose or dose rate, the emphasis in classifying these events is the degradation in the level of safety of the plant, not the magnitude of the associated dose or dose rate.

Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 39 of 274 This EAL addresses uncontrolled releases that are detected by sample analyses, particularly on unmonitored pathways, e.g., spills of radioactive liquids into storm drains, heat exchanger leakage in river water systems, etc.

CGS Basis:

Confirmed sample analyses in excess of two hundred times the site Offsite Dose Calculation Manual (Ref. 1) limits that continue for 15 minutes or longer represent an uncontrolled situation and hence, a potential degradation in the level of safety. This event escalates from the Unusual Event by raising the magnitude of the release by a factor of 100 over the Unusual Event level (i.e., 200 times ODCM).

The required release duration was reduced to 15 minutes in recognition of the raised severity.

CGS Basis Reference(s):

1. CGS Offsite Dose Calculation Manual Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 40 of 274 Category: R - Abnormal Rad Release / Rad Effluent Subcategory: 1 - Offsite Rad Conditions Initiating Condition: Off-site dose resulting from an actual or imminent release of gaseous radioactivity greater than 100 mrem TEDE or 500 mrem Thyroid CDE for the actual or projected duration of the release EAL:

RS1.1 Site Area Emergency Any VALID radiation monitor reading GT Table R-1 column "SAE" for GE 15 min. (Note 1)

NOTE 1: If dose assessment results are available, declaration should be based on dose assessment instead of radiation monitor values. Do not delay declaration awaiting dose assessment results.

The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the condition will likely exceed the applicable time.

Table R-I. Effluent Monitor, Classification Thresholds Release Point Monitor GE SAE Alert UE PRM-RE-1 B (I) 1.81 E+05 cps 1.25E+04 cps 8.62E+02 cps Reactor Building Exhaust PRM-RE-1C (H) 4.22E+01 cps ........

iA T Turbine Building Exhaust TEA-RIS-13 (I) ------- ------- 3.56E+04 cpm 3.33E+03 cpm 4',

w TEA-RIS-13A (H) 8.11E+00 PMU 8.11E-01 PMU WEA-RIS-14 (I) 5.55E+04 cpm Radwaste Building Exhaust WEA-RIS-14A(H) ------- 1.16E+01 PMU Radwaste Effluent FDR-RIS-606 ------- ---- --- 200 X Hi-Hi alarm 2 X Hi-Hi alarm "3 TSW Effluent TSW-RIS-5 2.OOE-03 pCi/cc 2.00E-05 pCi/cc

.27 Service Water Process A SW-RIS-604 2.OOE+04 cps 2.00E+02 cps Service Water Process B SW-RIS-605 2.00E+04 cps 2.OOE+02 cps Mode Applicability:

All Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 41 of 274 NEI 99-01 Basis:

This EAL addresses radioactivity releases that result in doses at or beyond the site boundary that exceed 10% of the EPA Protective Action Guides (PAGs). Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public.

While these failures are addressed by other EALs, this EAL provides appropriate diversity and addresses events which may not be able to be classified on the basis of plant status alone. It is important to note that for the more severe accidents the release may be unmonitored or there may be large uncertainties associated with the source term and/or meteorology.

The TEDE dose is set at 10% of the EPA PAG, while the 500 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.

The Table R-1 monitor list includes effluent monitors on all potential release pathways.

Since dose assessment is based on actual meteorology, whereas the monitor reading EAL is not, the results from these assessments may indicate that the classification is not warranted, or may indicate that a higher classification is warranted. For this reason, emergency implementing procedures call for the timely performance of dose assessments using actual meteorology and release information. If the results of these dose assessments are available when the classification is made (e.g., initiated at a lower classification level), the dose assessment results override the monitor reading EAL.

CGS Basis:

This EAL address gaseous radioactivity releases, that for whatever reason, cause effluent radiation monitor readings corresponding to site boundary doses that exceed either:

100 mRem TEDE 500 mRem CDE Thyroid The column "SAE" gaseous effluent release values in Table R-1 correspond to calculated doses of 10% of the EPA Protective Action Guidelines (TEDE or CDE Thyroid) (Ref. 2).

Instrumentation that may be used to assess this EAL is listed below (Ref. 1):

Reactor Building Elevated Discharge Monitor PRM-RE-1C This monitoring system measures radioactivity in the reactor building elevated release duct from the:

a) Offgas effluent, b) The gland seal system, c) Mechanical vacuum pump, Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 42 of 274 d) The Standby Gas Treatment System, and e) The exhaust air from the Reactor Building Ventilation System.

The purpose of this monitoring subsystem is to characterize the radioactivity being released through the reactor building elevated discharge vent and to provide alarms in the event of abnormal operation. The monitor has no control function but an alarm is provided in the Control Room if the setpoint is exceeded.

Turbine Building Ventilation Exhaust Monitor TEA-RIS-13 This monitoring system measures radioactivity in the turbine building exhaust. The monitor detects fission and activation products from the steam which may leak from the turbine or other equipment in the building. The purpose of this monitoring subsystem is to characterize the radioactivity being released through the turbine building discharge vent and to provide alarms in the event of abnormal operation. The monitor is read and recorded in the main Control Room. An alarm is provided in the Control Room if the setpoint is exceeded.

CGS Basis Reference(s):

1. CGS Offsite Dose Calculation Manual

2. Calculation NE-02-09-12 Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 43 of 274 Category: R - Abnormal Rad Release / Rad Effluent Subcateaory: 1 - Offsite Rad Conditions Initiating Condition: Off-site dose resulting from an actual or imminent release of gaseous radioactivity greater than 100 mrem TEDE or 500 mrem Thyroid CDE for the actual or projected duration of the release EAL:

RS1.2 Site Area Emergency Dose assessment using actual meteorology indicates doses GT 0.1 Rem TEDE or 0.5 Rem thyroid CDE at or beyond the site boundary Mode Applicability:

All NEI 99-01 Basis:

This EAL addresses radioactivity releases that result in doses at or beyond the site boundary that exceed 10% of the EPA Protective Action Guides (PAGs). Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public.

While these failures are addressed by other EALs, this EAL provides appropriate diversity and addresses events which may not be able to be classified on the basis of plant status alone. It is important to note that for the more severe accidents the release may be unmonitored or there may be large uncertainties associated with the source term and/or meteorology.

The TEDE dose is set at 10% of the EPA PAG, while the 500 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.

Since dose assessment is based on actual meteorology, whereas the monitor reading EAL is not, the results from these assessments may indicate that the classification is not warranted, or may indicate that a higher classification is warranted. For this reason, emergency implementing procedures call for the timely performance of dose assessments using actual meteorology and release information. If the results of these dose assessments are available when the classification is made (e.g., initiated at a lower classification level), the dose assessment results override the monitor reading EAL.

Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 44 of 274 CGS Basis:

For the purposes of this EAL, the Site Boundary for CGS is the 1950-meter radius around the plant as depicted in Figure 3-1 of the CGS ODCM (Ref. 1). The key-hole area between the river and this radius is not within the Site Boundary.

CGS Site Boundary CGS Basis Reference(s):

1. CGS Offsite Dose Calculation Manual Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 45 of 274 Category: R - Abnormal Rad Release / Rad Effluent Subcategory: 1 - Offsite Rad Conditions Initiating Condition: Off-site dose resulting from an actual or imminent release of gaseous radioactivity greater than 100 mrem TEDE or 500 mrem Thyroid CDE for the actual or projected duration of the release EAL:

RS1.3 Site Area Emergency Field survey results indicate closed window dose rate GT 0.1 Rem/hr expected to continue for GE 1 hr at or beyond the site boundary (Note 3)

OR Analyses of field survey samples indicate thyroid CDE GT 0.5 Rem for 1 hr of inhalation at or beyond the site boundary (Note 3)

NOTE 3: The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the condition will likely exceed the applicable time.

Mode Applicability:

All NEI 99-01 Basis:

This EAL addresses radioactivity releases that result in doses at or beyond the site boundary that exceed 10% of the EPA Protective Action Guides (PAGs). Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public.

While these failures are addressed by other EALs, this EAL provides appropriate diversity and addresses events which may not be able to be classified on the basis of plant status alone. It is important to note that for the more severe accidents the release may be unmonitored or there may be large uncertainties associated with the source term and/or meteorology.

The TEDE dose is set at 10% of the EPA PAG, while the 500 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.

CGS Basis:

The 500 mRem integrated CDE thyroid dose was established in consideration of the 1:5 ratio of the EPA Protective Action Guidelines for TEDE and thyroid exposure. In establishing the field survey Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 46 of 274 emergency action levels, a duration of one hour is assumed. Therefore, the dose rate EALs are based on a Site Boundary dose rate of 100 mRem/hr TEDE or 500 mRem for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of inhalation CDE thyroid, whichever is more limiting.

For the purposes of this EAL, the Site Boundary for CGS is the 1950-meter radius around the plant as depicted in Figure 3-1 of the CGS ODCM (Ref. 1). The key-hole area between the river and this radius is not within the Site Boundary.

CGS Site Boundary CGS Basis Reference(s):

1. CGS Offsite Dose Calculation Manual Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 47 of 274 Category: R - Abnormal Rad Release / Rad Effluent Subcategory: 1 - Offsite Rad Conditions Initiatinq Condition: Off-site dose resulting from an actual or imminent release of gaseous radioactivity greater than 1 Rem TEDE or 5 Rem Thyroid CDE for the actual or projected duration of the release using actual meteorology EAL:

RG1.1 General Emergency Any VALID radiation monitor reading GT Table R-1 column "GE" for GE 15 min. (Note 1)

NOTE 1: If dose assessment results are available, declaration should be based on dose assessment instead of radiation monitor values. Do not delay declaration awaiting dose assessment results.

The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the condition will likely exceed the applicable time.

Table.R4l- Effluent Monitor Classification Thresholds Release Point Monitor GE SAE Alert UE PRM-RE-1B (I) 1.81 E+05 cps 1.25E+04 cps 8.62E+02 cps Reactor Building Exhaust PRM-RE-lC (H) 4.22E+01 cps -

0 TEA-RIS-13 (I) -------- 3.56E+04 cpm 3.33E+03 cpm Tn ig a TEA-RIS-13A (H) 8.11E+00 PMU 8.11E-01 PMU ........

WEA-RIS-14 (I) ---- - 5.55E+04 cpm Radwaste Building Exhaust WEA-RIS-14A (H) - 1.16E+01 PMU --------

Radwaste Effluent FDR-RIS-606 ................- 200 X Hi-Hi alarm 2 X Hi-Hi alarm "O

'S TSW Effluent TSW-RIS-5 2.00E-03 pCi/cc 2.00E-05 pCi/cc Service Water Process A SW-RIS-604 2.OOE+04 cps 2.OOE+02 cps Service Water Process B SW-RIS-605 2.00E+04 cps 2.OOE+02 cps Mode Applicability:

All Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 48 of 274 NEI 99-01 Basis:

This EAL addresses radioactivity releases that result in doses at or beyond the site boundary that exceed the EPA Protective Action Guides (PAGs). Public protective actions will be necessary.

Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public and likely involve fuel damage.

While these failures are addressed by other EALs, this EAL provides appropriate diversity and addresses events which may not be able to be classified on the basis of plant status alone. It is important to note that for the more severe accidents the release may be unmonitored or there may be large uncertainties associated with the source term and/or meteorology.

The TEDE dose is set at the EPA PAG, while the 5000 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.

The Table R-1 monitor list includes effluent monitors on all potential release pathways.

Since dose assessment is based on actual meteorology, whereas the monitor reading EAL is not, the results from these assessments may indicate that the classification is not warranted, or may indicate that a higher classification is warranted. For this reason, emergency implementing procedures call for the timely performance of dose assessments using actual meteorology and release information. If the results of these dose assessments are available when the classification is made (e.g., initiated at a lower classification level), the dose assessment results override the monitor reading EAL.

CGS Basis:

This EAL address gaseous radioactivity releases, that for whatever reason, cause effluent radiation monitor readings corresponding to site boundary doses that exceed either:

• 1 Rem TEDE

• 5 Rem CDE 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 CDE Thyroid) (Ref. 2).

Instrumentation that may be used to assess this EAL is listed below (Ref. 1):

Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 49 of 274 Reactor Building Elevated Discharge Monitor PRM-RE-1C This monitoring system measures radioactivity in the reactor building elevated release duct from the:

a) Offgas effluent, b) the gland seal system, c) Mechanical vacuum pump, d) the Standby Gas Treatment System, and e) the exhaust air from the Reactor Building Ventilation System.

The purpose of this monitoring subsystem is to characterize the radioactivity being released through the reactor building elevated discharge vent and to provide alarms in the event of abnormal operation. The monitor has no control function but an alarm is provided in the Control Room if the setpoint is exceeded.

Turbine Building Ventilation Exhaust Monitor TEA-RIS-1 3 This monitoring system measures radioactivity in the turbine building exhaust. The monitor detects fission and activation products from the steam which may leak from the turbine or other equipment in the building. The purpose of this monitoring subsystem is to characterize the radioactivity being released through the turbine building discharge vent and to provide alarms in the event of abnormal operation. The monitor is read and recorded in the main Control Room. An alarm is provided in the Control Room if the setpoint is exceeded.

CGS Basis Reference(s):

1. CGS Offsite Dose Calculation Manual

2. Calculation NE-02-09-12 Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 50 of 274 Cateqory: R - Abnormal Rad Release / Rad Effluent Subcategory: 1 - Offsite Rad Conditions Initiating Condition: Off-site dose resulting from an actual or imminent release of gaseous radioactivity greater than 1 Rem TEDE or 5 Rem Thyroid CDE for the actual or projected duration of the release using actual meteorology EAL:

RG1.2 General Emeraency Dose assessment using actual meteorology indicates doses GT 1 Rem TEDE or 5 Rem thyroid CDE at or beyond the site boundary Mode Applicability:

All NEI 99-01 Basis:

This EAL addresses radioactivity releases that result in doses at or beyond the site boundary that exceed the EPA Protective Action Guides (PAGs). Public protective actions will be necessary.

Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public and likely involve fuel damage.

While these failures are addressed by other EALs, this EAL provides appropriate diversity and addresses events which may not be able to be classified on the basis of plant status alone. It is important to note that for the more severe accidents the release may be unmonitored or there may be large uncertainties associated with the source term and/or meteorology.

The TEDE dose is set at the EPA PAG, while the 5000 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.

Since dose assessment is based on actual meteorology, whereas the monitor reading EAL is not, the results from these assessments may indicate that the classification is not warranted, or may indicate that a higher classification is warranted. For this reason, emergency implementing procedures call for the timely performance of dose assessments using actual meteorology and release information. If the results of these dose assessments are available when the classification is made (e.g., initiated at a lower classification level), the dose assessment results override the monitor reading EAL.

Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 51 of 274 CGS Basis:

The General Emergency values are based on the boundary dose resulting from an actual or imminent release of gaseous radioactivity that exceeds 1 Rem TEDE or 5 Rem CDE thyroid for the actual or projected duration of the release. Actual meteorology is specifically identified since it gives the most accurate dose assessment. Actual meteorology (including forecasts) should be used whenever possible.

For the purposes of this EAL, the Site Boundary for CGS is the 1950-meter radius around the plant as depicted in Figure 3-1 of the CGS ODCM (Ref. 1). The key-hole area between the river and this radius is not within the Site Boundary.

Site Area Boundary CGS Basis Reference(s):

1. CGS Offsite Dose Calculation Manual Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 52 of 274 Categqory: R - Abnormal Rad Release / Rad Effluent Subcategory: 1 - Offsite Rad Conditions Initiating Condition: Off-site dose resulting from an actual or imminent release of gaseous radioactivity greater than 1 Rem TEDE or 5 Rem Thyroid CDE for the actual or projected duration of the release using actual meteorology EAL:

RG1.3 General Emergency Field survey results indicate closed window dose rates GT 1 Rem/hr expected to continue for GE 1 hr at or beyond the site boundary (Note 3)

OR Analyses of field survey samples indicate thyroid CDE GT 5 Rem for 1 hr of inhalation at or beyond the site boundary (Note 3)

NOTE 3: The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the condition will likely exceed the applicable time.

Mode Applicability:

All NEI 99-01 Basis:

This EAL addresses radioactivity releases that result in doses at or beyond the site boundary that exceed the EPA Protective Action Guides (PAGs). Public protective actions will be necessary.

Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public and likely involve fuel damage.

While these failures are addressed by other EALs, this EAL provides appropriate diversity and addresses events which may not be able to be classified on the basis of plant status alone. It is important to note that for the more severe accidents the release may be unmonitored or there may be large uncertainties associated with the source term and/or meteorology.

The TEDE dose is set at the EPA PAG, while the 5000 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.

Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 53 of 274 CGS Basis:

The 5 Rem integrated CDE thyroid dose was established in consideration of the 1:5 ratio of the EPA Protective Action Guidelines for TEDE and thyroid exposure. In establishing the dose rate emergency action levels, a duration of one hour is assumed. Therefore, the dose rate EALs are based on a Site Boundary dose rate of 1 Rem/hr TEDE or 5 Rem for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of inhalation CDE thyroid, whichever is more limiting.

For the purposes of this EAL, the Site Boundary for CGS is the 1950-meter radius around the plant as depicted in Figure 3-1 of the CGS ODCM (Ref. 1). The key-hole area between the river and this radius is not within the Site Boundary.

CGS Site Area Boundary CGS Basis Reference(s):

1. CGS Offsite Dose Calculation Manual Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 54 of 274 Cateqory: R - Abnormal Rad Release / Rad Effluent Subcategory: 2 - Onsite Rad Conditions/Fuel Pool Events Initiating Condition: Unplanned rise in plant radiation EAL:

RU2.1 Unusual Event UNPLANNED water level drop in the reactor cavity or spent fuel pool as indicated by fuel pool water level LT 22' 4" (FPC-LI-21) or H13.P626.FPC1 2-2 / H13.P627.FPC2.2-2 (FUEL POOL LEVEL HIGH/LOW)

AND VALID area radiation monitor reading rise on any of the following:

ARM-RIS-1 Reactor Building Fuel Pool Area ARM-RIS-2 Reactor Building Fuel Pool Area ARM-RIS-34 Reactor Building Elevation 606 Mode Applicability:

All NEI 99-01 Basis:

This EAL addresses increased radiation levels as a result of water level decreases above irradiated fuel or events that have resulted, or may result, in unplanned increases in radiation dose rates within plant buildings. These radiation increases represent a loss of control over radioactive material and represent a potential degradation in the level of safety of the plant.

The refueling pathway is the combination of cavities and pools in which spent fuel may be located.

While a radiation monitor could detect an increase in dose rate due to a drop in the water level, it might not be a reliable indication of whether or not the fuel is covered.

For example, an ARM reading may increase due to planned evolutions such as head lift, or even a fuel assembly being raised in the manipulator mast. Also, a monitor could in fact be properly responding to a known event involving transfer or relocation of a source, stored in or near the fuel pool or responding to a planned evolution such as removal of the reactor head. Generally, increased radiation monitor indications will need to combined with another indicator (or personnel report) of water loss.

Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 55 of 274 For refueling events where the water level drops below the RPV flange classification would be via CU2.1. This event escalates to an Alert per RA2.1 if irradiated fuel outside the reactor vessel is uncovered. For events involving irradiated fuel in the reactor vessel, escalation would be via the Fission Product Barrier Table for events in operating modes 1-3.

CGS Basis:

The fuel pool low level alarm is actuated by level switch FP-LS-4A when fuel pool water level drops below 605' 5-1/2" (LT 22' 4" on FPC-LI-21) (Ref. 1,2).

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 RPV and spent fuel pool as well as for spent fuel pool drain down events.

The maximum expected level reduction rate in the reactor cavity, with the gates removed, is 3 in/min with a 7500 gpm leak. For the Spent Fuel Pool with the gate installed, the maximum expected level reduction rate is 1.4 in/min with an 1150 gpm leak (Ref. 3).

For a loss of shielding, the source of the radiation is within the reactor cavity or fuel pool. Without the shielding provided by normal water inventory in the Fuel Pool, equipment pool, and/or reactor cavity, radiation levels from irradiated fuel and activation products will rise substantially. If the dryer and separator are in the equipment pool and are uncovered, unshielded exposure rates could be up to 3 Rem per hour on accessible portions of the refuel floor. With these components installed in the vessel when uncovered, the exposure rates from the vessel internals at the vessel flange would be 250 mRem per hour. A loss of Fuel Pool inventory through the gate to the reactor cavity will leave 1.5 feet of water over the top of the stored fuel. Radiation dose rates could exceed 100 Rem per hour at the edge of the Fuel Pool, assuming fuel with one day of decay (Ref. 3).

ARM-RIS-1 (102 - 106 mrem/hr) and ARM-RIS-2 (1 - 104 mrem/hr) are located in the fuel pool area of the 606' elevation of the Reactor Building. ARM-RIS-34 (10 104 mrem/hr) is located on the east side of the 606' elevation of the Reactor Building (Ref. 4).

CGS Basis Reference(s):

1. PPM 4.626.FPC1-2.2 (4.626.FPC2-2.2) Fuel Pool Level High/Low

2. PPM 4.627.FPC2-2.2 (4.627.FPC2-2.2) Fuel Pool Level High/Low

3. ABN-FPC-LOSS Loss of Fuel Pool Cooling

4. FSAR Table 12.3-1 Area Monitors Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 56 of 274 Cateqory: R - Radioactivity Release / Area Radiation Subcategory: 2 - Onsite Rad Conditions/Fuel Pool Events Initiating Condition: Unplanned rise in plant radiation EAL:

RU2.2 Unusual Event UNPLANNED VALID area radiation monitor readings or survey results rise by a factor of 1,000 over normal levels *

• Normal levels can be considered as the highest reading in the past 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> excluding the current peak value Mode Applicability:

All NEI 99-01 Basis:

This EAL addresses unplanned increases in radiation dose rates within plant buildings. These radiation increases represent a loss of control over radioactive material and represent a potential degradation in the level of safety of the plant. This EAL addresses increases in plant radiation levels that represent a loss of control of radioactive material resulting in a potential degradation in the level of safety of the plant.

This EAL excludes radiation level increases that result from planned activities such as use of radiographic sources and movement of radioactive waste materials. A specific list of ARMs is not required as it would restrict the applicability of the threshold. The intent is to identify loss of control of radioactive material in any monitored area.

CGS Basis:

The ARMs applicable to this EAL include ARMs located in the Turbine Building and Radwaste Building as well as the Reactor Building.

CGS Basis Reference(s):

None Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 57 of 274 Category: R - Abnormal Rad Release / Rad Effluent Subcategory: 2 - Onsite Rad Conditions/Fuel Pool Events Initiating Condition: Damage to irradiated fuel or loss of water level that has resulted or will result in the uncovering of irradiated fuel outside the RPV EAL:

RA2.1 Alert Irradiated fuel damage or loss of water level that causes a VALID high alarm on any of the following radiation monitors:

• ARM-RIS-1 Reactor Building Fuel Pool Area 0 ARM-RIS-2 Reactor Building Fuel Pool Area

• ARM-RIS-34 Reactor Building Elevation 606

• REA-RIS-609A-D Rx Bldg Vent Mode Applicability:

All NEI 99-01 Basis:

This EAL addresses increases in radiation dose rates within plant buildings, and may be a precursor to a radioactivity release to the environment. These events represent a loss of control over radioactive material and represent an actual or substantial potential degradation in the level of safety of the plant.

These events escalate from RU2.1 in that fuel activity has been released, or is anticipated due to fuel heatup. This IC applies to spent fuel requiring water coverage and is not intended to address spent fuel which is licensed for dry storage.

This EAL addresses radiation monitor indications of fuel uncovery and/or fuel damage.

Increased ventilation monitor readings may be indication of a radioactivity release from the fuel, confirming that damage has occurred. Increased background at the ventilation monitor due to water level decrease may mask increased ventilation exhaust airborne activity and needs to be considered.

While a radiation monitor could detect an increase in dose rate due to a drop in the water level, it might not be a reliable indication of whether or not the fuel is covered.

For example, a refueling bridge ARM reading may increase due to planned evolutions such as head lift, or even a fuel assembly being raised in the manipulator mast. Also, a monitor could in fact be Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 58 of 274 properly responding to a known event involving transfer or relocation of a source, stored in or near the fuel pool or responding to a planned evolution such as removal of the reactor head. Generally, increased radiation monitor indications will need to combined with another indicator (or personnel report) of water loss.

Escalation of this emergency classification level, if appropriate, would be based on RS1.1 or RG1.1.

CGS Basis:

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 RPV and spent fuel pool.

The maximum expected level reduction rate in the reactor cavity, with the gates removed, is 3 in/min with a 7500 gpm leak. For the Spent Fuel Pool with the gate installed, the maximum expected level reduction rate is 1.4 in/min with an 1150 gpm leak (Ref. 1).

For a loss of shielding, the source of the radiation is within the reactor cavity or fuel pool. Without the shielding provided by normal water inventory in the Fuel Pool, equipment pool, and/or reactor cavity, radiation levels from irradiated fuel and activation products will rise substantially. If the dryer and separator are in the equipment pool and are uncovered, unshielded exposure rates could be up to 3 Rem per hour on accessible portions of the refuel floor. With these components installed in the vessel when uncovered, the exposure rates from the vessel internals at the vessel flange would be 250 mRem per hour. A loss of Fuel Pool inventory through the gate to the reactor cavity will leave 1.5 feet of water over the top of the stored fuel. Radiation dose rates could exceed 100 Rem per hour at the edge of the Fuel Pool, assuming fuel with one day of decay (Ref. 1).

ARM-RIS-1 (102 _ 106 mrem/hr) and ARM-RIS-2 (1 - 104 mrem/hr) are located in the fuel pool area of the 606' elevation of the Reactor Building. ARM-RIS-34 (101 - 104 mrem/hr) is located on the east side of the 606' elevation of the Reactor Building (Ref. 2). The ARM alarm setpoints are:

ARM-RIS GE 300 mR/hr (Ref. 3)

ARM-RIS GE 15 mR/hr (Ref. 3)

ARM-RIS GE 1 R/hr (Ref. 4)

REA-RIS-609A-D are the Reactor Building Exhaust Plenum radiation monitors. This system monitors the radiation level of the reactor building ventilation system exhaust plenum prior to its discharge from the building into the elevated release duct. A high radioactivity level in the exhaust system could be due to fission gases from damaged or leaking spent fuel or an accident (Ref. 5). Actuation of the High-High alarm (GT 13 mR/hr) actuates a Secondary Containment isolation and actuates SGT (Ref. 6).

CGS Basis Reference(s):

1. ABN-FPC-LOSS Loss of Fuel Pool Cooling

2. CGS FSAR Table 12.3-1 Area Monitors

3. PPM 4.602.A5-1.1 Refueling Floor Area Rad High Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 59 of 274

4. PPM 4.602.A5-3.1 Reactor Building Rad High

5. FSAR Section 11.5.2.1.2 Reactor Building Exhaust Plenum Radiation Monitoring System

6. PPM 4.602.A5-1.4 Reactor Building Exh Plenum Rad Hi-Hi Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 60 of 274 Category: R - Abnormal Rad Release / Rad Effluent Subcategory: 2 - Onsite Rad Conditions/Fuel Pool Events Initiating Condition: Damage to irradiated fuel or loss of water level that has resulted or will result in the uncovering of irradiated fuel outside the RPV EAL:

Mode Applicability:

All NEI 99-01 Basis:

This event represents a loss of control over radioactive material and represents an actual or substantial potential degradation in the level of safety of the plant.

This event escalates from RU2.1 in that fuel activity release is anticipated due to fuel heatup. This EAL applies to spent fuel requiring water coverage and is not intended to address spent fuel which is licensed for dry storage.

Indications may include instrumentation such as water level and local area radiation monitors, and personnel (e.g., refueling crew) reports. If available, video cameras may allow remote observation.

Depending on available level instrumentation, the declaration threshold may need to be based on indications of water makeup rate or decrease in water storage tank level.

Escalation of this emergency classification level, if appropriate, would be based on RS1.1 or RG1.1.

CGS Basis:

None CGS Basis Reference(s):

None Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 61 of 274 Category: R - Abnormal Rad Release / Rad Effluent Subcategory: 3 - CR Rad Initiating Condition: Rise in radiation levels within the facility that impedes operation of systems required to maintain plant safety functions EAL:

RA3.1 Alert Dose rates GT 15 mR/hr within the Control Room envelope (D21 -S004 CH-19)

Mode Applicability:

All NEI 99-01 Basis:

This EAL addresses increased radiation levels that impact continued operation in areas requiring continuous occupancy to maintain safe operation or to perform a safe shutdown.

The cause and/or magnitude of the increase in radiation levels is not a concern of this EAL. The Emergency Director must consider the source or cause of the increased radiation levels and determine if any other EAL may be involved.

The value of 15mR/hr is derived from the GDC 19 value of 5 rem in 30 days with adjustment for expected occupancy times. Although Section III.D.3 of NUREG-0737, "Clarification of TMI Action Plan Requirements", provides that the 15 mR/hr value can be averaged over the 30 days, the value is used here without averaging, as a 30 day duration implies an event potentially more significant than an Alert.

Areas requiring continuous occupancy include the Control Room.

CGS Basis:

The CGS Control Room requires continuous occupancy because of its importance to assure safe plant operations and control of site security functions (Secondary Alarm Station).

Control Room ARM (D21-S004 CH-19) measures area radiation in a range of 1 - 104 mR/hr (Ref. 1).

CGS Basis Reference(s):

1. FSAR Table 12.3-1 Area Monitors Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 62 of 274 Cateaory C - Cold Shutdown / Refueling System Malfunction EAL Group: Cold Conditions (RCS temperature LE 2000 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 that occurs at the end of an extended outage has less significance than a similar loss occurring during the first week after shutdown. Compounding these events is the likelihood that instrumentation necessary for assessment may also be inoperable. The cold shutdown and refueling system malfunction EALs are based on performance capability to the extent possible with consideration given to RCS integrity, Containment Closure, and fuel clad integrity for the applicable operating modes (4 - Cold Shutdown, 5 - Refuel, D - Defueled).

The events of this category pertain to the following subcategories:

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

2. Loss of DC Power Loss of vital critical DC 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 125 VDC power sources.

3. RPV Level RPV water level is a measure of inventory available to ensure adequate core cooling and, therefore, maintain fuel clad integrity. The RPV provides a volume for the coolant that covers the reactor core. The RPV 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.

4. RCS Temperature Uncontrolled or inadvertent temperature or pressure increases are indicative of a potential loss of safety functions.

Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 63 of 274

5. Communications Certain events that degrade plant operator ability to effectively communicate with essential personnel within or external to the plant warrant emergency classification.

6. Inadvertent Criticality Inadvertent criticalities pose potential personnel safety hazards as well being indicative of losses of reactivity control.

Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 64 of 274 Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - Loss of AC Power Initiating Condition: AC power capability to emergency buses reduced to a single power source for 15 minutes or longer such that any additional single failure would result in complete loss of AC power to emergency buses EAL:

CU1.1 Unusual Event AC power capability, Table C-5, to emergency buses SM-7 and SM-8 reduced to a single power source for GE 15 min. (Note 3) such that any additional single power source failure would result in a loss of all emergency bus AC power NOTE 3: The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the condition will likely exceed the applicable time.

Table C-54 AC Power Sources (C !Conditions):,...

Offsite

• Startup Transformer TR-S

• Backup Transformer TR-B

• Backfeed 500 KV power through Main Transformers to the Normal Transformers TR-N1/N2 Onsite

• DG1

• DG2 Mode Applicability:

4 - Cold Shutdown, 5 - Refuel Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 65 of 274 NEI 99-01 Basis:

The condition indicated by this IC is the degradation of the off-site and on-site AC power systems such that any additional single failure would result in a station blackout. This condition could occur due to a loss of off-site power with a concurrent failure of all but one emergency generator to supply power to its emergency busses. The subsequent loss of this single power source would escalate the event to an Alert in accordance with CA1.1.

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

CGS Basis:

Table C-5 provides the list of AC power sources available to power emergency buses in cold conditions. (Ref. 1, 2)

Station Startup and shutdown 230KV power comes from the Ashe substation, located just north of the plant, through Startup transformer TR-S. The startup transformer usually supplies station auxiliary loads when the main generator is not available. Station Normal power, 25KV, is supplied from the main generator via Normal Auxiliary transformers TR-N1 and TR-N2. When the main generator is ready, all station auxiliary loads are transferred from the Startup transformer to the Normal Auxiliary transformers. Station Backup 115KV power from the Benton Substation feeder, located about 4 miles southeast of the plant, can be supplied to emergency buses SM-7 and SM-8. (Ref. 3, 4)

The buses addressed in this EAL are 4160V SM-7 and SM-8 (Ref. 5, 6).

Bus SM-7 is normally energized from TR-N1 thru SM-1 when the main generator is operating and from TR-S thru SM-1 when the main generator is not operating. SM-7 can receive backup power from TR-B or emergency power from DG1.

• Bus SM-8 is normally energized from TR-N1 thru SM-3 when the main generator is operating and from TR-S thru SM-3 when the main generator is not operating. SM-8 can receive backup power from TR-B or emergency power from DG2.

Credit is not taken in this EAL for SM-4/DG3 crosstie capability because:

Establishing the crosstie to SM-7 or SM-8 is assumed to require more than 15 minutes.

Prolonged loss of offsite AC power reduces the required redundancy and potentially degrades the level of safety by rendering the plant more vulnerable to a complete loss of AC power (Station Blackout). (Ref. 7, 8)

The HPCS pump does not have any long-term decay heat removal systems, in particular, wetwell cooling.

Power to buses SM-7 and SM-8 may come from either its respective standby diesel generator or from the Switch Yard through the Normal, Startup or Backup Transformers. Regardless of the source of power, failure of the remaining power source would result, at least temporarily, in a station blackout.

Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 66 of 274 The determining factor of whether or not to classify then becomes the amount of time required to restore power to an emergency bus.

Each emergency bus has multiple sources of power. In order of preference, they are (Ref. 5, 6):

1. Normal Transformers: TR-N1 (SM-7 and/or SM-8)

2. Startup Transformers: TR-S (SM-7 and/or SM-8)

3. Backup Transformers: TR B (SM-7 and/or SM-8)

4. Emergency Diesel Generators: DG 1 (SM-7), DG 2 (SM-8)

It is possible to remove startup power from service and continue to supply the plant during shutdown conditions by backfeeding 500 KV power from Ashe Substation through the Main Transformers, the Normal Transformers and associated "N"breakers. This involves disconnecting the Main Generator from the Isolated Phase conductors (25 KV system) and overriding various interlocks.

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

CGS Basis Reference(s):

1. FSAR Figure 8.1-2.1 Main One-Line Diagram - Main Buses

2. FSAR Figure 8.1-2.2 Main One-Line Diagram - Emergency Buses

3. FSAR Section 8.2

4. 01-53 Offsite Power

5. FSAR Section 8.3

6. ABN-ELEC-LOOP Loss Of All Off-Site Electrical Power

7. FSAR Section 8A.1

8. PPM 5.6.1 Station Blackout (SBO)

Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 67 of 274 Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - Loss of AC Power Initiating Condition: Loss of all offsite and all onsite AC power to emergency buses for 15 minutes or longer EAL:

CA1.1 Alert Loss of all offsite and all onsite AC power, Table C-5, to emergency buses SM-7 and SM-8 for GE 15 min. (Note 3)

NOTE 3: The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the condition will likely exceed the applicable time.

Table C-5 AC.Power Sources (Cold Conditions) ......

Offsite

" Startup Transformer TR-S

" Backup Transformer TR-B

" Backfeed 500 KV power through Main Transformers to the Normal Transformers TR-N 1/N2 Onsite

" DG1

" DG2 Mode Applicability:

4 - Cold Shutdown, 5 - Refuel, D - Defueled Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 68 of 274 NEI 99-01 Basis:

Loss of all AC power compromises all plant safety systems requiring electric power including RHR, ECCS, Containment Heat Removal, Spent Fuel Heat Removal and the Ultimate Heat Sink.

The event can be classified as an Alert when in cold shutdown, refueling, or defueled mode because of the significantly reduced decay heat and lower temperature and pressure, increasing the time to restore one of the emergency busses, relative to that specified for the Site Area Emergency EAL.

Escalating to Site Area Emergency, if appropriate, is by Abnormal Rad Levels / Radiological Effluent ICs.

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

CGS Basis:

Table C-5 provides the list of AC power sources available to power emergency buses in cold conditions. (Ref. 1, 2)

Station Startup and shutdown 230KV power comes from the Ashe substation, located just north of the plant, through Startup transformer TR-S. The startup transformer usually supplies station auxiliary loads when the main generator is not available. Station Normal power, 25KV, is supplied from the main generator via Normal Auxiliary transformers TR-N1 and TR-N2. When the main generator is ready, all station auxiliary loads are transferred from the Startup transformer to the Normal Auxiliary transformers. Station Backup 115KV power from the Benton Substation feeder, located about 4 miles southeast of the plant, can be supplied to emergency buses SM-7 and SM-8. (Ref. 3, 4)

The buses addressed in this EAL are 4160V SM-7 and SM-8 (Ref. 5, 6).

Bus SM-7 is normally energized from TR-N1 thru SM-1 when the main generator is operating and from TR-S thru SM-1 when the main generator is not operating. SM-7 can receive backup power from TR-B or emergency power from DGI.

Bus SM-8 is normally energized from TR-N1 thru SM-3 when the main generator is operating and from TR-S thru SM-3 when the main generator is not operating. SM-8 can receive backup power from TR-B or emergency power from DG2.

Credit is not taken in this EAL for SM-4/DG3 crosstie capability because:

Establishing the crosstie to SM-7 or SM-8 is assumed to require more than 15 minutes.

Prolonged loss of offsite AC power reduces the required redundancy and potentially degrades the level of safety by rendering the plant more vulnerable to a complete loss of AC power (Station Blackout). (Ref. 7, 8)

The HPCS pump does not have any long-term decay heat removal systems, in particular, wetwell cooling Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 69 of 274 Power to buses SM-7 and SM-8 may come from either its respective standby diesel generator or from the Switch Yard through the Normal, Startup or Backup Transformers. Regardless of the source of power, failure of the remaining power source would result, at least temporarily, in a station blackout.

The determining factor of whether or not to classify then becomes the amount of time required to restore power to an emergency bus.

Each emergency bus has multiple sources of power. In order of preference, they are (Ref. 5, 6):

1. Normal Transformers: TR-N1 (SM-7 and/or SM-8)

2. Startup Transformers: TR-S (SM-7 and/or SM-8)

3. Backup Transformers: TR B (SM-7 and/or SM-8)

4. Emergency Diesel Generators: DG 1 (SM-7), DG 2 (SM-8)

It is possible to remove startup power from service and continue to supply the plant during shutdown conditions by backfeeding 500 KV power from Ashe Substation through the Main Transformers, the Normal Transformers and associated "N" breakers. This involves disconnecting the Main Generator from the Isolated Phase conductors (25 KV system) and overriding various interlocks.

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

CGS Basis Reference(s):

1. FSAR Figure 8.1-2.1 Main One-Line Diagram - Main Buses

2. FSAR Figure 8.1-2.2 Main One-Line Diagram - Emergency Buses

3. FSAR Section 8.2

4. 01-53 Offsite Power

5. FSAR Section 8.3

6. ABN-ELEC-LOOP Loss Of All Off-Site Electrical Power

7. FSAR Section 8A.1

8. PPM 5.6.1 Station Blackout (SBO)

Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 70 of 274 Cateqory: C - Cold Shutdown / Refueling System Malfunction Subcateqory: 2 - Loss of DC Power Initiating Condition: Loss of required DC power for 15 minutes or longer EAL:

CU2.1 Unusual Event LT 108 VDC bus voltage indications on required 125 VDC buses DP-S1-1 and DP-S1-2 for GE 15 min. (Note 3)

NOTE 3: The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the condition will likely exceed the applicable time.

Mode Applicability:

4 - Cold Shutdown, 5 - Refuel NEI 99-01 Basis:

The purpose of this EAL and its associated EALs is to recognize a loss of DC power compromising the ability to monitor and control the removal of decay heat during Cold Shutdown or Refueling operations.

It is intended that the loss of the operating (operable) train is to be considered. If this loss results in the inability to maintain cold shutdown, the escalation to an Alert will be per CA4.1.

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

CGS Basis:

The 125 VDC system is illustrated in Figure C-2 (Ref. 1,2).

The 125 VDC Class 1 E DC power system consists of three electrically independent and separate distribution systems (S1-1, S1-2, and S1 -HPCS). S1 -HPCS is not included in this EAL. Each DC distribution system has a battery and a battery charger that are normally connected to the bus such that these two sources of power are operating in parallel. The charger is normally supplying system electrical loads with the battery on a float charge. Each battery has the necessary amp-hour discharge capacity to sustain system loads for a minimum of two hours. This capacity is specifically for a loss of power to the charger coincident with a design basis accident. The batteries have capacity to carry design load at 60°F without decreasing battery voltage below 1.81 volts/cell (or 108 VDC, Ref. 3) with loss of output from the battery chargers during the specified period. Battery capacity is sufficient to provide starting currents while operating at full load. (Ref. 4)

Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 71 of 274 This EAL is the cold condition equivalent of the hot condition loss of DC power EAL SS2.1.

CGS Basis Reference(s):

1. E505-1 DC One Line Diagram

2. E505-2 DC One Line Diagram

3. Calculation No. 2.05.01 Battery Sizing, Voltage Drop, and Charger Studies for Div. 1 & 2 Systems

4. FSAR Section 8.3.2 Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 72 of 274 Figure C-2:125 VDC System (Ref. 1, 2) 1)73 JJTR-7-73 CRTBS73y 40 MMC-S1 -2)

DP-SI-2A NSSS BOARD

-INVERTER IN-2A INVERTER IN-2B DP-SI-2D DIV.2 CRIT SWGR

&RBEOTE S/D PANEL DP-SI-1E D/G #1 DIST. PANEL DP-S1-2E DP-SI-IF DG #2 DIST. PANEL DIV. 1 CRfl SWGR ALT-REMOTE S/D PANEL MC-S2-1ACONTPWR Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 73 of 274 Categqory: C - Cold Shutdown / Refueling System Malfunction Subcateqory: 3 - RPV Level Initiating Condition: Unplanned loss of RPV inventory EAL:

CU3.1 Unusual Event In Mode 4, RCS leakage results in the inability to maintain or restore RPV level GT +13.0 in. for GE 15 min. (Note 3)

NOTE 3: The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the condition will likely exceed the applicable time.

Mode Applicability:

4 - Cold Shutdown NEI 99-01 Basis:

This EAL is considered to be a potential degradation of the level of safety of the plant. The inability to maintain or restore level is indicative of loss of RCS inventory.

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

Prolonged loss of RCS Inventory may result in escalation to the Alert emergency classification level via either CA3.1 or CA4.1.

CGS Basis:

Figure C-3 illustrates the elevations of the RPV level instrument ranges (Ref. 1, 2, 3).

+13.0 in. is the RPV low water level scram setpoint (Ref. 4, 5).

RPV level is monitored from -310 in. to +400 in. to ensure adequate coverage for expected and postulated conditions of RPV level. RPV level measurement is derived by the differential pressure that exists between a reference leg and variable leg. All instruments are referenced to a benchmark at 527.5 in. above the inside bottom head of the reactor vessel. This benchmark corresponds to the bottom edge of the steam dryer skirt and is the 0 in. reference indication on the RPV level instruments. RPV level monitoring is subdivided into five ranges identified as:

Narrow provides indication and control signals for normal plant operation and protection system actuation.

Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 74 of 274

• Wide provides indication and control signals for transient conditions below the normal operating band and emergency equipment actuation.

• Upset provides indication for transient conditions above normal operating band.

• Shutdown provides indication for vessel flood up and activities.

• Fuel Zone provides indication for long term accident conditions where reactor level cannot be restored.

In preparation for refueling operations, the Upset and Shutdown RPV level instruments are modified as follows to provide continuous level indication from within the RPV to the refuel floor:

• RFW LR 608, RPV Upset Level Range Recorder: Channel 1 Red Pen is spanned for 0 to 500 in.

• MS LI 605, RPV Shutdown Level Range: The existing scale is removed from MS LI 605 and replaced with a temporary scale which reads 0 to 500 in. with marks at 107.5 in. for main steam line elevation and 216 in. for RPV flange elevation.

• TDAS Point X327 is spanned for 0 to 500 in.

Two individuals are also stationed on the refuel floor to monitor RPV level concurrently with RPV disassembly activities. (Ref. 6, 7)

This Cold Shutdown EAL represents the hot condition EAL SU8.1, in which RCS leakage is associated with Technical Specification limits. In Cold Shutdown, these limits are not applicable; hence, the use of RPV level as the parameter of concern in this EAL (Ref. 8).

CGS Basis Reference(s):

1. FSAR Section 7.5.1.1

2. FSAR Table 7.5-1

3. FSAR Figure 7.7-1

4. Technical Specifications Table 3.3.1.1-1

5. PPM 5.1.1 RPV Control

6. PPM 10.27.39 Refueling Reactor Vessel Level (Temporary)

7. SOP-CAVITY-FILL Reactor Cavity and Dryer Separator Pit Fill

8. Technical Specifications 3.4.7 Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 75 of 274 Figure C-3 RPV Levels (Ref. 1, 2, 3)

- 110

- 161.2 Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 76 of 274 Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 3 - RPV Level Initiating Condition: Unplanned loss of RPV inventory EAL:

CU3.2 Unusual Event In Mode 5, UNPLANNED RPV level drop below EITHER of the following for GE 15 min. (Note 3):

• 216 in. (RPV flange)

• RPV level band (when the RPV level band is established below the RPV flange)

NOTE 3: The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the condition will likely exceed the applicable time.

Mode Applicability:

5 - Refuel NEI-99-01 Basis:

This EAL is a precursor of more serious conditions and considered to be a potential degradation of the level of safety of the plant.

Refueling evolutions that decrease RPV water level below the RPV flange are carefully planned and procedurally controlled. An UNPLANNED event that results in water level decreasing below the RPV flange, or below the planned RPV water level for the given evolution (if the planned RPV water level is already below the RPV flange), warrants declaration of a UE due to the reduced RPV inventory that is available to keep the core covered.

The allowance of 15 minutes was chosen because it is reasonable to assume that level can be restored within this time frame using one or more of the redundant means of refill that should be available. If level cannot be restored in this time frame then it may indicate a more serious condition exists.

Continued loss of RCS Inventory will result in escalation to the Alert emergency classification level via either CA3.1 or CA4.1.

This EAL involves a decrease in RPV level below the top of the RPV flange that continues for 15 minutes due to an UNPLANNED event. This EAL is not applicable to decreases in flooded reactor cavity level, which is addressed by RU2.1, until such time as the level decreases to the level of the RPV flange.

Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 77 of 274 CGS Basis:

Figure C-3 illustrates the elevations of the RPV level instrument ranges (Ref. 1, 2, 3).

216 in. is the RPV flange level (Ref. 4).

RPV level is monitored from -310 in. to +400 in. to ensure adequate coverage for expected and postulated conditions of RPV level. RPV level measurement is derived by the differential pressure that exists between a reference leg and variable leg. All instruments are referenced to a benchmark at 527.5 in. above the inside bottom head of the reactor vessel. This benchmark corresponds to the bottom edge of the steam dryer skirt and is the 0 in. reference indication on the RPV level instruments. RPV level monitoring is subdivided into five ranges identified as:

0 Narrow provides indication and control signals for normal plant operation and protection system actuation.

• Wide provides indication and control signals for transient conditions below the normal operating band and emergency equipment actuation.

• Upset provides indication for transient conditions above normal operating band.

• Shutdown provides indication for vessel flood up and activities.

• Fuel Zone provides indication for long term accident conditions where reactor level cannot be restored.

In preparation for refueling operations, the Upset and Shutdown RPV level instruments are modified as follows to provide continuous level indication from within the RPV to the refuel floor:

RFW LR 608, RPV Upset Level Range Recorder: Channel 1 Red Pen is spanned for 0 to 500 in.

MS LI 605, RPV Shutdown Level Range: The existing scale is removed from MS LI 605 and replaced with a temporary scale which reads 0 to 500 in. with marks at 107.5 in. for main steam line elevation and 216 in. for RPV flange elevation.

TDAS Point X327 is spanned for 0 to 500 in.

Two individuals are also stationed on the refuel floor to monitor RPV level concurrently with RPV disassembly activities. (Ref. 4, 5)

Depending on the configuration of the reactor cavity and spent fuel pool (gates installed or removed) and the status of refueling operations (all spent fuel seated in storage racks/RPV or a bundle raised on the fuel grapple), a loss of inventory may reduce water shielding above irradiated components or Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 78 of 274 spent fuel. EALs in Subcategory R.2 should be assessed for emergency classification due to the radiological consequences of such events.

This Cold Shutdown EAL represents the hot condition EAL SU8.1, in which RCS leakage is associated with Technical Specification limits. In Cold Shutdown, these limits are not applicable; hence, the use of RPV level as the parameter of concern in this EAL (Ref. 6).

CGS Basis Reference(s):

1. FSAR Section 7.5.1.1

2. FSAR Table 7.5-1

3. FSAR Figure 7.7-1

4. PPM 10.27.39 Refueling Reactor Vessel Level (Temporary)

5. SOP-CAVITY-FILL Reactor Cavity and Dryer Separator Pit Fill

6. Technical Specifications 3.4.7 Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 79 of 274 Figure C-3 RPV Levels (Ref. 1, 2, 3)

(LEVEL NO.

HEIGHTABOVE VESSELZERO TRIPPOINr (INCICATED).

(IN.) _ _ _ _

()582 + 54~5 (7 568 +4(15 (4) 559 + 31.5 900o

-4 I NOTE: SCALE ININCHES (3) 540.5 + 13.0

-110

- 161.2

- 310 Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 80 of 274 Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 3 - RPV Level Initiating Condition: Unplanned loss of RPV inventory EAL:

CU3.3 Unusual Event In Mode 5, RPV level cannot be monitored with a loss of RPV inventory as indicated by any unexplained RPV leakage indication, Table C-1 Table C-I RPV Leakage Indicaions.

" Excessive Reactor Building equipment/floor drain sump pump-out rate

• Wetwell level rise

• Observation of unisolable RCS leakage Mode Applicability:

5 - Refuel NEI 99-01 Basis:

This EAL is a precursor of more serious conditions and considered to be a potential degradation of the level of safety of the plant.

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

Escalation to the Alert emergency classification level would be via either CA3.1 or CA4.1.

Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 81 of 274 CGS Basis:

In preparation for refueling operations, the Upset and Shutdown RPV level instruments are modified as follows to provide continuous level indication from within the RPV to the refuel floor:

RFW LR 608, RPV Upset Level Range Recorder: Channel 1 Red Pen is spanned for 0 to 500 in.

MS LI 605, RPV Shutdown Level Range: The existing scale is removed from MS LI 605 and replaced with a temporary scale which reads 0 to 500 in. with marks at 107.5 in. for main steam line elevation and 216 in. for RPV flange elevation.

TDAS Point X327 is spanned for 0 to 500 in.

Two individuals are also stationed on the refuel floor to monitor RPV level concurrently with RPV disassembly activities. (Ref. 1, 2)

In this EAL, all RPV level indication is unavailable and the RPV inventory loss must be detected by the leakage indications listed in Table C-1. Level increases must be evaluated against other potential sources of leakage such as cooling water sources inside the drywell to ensure they are indicative of RPV leakage. Reactor Building equipment or floor drain sump level rise may be indicative of RPV inventory losses external to the Primary Containment from systems connected to the RPV (Ref. 4, 5).

With RHR System operating in the Shutdown Cooling mode, an unexplained rise in wetwell level could be indicative of RHR valve misalignment or leakage (Ref. 6). If the make-up rate to the RPV unexplainably rises above the pre-established rate, a loss of RPV inventory may be occurring even if the source of the leakage cannot be immediately identified. Visual observation of leakage from systems connected to the RCS in areas outside the Primary Containment that cannot be isolated could be indicative of a loss of RPV inventory.

Depending on the configuration of the reactor cavity and spent fuel pool (gates installed or removed) and the status of refueling operations (all spent fuel seated in storage racks/RPV or a bundle raised on the fuel grapple), a loss of inventory may reduce water shielding above irradiated components or spent fuel. EALs in Subcategory R.2 should be assessed for emergency classification due to the radiological consequences of such events.

CGS Basis Reference(s):

1. SOP-CAVITY-FILL Reactor Cavity and Dryer Separator Pit Fill

2. Technical Specifications 3.4.7

3. FSAR Section 7.6.1.3

4. SOP-EDR-OPS Equipment Drain System Operation

5. SOP-FDR-OPS Floor Drain System Operation

6. SOP-RHR-SDC RHR Shutdown Cooling Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 82 of 274 Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 3 - RPV Level Initiating Condition: Loss of RPV inventory EAL:

CA3.1 Alert RPV level LT -50 in.

OR RPV level cannot be monitored for GE 15 min. (Note 3) with any unexplained RPV leakage indication, Table C-1 NOTE 3: The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the condition will likely exceed the applicable time.

Table C-1 RPV Leakage Indications 0 Excessive Reactor Building equipment/floor drain sump pump-out rate

• Wetwell level rise

• Observation of unisolable RCS leakage Mode Applicability:

4 - Cold Shutdown, 5 - Refuel NEI 99-01 Basis:

This EAL serves as precursors to a loss of ability to adequately cool the fuel. The magnitude of this loss of water indicates that makeup systems have not been effective and may not be capable of preventing further RPV level decrease and potential core uncovery. This condition will result in a minimum emergency classification level of an Alert.

1 st Condition The BWR Low-Low ECCS Actuation Setpoint/Level 2 was chosen because it is a standard setpoint at which some available injection systems automatically start. The inability to restore and maintain level after reaching this setpoint would be indicative of a failure of the RCS barrier.

2 nd Condition Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 83 of 274 In the cold shutdown mode, normal RPV level instrumentation systems will usually be available. In the refueling mode, normal means of RPV level indication may not be available. Redundant means of RPV level indication will usually be installed (including the ability to monitor level visually) to assure that the ability to monitor level will not be interrupted. However, if all level indication were to be lost during a loss of RCS inventory event, the operators would need to determine that RPV inventory loss was occurring by observing sump and tank level changes. Sump and tank level increases must be evaluated against other potential sources of leakage such as cooling water sources inside the containment to ensure they are indicative of RCS leakage.

The 15-minute duration for the loss of level indication was chosen because it is half of the CS2.3 Site Area Emergency EAL duration. Significant fuel damage is not expected to occur until the core has been uncovered for greater than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> per the analysis referenced in the CG3.2 basis. Therefore this EAL meets the definition for an Alert.

If RPV level continues to lower then escalation to Site Area Emergency will be via CS3.1, CS3.2, or CS3.3.

CGS Basis:

The threshold RPV level of -50 in. is the low-low ECCS actuation setpoint (Ref. 1, 2). RPV level is normally monitored using the instruments in Figure C-3 (Ref. 3, 4, 5).

In Cold Shutdown mode, the RCS will normally be intact and standard RPV level monitoring means are available. In the Refuel mode, the RCS is not intact and RPV level may be monitored by different means, including the ability to monitor level visually.

In the second condition of this EAL, all RPV level indication is unavailable and the RPV inventory loss must be detected by the leakage indications listed in Table C-1. Level increases must be evaluated against other potential sources of leakage such as cooling water sources inside the drywell to ensure they are indicative of RPV leakage. Reactor Building equipment or floor drain sump level rise may be indicative of RPV inventory losses external to the Primary Containment from systems connected to the RPV (Ref. 7, 8). With RHR System operating in the Shutdown Cooling mode, ah unexplained rise in wetwell level could be indicative of RHR valve misalignment or leakage (Ref. 9). If the make-up rate to the RPV unexplainably rises above the pre-established rate, a loss of RPV inventory may be occurring even if the source of the leakage cannot be immediately identified. Visual observation of leakage from systems connected to the RCS in areas outside the Primary Containment that cannot be isolated could be indicative of a loss of RPV inventory.

Depending on the configuration of the reactor cavity and spent fuel pool (gates installed or removed) and the status of refueling operations (all spent fuel seated in storage racks/RPV or a bundle raised on the fuel grapple), a loss of inventory may reduce water shielding above irradiated components or spent fuel. EALs in Subcategory R.2 should be assessed for emergency classification due to the radiological consequences of such events.

Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 84 of 274 CGS Basis Reference(s):

1. Technical Specifications Table 3.3.5.1-1

2. PPM 5.1.1 RPV Control

3. FSAR Section 7.5.1.1

4. FSAR Table 7.5-1

5. FSAR Figure 7.7-1

6. FSAR Section 7.6.1.3

7. SOP-EDR-OPS Equipment Drain System Operation

8. SOP-FDR-OPS Floor Drain System Operation

9. SOP-RHR-SDC RHR Shutdown Cooling Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 85 of 274 Figure C-3 RPV Levels (Ref. 3, 4, 5)

LEVEL HEIGHT ABOVE TRIP POINT NO. VESSELZERO (INDICATED)

(IN.)

(8) 582 + 54.5 (7) 568 + 4M5 (4) 559 + 31.5 (3) 540.5 + 13.0 (2) 477.5 - 50.0 (1) 398.5 - 129.0 0 UFEACTOR Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 86 of 274 Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 3 - RPV Level Initiating Condition: Loss of RPV inventory affecting core decay heat removal capability EAL:

CS3.1 Site Area Emergency With CONTAINMENT CLOSURE not established (i.e., primary and secondary containment not functional), RPV level LT -56 in.

Mode Applicability:

4 - Cold Shutdown, 5 - Refuel NEI 99-01 Basis:

Under the conditions specified by this EAL, continued decrease in RPV level is indicative of a loss of inventory control. Inventory loss may be due to an RCS breach, pressure boundary leakage, or continued boiling in the RPV. Thus, declaration of a Site Area Emergency is warranted.

Escalation to a General Emergency is via CG3.1 or RG1.1.

CGS Basis:

When RPV level decreases to -56 in., water level is six inches below the low-low ECCS actuation setpoint (Ref. 1, 2).

The inability to restore and maintain level after reaching this setpoint infers a failure of the RCS barrier and Potential Loss of the Fuel Clad barrier.

Containment Closure is defined as the SDSP actions taken to secure primary or secondary containment and its associated structures, systems, and components as a functional barrier to fission product release under existing plant conditions. This definition is less restrictive than Technical Specification criteria governing Primary and Secondary Containment operability. If the Technical Specification criteria are met, therefore, Containment Closure has been established. (Ref. 3, 4, 5)

CGS Basis Reference(s):

1. Technical Specifications Table 3.3.5.1-1

2. PPM 5.1.1 RPV Control

3. Technical Specifications 3.6.1.1

4. Technical Specifications 3.6.4.1

5. PPM 1.20.3 Outage Risk Management Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 87 of 274 Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 3 - RPV Level Initiating Condition: Loss of RPV inventory affecting core decay heat removal capability EAL:

CS3.2 Site Area Emergency With CONTAINMENT CLOSURE established (i.e., primary or secondary containment functional),

RPV level LT -161 in.

Mode Applicability:

4 - Cold Shutdown, 5 - Refuel NEI 99-01 Basis:

Under the conditions specified by this EAL, continued decrease in RCS/RPV level is indicative of a loss of inventory control. Inventory loss may be due to an RCS breach, pressure boundary leakage, or continued boiling in the RPV. Thus, declaration of a Site Area Emergency is warranted.

Escalation to a General Emergency is via CG3.1 or AG1.1.

CGS Basis:

When RPV level drops the top of active fuel (an indicated RPV level of -161 in.), core uncovery starts to occur (Ref. 1, 2).

Containment Closure is defined as the SDSP actions taken to secure primary or secondary containment and its associated structures, systems, and components as a functional barrier to fission product release under existing plant conditions. This definition is less restrictive than Technical Specification criteria governing Primary and Secondary Containment operability. If the Technical Specification criteria are met, therefore, Containment Closure has been established. (Ref. 3, 4)

CGS Basis Reference(s):

1. Calculation NE-02-03-05 Attachment 3 Note 8

2. PPM 5.1.1 RPV Control

3. Technical Specifications 3.6.1.1

4. Technical Specifications 3.6.4.1

5. PPM 1.20.3 Outage Risk Management Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 88 of 274 Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 3 - RPV Level Initiating Condition: Loss of RPV inventory affecting core decay heat removal capability EAL:

CS3.3 Site Area Emergency RPV level cannot be monitored for GE 30 min. (Note 3) with a loss of RPV inventory as indicated by any of the following:

• Any unexplained RPV leakage indication, Table C-1

• Erratic Source Range Monitor indication NOTE 3: The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the condition will likely exceed the applicable time.

Table C-1 RPV Leakage Indications

• Excessive Reactor Building equipment/floor drain sump pump-out rate

• Wetwell level rise

" Observation of unisolable RCS leakage Mode Applicability:

4 - Cold Shutdown, 5 - Refuel NEI 99-01 Basis:

Under the conditions specified by this EAL, continued decrease in RPV level is indicative of a loss of inventory control. Inventory loss may be due to an RCS breach, pressure boundary leakage, or continued boiling in the RPV. Thus, declaration of a Site Area Emergency is warranted.

Escalation to a General Emergency is via CG3.2 or AG1.1.

In the cold shutdown mode, normal RPV level and RPV level instrumentation systems will usually be available. In the refueling mode, normal means of RPV level indication may not be available.

Redundant means of RPV level indication will usually be installed (including the ability to monitor level visually) to assure that the ability to monitor level will not be interrupted. However, if all level indication were to be lost during a loss of RPV inventory event, the operators would need to determine that RPV inventory loss was occurring by observing sump and tank level changes. Sump and tank level Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 89 of 274 increases must be evaluated against other potential sources of leakage such as cooling water sources inside the containment to ensure they are indicative of RCS leakage.

The 30-minute duration allows sufficient time for actions to be performed to recover inventory control equipment.

As water level in the RPV lowers, the dose rate above the core will increase. The dose rate due to this core shine should result in site specific monitor indication and possible alarm.

Post-TMI studies indicated that the installed nuclear instrumentation will operate erratically when the core is uncovered and that this should be used as a tool for making such determinations.

CGS Basis:

If RPV level monitoring capability is unavailable, all RPV level indication is unavailable and the RPV inventory loss must be detected by the leakage indications listed in Table C-1. Level increases must be evaluated against other potential sources of leakage such as cooling water sources inside the drywell to ensure they are indicative of RPV leakage. Reactor Building equipment or floor drain sump level rise may be indicative of RPV inventory losses external to the Primary Containment from systems connected to the RPV (Ref. 2, 3). With RHR System operating in the Shutdown Cooling mode, an unexplained rise in wetwell level could be indicative of RHR valve misalignment or leakage (Ref. 4). If the make-up rate to the RPV unexplainably rises above the pre-established rate, a loss of RPV inventory may be occurring even if the source of the leakage cannot be immediately identified.

Visual observation of leakage from systems connected to the RCS in areas outside the Primary Containment that cannot be isolated could be indicative of a loss of RPV inventory.

Four channels of log count rate meters are available in the Main Control Room to detect erratic source range monitor indications: Channels A-D on Panel H13-P603, Channels A & C on Panel H13-P606 and Channels B & D on Panel H13-P633 (Ref. 5).

CGS Basis Reference(s):

1. FSAR Section 7.6.1.3

2. SOP-EDR-OPS Equipment Drain System Operation

3. SOP-FDR-OPS Floor Drain System Operation

4. SOP-RHR-SDC RHR Shutdown Cooling

5. FSAR Section 7.7.1.7 Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 90 of 274 Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 3 - RPV Level Initiating Condition: Loss of RPV inventory affecting fuel clad integrity with Containment challenged EAL:

CG3.1 General Emergency RPV level LT -161 in. for GE 30 min. (Note 3)

AND Any Containment Challenge indication, Table C-2 NOTE 3: The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the condition will likely exceed the applicable time.

Table C-2 Containment Challenge Indications:;:i.

" Containment Closure no established (i.e.,

primary and secondary containment not functional)

• Explosive mixture exists inside PC (H 2 GE 6% and 02 GE 5%)

• Unplanned rise in PC pressure

• RB area radiation GT = Maximum Safe Operating level (PPM 5.3.1 Table 24)

Mode Applicability:

4 - Cold Shutdown, 5 - Refuel NEI 99-01 Basis:

This EAL represents the inability to restore and maintain RPV level to above the top of active fuel with containment challenged. Fuel damage is probable if RPV level cannot be restored, as available decay heat will cause boiling, further reducing the RPV level. With the containment breached or challenged then the potential for unmonitored fission product release to the environment is high. This represents a direct path for radioactive inventory to be released to the environment. This is consistent with the Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 91 of 274 definition of a General Emergency. The General Emergency is declared on the occurrence of the loss or imminent loss of function of all three barriers.

A number of variables can have a significant impact on heat removal capability challenging the fuel clad barrier. Examples include: initial RPV level, shutdown heat removal system design Analysis indicates that core damage may occur within an hour following continued core uncovery therefore, 30 minutes was conservatively chosen.

If Containment Closure is re-established prior to exceeding the 30 minute core uncovery time limit then escalation to GE would not occur.

CGS Basis:

When RPV level drops the top of active fuel (an indicated RPV level of -161 in.), core uncovery starts to occur (Ref. 1, 2).

Four conditions are associated with a challenge to Primary Containment (PC) integrity:

• Containment Closure is defined as the SDSP actions taken to secure primary or secondary containment and its associated structures, systems, and components as a functional barrier to fission product release under existing plant conditions. This definition is less restrictive than Technical Specification criteria governing Primary and Secondary Containment operability. If the Technical Specification criteria are met, therefore, Containment Closure has been established. (Ref. 3, 4, 5)

" Explosive (deflagration) mixtures in the primary containment are assumed to be elevated concentrations of hydrogen and oxygen. BWR industry evaluation of hydrogen generation for development of EOPs/SAGs indicates that any hydrogen concentration above minimum detectable is not to be expected within the short term. Post-LOCA hydrogen generation primarily caused by radiolysis is a slowly evolving, long-term condition. Hydrogen concentrations that rapidly develop are most likely caused by metal-water reaction. A metal-water reaction is indicative of an accident more severe than accidents considered in the plant design basis and would be indicative, therefore, of a potential threat to primary containment integrity. Hydrogen concentration of approximately 6% is considered the global deflagration concentration limit (Ref. 6).

The specified values for this threshold are the minimum global deflagration concentration limits (6% hydrogen and 5% oxygen, Ref. 5) and readily recognizable because 6% hydrogen is well above the EOP flowchart entry condition (Ref. 8). The minimum global deflagration hydrogen/oxygen concentrations (60/o/5%, respectively) require intentional primary containment venting, which is defined to be a loss of the primary containment barrier.

Atmosphere samples from a minimum of two locations inside the primary containment and one location in the suppression chamber are sequentially monitored for hydrogen and oxygen percentage levels by each of two redundant analyzer systems. The analyzers are single range (0 to 30% hydrogen and 0 to 30% oxygen). Two redundant (divisional) recorders are provided in the Main Control Room CMS-02/H2R-1 (H13-P827) and CMS-02/H2R-2 Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 92 of 274 (H13-P811). Hydrogen and oxygen concentrations can also be displayed on the plant computers (TDAS points 410 & 471 and 02 on TDAS points 395 and 456). The primary containment high hydrogen alarm is set at 3.56%. The minimum detectable primary containment hydrogen concentration is 0.6%. (Ref. 9-12)

When the monitoring systems for hydrogen or oxygen become unavailable, the concentration of these gases must be determined by manual sample and analysis. This activity is a function of Chemistry using PPM 12.17.3 to sample and analyze the primary containment atmosphere.

(Ref. 7)

" Any unplanned rise in PC pressure in the Cold Shutdown or Refueling mode indicates Containment Closure cannot be assured and the Primary Containment cannot be relied upon as a barrier to fission product release.

" RB (Reactor Building) area radiation monitors should provide indication of increased release that may be indicative of a challenge to Containment Closure. The Maximum Safe Operating levels are indicative of problems in the secondary containment that are spreading. The locations into which the primary system discharge is of concern correspond to the areas addressed in Table 24 of the EOP flowcharts (Ref. 13).

If RPV level is restored and maintained above the top of active fuel before a Containment Challenge condition occurs and subsequently a Containment Challenge condition is reached, this EAL is not met.

CGS Basis Reference(s):

1. Calculation NE-02-03-05 Attachment 3 Note 8

2. PPM 5.1.1 RPV Control

3. Technical Specifications 3.6.1.1

4. Technical Specifications 3.6.4.1

5. PPM 1.20.3 Outage Risk Management

6. BWROG EPG/SAG Revision 2, Sections PC/G

7. PPM 5.7.1 RPV and Primary Containment Flooding SAG, Table 19

8. PPM 5.2.1 Primary Containment Control

9. FSAR Section 7.5.1.5.4

10. PPM 5.0.10 Flowchart Training Manual

11. PPM 4.814.J1 814.J1 Annunciator Panel Alarms, 2-2

12. PPM 4.814.J2 814.J2 Annunciator Panel Alarms, 2-2

13. PPM 5.3.1 Secondary Containment Control Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 93 of 274 PPM 5.3.1 Table 24 - RB Area Radiation 241 RB Area Radiation MxmmSf Maximum Safe Area Instrument Alarm Operating Value (mRlir) (mR/hr) (Rihr)

East CRD Area ARM-RIS-4 104 N/A West CRD Area ARM-RIS-5 10 4 N/A H2 Recoin ber Area ARM-RIS-6 104 NA TIP Drive Area ARM-RIS-7 104 N/A SGT Filter Area ARM-RIS-8 14 N/A RHR Pump A Room ARM-RIS-9 104 NIA RHR Pump B Room ARM-RIS-10 10 N/A RHR Pump C Room ARM-RIS-il 2 104 N/A RCIC Pump Room ARM-RIS-12 04 N/A HPCS Pump Room ARM-RIS-1 3 104 NiA CRD Pump Room ARM-RIS-23 104 N/A 471 Elev - West Area ARM-RIS-24 104 N/A 471 Elev - NE ARM-RIS-32 N/A 10 501 Elev - NW ARM-RIS-33 N/A 10 Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 94 of 274 Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 3- RPV Level Initiating Condition: Loss of RPV inventory affecting fuel clad integrity with containment challenged EAL:

CG3.2 General Emergency RPV level cannot be monitored with core uncovery indicated by any of the following for GE 30 min. (Note 3):

• Any unexplained RPV leakage indication, Table C-1

• Erratic Source Range Monitor indication AND Any Containment Challenge indication, Table C-2 NOTE 3: The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the condition will likely exceed the applicable time.

Table C-I RPV Leakage Indications

• Excessive Reactor Building equipment/floor drain sump pump-out rate

• Wetwell level rise

• Observation of unisolable RCS leakage Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 95 of 274 Table C-2 Containment Challenge Indications

" Containment Closure not established (i.e.,

primary and secondary containment not functional)

• Explosive mixture exists inside PC (H2 GE 6% and 02 GE 5%)

• Unplanned rise in PC pressure

• RB area radiation GT a Maximum Safe Operating level (PPM 5.3.1 Table 24)

Mode Applicability:

4 - Cold Shutdown, 5 - Refuel NEI 99-01 Basis:

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

A number of variables can have a significant impact on heat removal capability challenging the fuel clad barrier. Examples include initial vessel level, shutdown heat removal system design Analysis indicates that core damage may occur within an hour following continued core uncovery therefore, 30 minutes was conservatively chosen.

If Containment Closure is re-established prior to exceeding the 30 minute core uncovery time limit then escalation to General Emergency would not occur.

Sump and tank 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.

Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 96 of 274 In the cold shutdown mode, normal RPV level and RPV level instrumentation systems will usually be available. In the refueling mode, normal means of RPV level indication may not be available.

Redundant means of RPV level indication will usually be installed (including the ability to monitor level visually) to assure that the ability to monitor level will not be interrupted. However, if all level indication were to be lost during a loss of RPV inventory event, the operators would need to determine that RPV inventory loss was occurring by observing sump and tank level changes. Sump and tank 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.

As water level in the RPV lowers, the dose rate above the core will increase. The dose rate due to this core shine should result in site specific monitor indication and possible alarm.

Post-TMI studies indicated that the installed nuclear instrumentation will operate erratically when the core is uncovered and that this should be used as a tool for making such determinations.

CGS Basis:

If RPV level monitoring capability is unavailable, all RPV level indication is unavailable and the RPV inventory loss must be detected by erratic source range monitor indications and the leakage indications listed in Table C-1.

Four channels of log count rate meters are available in the Main Control Room to detect erratic source range monitor indications: Channels A-D on Panel H13-P603, Channels A & C on Panel H13-P606 and Channels B & D on Panel H13-P633 (Ref. 1).

Reactor Building equipment or floor drain sump level rise may be indicative of RPV inventory losses external to the Primary Containment from systems connected to the RPV (Ref. 3, 4). With RHR System operating in the Shutdown Cooling mode, an unexplained rise in wetwell level could be indicative of RHR valve misalignment or leakage (Ref. 5). If the make-up rate to the RPV unexplainably rises above the pre-established rate, a loss of RPV inventory may be occurring even if the source of the leakage cannot be immediately identified. Visual observation of leakage from systems connected to the RCS in areas outside the Primary Containment that cannot be isolated could be indicative of a loss of RPV inventory.

Four conditions are associated with a challenge to Primary Containment (PC) integrity:

Containment Closure is defined as the SDSP actions taken to secure primary or secondary containment and its associated structures, systems, and components as a functional barrier to fission product release under existing plant conditions. This definition is less restrictive than Technical Specification criteria governing Primary and Secondary Containment operability. If the Technical Specification criteria are met, therefore, Containment Closure has been established. (Ref. 6, 7, 8)

Explosive (deflagration) mixtures in the primary containment are assumed to be elevated concentrations of hydrogen and oxygen. BWR industry evaluation of hydrogen generation for development of EOPs/SAGs indicates that any hydrogen concentration above minimum detectable is not to be expected within the short term. Post-LOCA hydrogen generation primarily caused by radiolysis is a slowly evolving, long-term condition. Hydrogen concentrations that rapidly develop are most likely caused by metal-water reaction. A metal-water reaction is indicative of an accident more severe than accidents considered in the plant design basis and would be indicative, therefore, of a potential threat to primary containment Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 97 of 274 integrity. Hydrogen concentration of approximately 6% is considered the global deflagration concentration limit (Ref. 9).

The specified values for this threshold are the minimum global deflagration concentration limits (6% hydrogen and 5% oxygen, Ref. 10) and readily recognizable because 6% hydrogen is well above the EOP flowchart entry condition (Ref. 11). The minimum global deflagration hydrogen/oxygen concentrations (60/o/5%, respectively) require intentional primary containment venting, which is defined to be a loss of the primary containment barrier.

Atmosphere samples from a minimum of two locations inside the primary containment and one location in the suppression chamber are sequentially monitored for hydrogen and oxygen percentage levels by each of two redundant analyzer systems. The analyzers are single range (0 to 30% hydrogen and 0 to 30% oxygen). Two redundant (divisional) recorders are provided in the Main Control Room CMS-02/H2R-1 (H13-P827) and CMS-O2/H2R-2 (H13-P811). Hydrogen and oxygen concentrations can also be displayed on the plant computers (TDAS points 410 & 471 and 02 on TDAS points 395 and 456). The primary containment high hydrogen alarm is set at 3.56%. The minimum detectable primary containment hydrogen concentration is 0.6%. (Ref. 12-15)

When the monitoring systems for hydrogen or oxygen become unavailable, the concentration of these gases must be determined by manual sample and analysis. This activity is a function of Chemistry using PPM 12.17.3 to sample and analyze the primary containment atmosphere.

(Ref. 11)

" Any unplanned rise in PC pressure in the Cold Shutdown or Refueling mode indicates Containment Closure cannot be assured and the Primary Containment cannot be relied upon as a barrier to fission product release.

• RB (Reactor Building) area radiation monitors should provide indication of increased release that may be indicative of a challenge to Containment Closure. The Maximum Safe Operating levels are indicative of problems in the secondary containment that are spreading. The locations into which the primary system discharge is of concern correspond to the areas addressed in Table 24 of the EOP flowcharts (Ref.16).

CGS Basis Reference(s):

1. FSAR Section 7.7.1.7

2. FSAR Section 7.6.1.3

3. SOP-EDR-OPS Equipment Drain System Operation 4 SOP-FDR-OPS Floor Drain System Operation

5. SOP-RHR-SDC RHR Shutdown Cooling

6. Technical Specifications 3.6.1.1 7 Technical Specifications 3.6.4.1

8. PPM 1.20.3 Outage Risk Management

8. BWROG EPG/SAG Revision 2, Sections PC/G

10. PPM 5.7.1 RPV and Primary Containment Flooding SAG, Table 19

11. PPM 5.2.1 Primary Containment Control

12. FSAR Section 7.5.1.5.4

13. PPM 5.0.10 Flowchart Training Manual

14. PPM 4.814.J1 814.J1 Annunciator Panel Alarms, 2-2

15. PPM 4.814.J2 814.J2 Annunciator Panel Alarms, 2-2 Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 98 of 274

16. PPM 5.3.1 Secondary Containment Control Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 99 of 274 PPM 5.3.1 Table 24 - RB Area Radiation 241 R8 Area Radiation MxmmSf Maximum Sate Area Instrument Alarm Operating Value (mR/hr) (mRJhr) (Rihr)

East CRD Area ARM-RISA4 104 N/A West CRD Area ARM-RIS-5 104 N/A H2 Recomber Area ARM-RIS-6 104 N/A TIP Drive Area ARM-RIS-7 10`4 N/A SGT Filter Area ARM-RIS-8 4 N/A

<10 RHR Pump A Room ARM-RIS-9 104 N/A RHR Pump B Room ARM-RIS-10 0. 6 4 N/A RHR Pump C Room ARM-RIS-l, 104 N/A RCOC Pump Room ARM-RIS- 2 104 N/A HPCS Pump Room ARM-RIS-13 104 N/A CRD Pump Room ARM-RIS-23 "14 N/A 471 Elev - West Area ARM-RIS-24 104 NtA 471 Elev - NE ARM-RIS-32 N/A 10 501 Elev - NW ARM-RIS-33 f N/A 10 Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 100 of 274 Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 4 - RCS Temperature Initiating Condition: Unplanned loss of decay heat removal capability with irradiated fuel in the RPV EAL:

CU4.1 Unusual Event UNPLANNED event results in RCS temperature exceeding the Technical Specification cold shutdown temperature limit (2000 F) due to loss of decay heat removal capability Mode Applicability:

4 - Cold Shutdown, 5 - Refuel NEI 99-01 Basis:

This EAL is be a precursor of more serious conditions and, as a result, is considered to be a potential degradation of the level of safety of the plant. In cold shutdown the ability to remove decay heat relies primarily on forced cooling flow. Operation of the systems that provide this forced cooling may be jeopardized due to the unlikely loss of electrical power or RCS inventory. Since the RCS usually remains intact in the cold shutdown mode a large inventory of water is available to keep the core covered.

During refueling the level in the RPV will normally be maintained above the RPV flange. Refueling evolutions that decrease water level below the RPV flange are carefully planned and procedurally controlled. Loss of forced decay heat removal at reduced inventory may result in more rapid increases in RPV temperatures depending on the time since shutdown.

Escalation to Alert would be via CA3.1 based on an inventory loss or CA4.1 based on exceeding its temperature criteria.

CGS Basis:

The Technical Specification cold shutdown temperature limit is 200°F (Ref. 1).

Recirculation suction temperature, RRC-TR-650 pt 1(Ref 2), is the primary temperature measurement instrument when RPV pressure is less than 100 psig and the associated RRC pump is operating. It is the alternate method when RPV pressure is greater than or equal to 100 psig. RPV pressure is the primary temperature measurement when above 100 psig and MSPI 9 is the preferred pressure monitoring tool.

Monitoring of the RWCU bottom head drain temperature element, RWCU-TE-21, as read on RWCU-TI-607 pt 5 (H13 P602) or MS-TR-6 pt 316 (RB 522) is acceptable only if a RRC pump is operating for forced flow and RWCU flow of greater than 50 gpm exists. (Ref. 2)

Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 101 of 274 With flow through the RHR Heat Exchanger, the inlet temperature (TDAS pt. X045) is indicative of RRC system temperature. If adequate core flow cannot be provided, RPV metal temperature can be monitored on MS-TR-6. (Ref. 3)

If RCS temperature exceeds 200 0 F, an operating mode change occurs. Although the event may have originated in cold conditions, the emergency classification shall be based on the operating mode that existed at the time the event occurred (prior to any protective system or operator action initiated in response to the condition). For events that occur in Cold Shutdown or Refuel, escalation is via EALs that have Cold Shutdown or Refuel for mode applicability, even if Hot Shutdown (or a higher mode) is entered during any subsequent heat-up. In particular, the fission product barrier EALs are applicable only to events that initiate in Hot Shutdown or higher.

CGS Basis Reference(s):

1. Technical Specifications Table 1.1-1

2. OSP-RCS-C102 RPV Non-Critical Cooldown Surveillance

3. SOP-RHR-SDC RHR Shutdown Cooling Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 102 of 274 Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 4 - RCS Temperature Initiating Condition: Unplanned loss of decay heat removal capability with irradiated fuel in the RPV EAL:

CU4.2 Unusual Event Loss of all RCS temperature and RPV level indication for GE 15 min. (Note 3)

NOTE 3: The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the condition will likely exceed the applicable time.

Mode Applicability:

4 - Cold Shutdown, 5 - Refuel NEI 99-01 Basis:

This EAL is be a precursor of more serious conditions and, as a result, is considered to be a potential degradation of the level of safety of the plant. In cold shutdown the ability to remove decay heat relies primarily on forced cooling flow. Operation of the systems that provide this forced cooling may be jeopardized due to the unlikely loss of electrical power or RCS inventory. Since the RCS usually remains intact in the cold shutdown mode a large inventory of water is available to keep the core covered.

During refueling the level in the RPV will normally be maintained above the RPV flange. Refueling evolutions thiat decrease water level below the RPV flange are carefully planned and procedurally controlled. Loss of forced decay heat removal at reduced inventory may result in more rapid increases in RPV temperatures depending on the time since shutdown.

Normal means of core temperature indication and RCS level indication may not be available in the refueling mode. Redundant means of RPV level indication are therefore procedurally installed to assure that the ability to monitor level will not be interrupted. However, if all level and temperature indication were to be lost in either the cold shutdown of refueling modes, this EAL would result in declaration of a UE if both temperature and level indication cannot be restored within 15 minutes from the loss of both means of indication. Escalation to Alert would be via CA3.1 based on an inventory loss or CA4.1 based on exceeding its temperature criteria.

CGS Basis:

RPV water level is normally monitored using the instruments in Figure C-3 (Ref. 1, 2, 3).

Recirculation suction temperature, RRC-TR-650 pt 1(2), is the primary temperature measurement instrument when RPV pressure is less than 100 psig and the associated RRC pump is operating. It is the alternate method when RPV pressure is greater than or equal to 100 psig. RPV pressure is the Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 103 of 274 primary temperature measurement when above 100 psig and MS-PI-9 is the preferred pressure monitoring tool.

Monitoring of the RWCU bottom head drain temperature element, RWCU-TE-21, as read on RWCU-TI-607 pt 5 (H13 P602) or MS-TR-6 pt 316 (RB 522) is acceptable only if a RRC pump is operating for forced flow and RWCU flow of greater than 50 gpm exists. (Ref. 4)

With flow through the RHR Heat Exchanger, the inlet temperature (TDAS pt. X045) is indicative of RRC system temperature. If adequate core flow cannot be provided, RPV metal temperature can be monitored on MS-TR-6. (Ref. 5)

If RCS temperature exceeds 200 0 F, an operating mode change occurs. Although the event may have originated in cold conditions, the emergency classification shall be based on the operating mode that existed at the time the event occurred (prior to any protective system or operator action initiated in response to the condition). For events that occur in Cold Shutdown or Refuel, escalation is via EALs that have Cold Shutdown or Refuel for mode applicability, even if Hot Shutdown (or a higher mode) is entered during any subsequent heat-up. In particular, the fission product barrier EALs are applicable only to events that initiate in Hot Shutdown or higher.

CGS Basis Reference(s):

1. FSAR Table 7.5-1

2. FSAR Figure 7.7-1

3. FSAR Section 7.6.1.3

4. OSP-RCS-C102 RPV Non-Critical Cooldown Surveillance

5. SOP-RHR-SDC RHR Shutdown Cooling Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 104 of 274 Figure C-3 RPV Levels (Ref. 1, 2, 3)

!Ii LEVEL HE3GHTABOVE TRIPPOINt NO. VESSELZERO (INOCATED) 56582 + 45'L (7) 568 +4(L5 I#1

- ~aa --

'Mi

-110'

- 161.2

- 310' Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 105 of 274 Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 4 - RCS Temperature Initiating Condition: Inability to maintain plant in cold shutdown EAL:

CA4.1 Alert An UNPLANNED event results in RCS temperature GT 2001F for GT Table C-3 duration OR An UNPLANNED event results in RPV pressure increase GT 10 psi due to a loss of RCS cooling Table C-3; RCS Reheat Duration Thresholds

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

1. RCS intact (Containment Closure N/A) 60 min.*

2. Containment Closure established (i.e., primary or secondary containment functional) 20 min.*

AND RCS not intact

3. Containment Closure not established (i.e.,

primary and secondary containment not functional) 0 min.

AND RCS not intact Mode Applicability:

4 - Cold Shutdown, 5 - Refuel NEI 99-01 Basis:

1st Condition The RFS Reheat Duration Threshold table addresses complete loss of functions required for core cooling for greater than 60 minutes during refueling and cold shutdown modes when RCS integrity is Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 106 of 274 established. RCS integrity should be considered to be in place when the RCS pressure boundary is in its normal condition for the cold shutdown mode of operation (e.g., no freeze seals or nozzle dams).

The status of Containment Closure in this condition is immaterial given that the RCS is providing a high pressure barrier to fission product release to the environment. The 60 minute time frame should allow sufficient time to restore cooling without there being a substantial degradation in plant safety.

The RCS Reheat Duration Threshold table also addresses the complete loss of functions required for core cooling for greater than 20 minutes during refueling and cold shutdown modes when Containment Closure is established but RCS integrity is not established. The allowed 20 minute time frame was included to allow operator action to restore the heat removal function, if possible.

Finally, complete loss of functions required for core cooling during refueling and cold shutdown modes when neither Containment Closure nor RCS integrity are established. RCS integrity is in place when the RCS pressure boundary is in its normal condition for the cold shutdown mode of operation (e.g.,

no freeze seals or nozzle dams). No delay time is allowed because the evaporated reactor coolant that may be released into the Containment during this heatup condition could also be directly released to the environment.

The note (*) in Table C-3 indicates that this EAL is not applicable if actions are successful in restoring an RCS heat removal system to operation and RCS temperature is being reduced within the specified time frame.

2nd Condition The 10 psi pressure increase addresses situations where, due to high decay heat loads, the time provided to restore temperature control, should be less than 60 minutes. The RCS pressure setpoint chosen should be 10 psi or the lowest pressure that the site can read on installed Control Board instrumentation that is equal to or greater than 10 psi.

Escalation to Site Area Emergency would be via CS3.1, CS3.2 or CS3.3 should boiling result in significant RPV level loss leading to core uncovery.

A loss of Technical Specification components alone is not intended to constitute an Alert. The same is true of a momentary unplanned excursion above the Technical Specification cold shutdown temperature limit when the heat removal function is available.

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

CGS Basis:

200OF is the Technical Specification cold shutdown temperature limit (Ref. 1).

10 psi is one-half of the 20 psi minor division on the Wide Range RPV pressure instrument, RFW-PI-605, on Main Control Room Panel H13- P603 (Ref. 2). This instrument has a range of 0 to 1200 psig. This RPV pressure indication is also displayed on plant computer point B016 (Ref. 3).

Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 107 of 274 Recirculation suction temperature, RRC TR 650 pt 1(2), is the primary temperature measurement instrument when RPV pressure is less than 100 psig and the associated RRC pump is operating. It is the alternate method when RPV pressure is greater than or equal to 100 psig. RPV pressure is the primary temperature measurement when above 100 psig and MSPI 9 is the preferred pressure monitoring tool.

Monitoring of the RWCU bottom head drain temperature element, RWCU TE 21, as read on RWCU TI 607 pt 5 (H13 P602) or MS TR 6 pt 316 (RB 522) is acceptable only if a RRC pump is operating for forced flow and RWCU flow of greater than 50 gpm exists. (Ref. 4)

With flow through the RHR Heat Exchanger, the inlet temperature (TDAS pt. X045) is indicative of RRC system temperature. If adequate core flow cannot be provided, RPV metal temperature can be monitored on MS TR 6. (Ref. 5)

Containment Closure is defined as the SDSP actions taken to secure primary or secondary containment and its associated structures, systems, and components as a functional barrier to fission product release under existing plant conditions. This definition is less restrictive than Technical Specification criteria governing Primary and Secondary Containment operability. If the Technical Specification criteria are met, therefore, Containment Closure has been established. (Ref. 6, 7, 8)

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).

If RCS temperature exceeds 200'F, an operating mode change occurs. Although the event may have originated in cold conditions, the emergency classification shall be based on the operating mode that existed at the time the event occurred (prior to any protective system or operator action initiated in response to the condition). For events that occur in Cold Shutdown or Refuel, escalation is via EALs that have Cold Shutdown or Refuel for mode applicability, even if Hot Shutdown (or a higher mode) is entered during any subsequent heat-up. In particular, the fission product barrier EALs are applicable only to events that initiate in Hot Shutdown or higher.

CGS Basis Reference(s):

1. Technical Specifications Table 1.1-1

2. Instrument Master Datasheet for EPN RFW-PI-605

3. PPM 10.27.36 Reactor Pressure High Alarm - CC

4. OSP-RCS-C102 RPV Non-Critical Cooldown Surveillance

5. SOP-RHR-SDC RHR Shutdown Cooling

6. Technical Specifications 3.6.1.1

7. Technical Specifications 3.6.4.1

8. PPM 1.20.3 Outage Risk Management Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 108 of 274 Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 5 - Communications Initiating Condition: Loss of all onsite or offsite communications capabilities EAL:

CU5.1 Unusual Event Loss of all Table C-4 onsite communication methods affecting the ability to perform routine operations OR Loss of all Table C-4 offsite communication methods affecting the ability to perform offsite notifications Table C-4 Communications Systems:

System Onsite (external)

(internal) Offsite Plant Public Address (PA) System X Plant Telephone System X Plant Radio System Operations and X Security Channels Offsite calling capability from the X Control Room via direct telephone and fax lines Long distance calling capability on X the commercial phone system Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 109 of 274 Mode Applicability:

4 - Cold Shutdown, 5 - Refuel, D - Defueled NEI 99-01 Basis:

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

The availability of one method of ordinary off-site communications is sufficient to inform federal, state, and local authorities of plant issues. This EAL is intended to be used only when extraordinary means (e.g., relaying of information from radio transmissions, individuals being sent to off-site locations, etc.)

are being utilized to make communications possible.

CGS Basis:

Onsite and offsite communications include one or more of the systems listed in Table C-4 (Ref. 1, 2).

Public Address (PA) System The public address system provides a way of contacting personnel in the various buildings of the plant and locations of the site that might be inaccessible using other means of communication. The building-wide alarm system alerts (via the public address system speakers) operating personnel to fire hazards and other trouble conditions for which plant management finds it necessary to alert plant personnel.

Plant Telephone System This system consists of interconnections to the public telephone network (and trunks to the PBX) with individual direct lines that provide inward and outward dialing access to most plant locations.

Plant Radio System Operations and Security Channels The radio communications system is used for communications with personnel involved in maintenance and security in and around the plant complex by means of hand-held portable radio units, mobile radio units, and paging receivers. The telephone link to BPA provides a direct communication link to the BPA Dittmer Control Center. The radio communications system provides a communications link for security and emergency communications to local law enforcement agencies and emergency control centers.

Offsite callinq capability from the Control Room via direct telephone and fax lines This communications method includes following dedicated phone networks that are available for emergency communications in addition to the normal Energy Northwest phone network:

• Energy Northwest Emergency Center Network

• Response Agency Network Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 110 of 274 0 NRC Emergency Notification System Various locations such as the Control Room, Technical Support Center, Emergency Operations Facility, Joint Information Center, Department of Energy-RL, Washington State Emergency Operations Center, Oregon State Emergency Coordination Center and the Benton and Franklin County Emergency Operations Centers have facsimile transceivers. The facsimile transceivers enable the transmission and receipt of printed material. The facsimile system which connects the Energy Northwest emergency centers with the county and state emergency centers uses dedicated phone lines.

Lona distance calling capability on the commercial phone system The Energy Northwest Richland phone system is a computer based, software controlled telephone exchange (Computerized Branch Exchange). It is equipped with redundant computerized processor units and is served by an uninterruptible power supply. The direct-dial private telephone system provides communication between the Energy Northwest facilities. The phone system is arranged such that plant telephones can reach other Energy Northwest facilities by direct-dialing and without the need of an operator.

This EAL is the cold condition equivalent of the hot condition EAL SU6.1.

CGS Basis Reference(s):

1. Emergency Plan Section 6.6

2. FSAR Section 9.5.2 Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 111 of 274 Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 6 - Inadvertent Criticality Initiating Condition: Inadvertent criticality EAL:

CU6.1 Unusual Event UNPLANNED sustained positive period observed on nuclear instrumentation Mode Applicability:

4 - Cold Shutdown, 5 - Refuel NEI 99-01 Basis:

This EAL addresses criticality events that occur in Cold Shutdown or Refueling modes such as fuel mis-loading events and inadvertent dilution events. This IC indicates a potential degradation of the level of safety of the plant, warranting a UE classification.

Escalation would be by Emergency Director Judgment.

CGS Basis:

This condition may be identified by (Ref. 1):

• Period meters A, B, C, and D on Main Control Room panel H13-P603 and H13-P633

• Amber period lights for each of the four channels on Main Control Room panel H13-P603 and H13-P633

• Annunciator P603-A7 SRM PERIOD FAST which is actuated at +60 seconds by any one of the four SRM channels (Ref. 2)

CGS Basis Reference(s):

1. FSAR Section 7.7.1.7

2. PPM 4.603.A7 Annunciator Response 5-6, SOURCE RANGE MONITORS FAST PERIOD Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 112 of 274 Category H - Hazards & Other Conditions Affectinq 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.

The events of this category pertain to the following subcategories:

1. Natural & Destructive Phenomena Natural events include hurricanes, earthquakes or tornados that have potential to cause plant structure or equipment damage of sufficient magnitude to threaten personnel or plant safety. Non-naturally occurring events that can cause damage to plant facilities and include aircraft crashes, missile impacts, etc.

2. Fire or Explosion Fires can pose significant hazards to personnel and reactor safety. Appropriate for classification are fires within the site Protected Area or which may affect operability of vital equipment.

3. Hazardous Gas Non-naturally occurring events that can cause damage to plant facilities and include toxic, corrosive, asphyxiant, or flammable gas leaks.

4. Security Unauthorized entry attempts into the Protected Area, bomb threats, sabotage attempts, and actual security compromises threatening loss of physical control of the plant.

5. 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.

6. Judgment The EALs defined in other categories specify the predetermined symptoms or events that are indicative of emergency or potential emergency conditions and thus warrant classification. While these EALs have been developed to address the full spectrum of possible emergency conditions which may warrant classification and subsequent implementation of the Emergency Plan, a provision for classification of emergencies based on operator/management experience and judgment is still necessary. The EALs of this category provide the Emergency Director the latitude to classify emergency conditions consistent with the established classification criteria based upon Emergency Director judgment.

Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 113 of 274

7. 618-11 Area Events that are indicative of fires, explosions or other events that may cause a hazardous release from the 618-11 Waste Burial Ground.

Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 114 of 274 Category: H - Hazards & Other Conditions Affecting Plant Safety Subcategory: 1 - Natural & Destructive Phenomena Initiating Condition: Natural or destructive phenomena affecting the Protected Area EAL:

HU1.1 Unusual Event Seismic event identified by any two of the following:

" Earthquake felt in plant

" H13.P851.S1.2-5 (MINIMUM SEISMIC EARTHQUAKE EXCEEDED) activated

" National Earthquake Information Center (Note 4)

NOTE 4: The NEIC can be contacted by calling (303) 273-8500. Select option #1 and inform the analyst you wish to confirm recent seismic activity in the vicinity of Columbia Generating Station. Provide the analyst with the following CGS coordinates: 460 28' 18" (46.4720) north latitude, 1190 19' 58" (119.3330) west longitude.

Mode Applicability:

All NEI 99-01 Basis:

This EAL is categorized on the basis of the occurrence of an event of sufficient magnitude to be of concern to plant operators.

Damage may be caused to some portions of the site, but should not affect ability of safety functions to operate.

As defined in the EPRI-sponsored Guidelines for Nuclear Plant Response to an Earthquake, dated October 1989, a "felt earthquake" is: An earthquake of sufficient intensity such that: (a) the vibratory ground motion is felt at the nuclear plant site and recognized as an earthquake based on a consensus of control room operators on duty at the time, and (b) for plants with operable seismic instrumentation, the seismic switches of the plant are activated.

The National Earthquake Information Center can confirm if an earthquake has occurred in the area of the plant.

CGS Basis:

An earthquake felt "in plant" pertains to an earthquake felt within the CGS power block.

CGS seismic instrumentation consists of a Kinemetrics SMA-3 Strong Motion Accelerograph and associated sensors that are equipped with seismic triggers set to initiate recording at an acceleration equal to or exceeding 0.01 g (Ref. 2). Three time-history triaxial acceleration sensors are provided.

Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 115 of 274 These sensors transmit electrical signals to the main Control Room where they are recorded on magnetic tape. A playback unit is provided in the main Control Room immediately below the magnetic tape recorders. The operator can obtain a visual record on paper tape by withdrawing the magnetic cassette from the recorder and inserting it into the playback unit. One time-history triaxial acceleration sensor is located in the basement of the reactor building on the foundation. Another is located at a higher elevation on reactor building floor el. 522 ft. The time-history records are provided to facilitate the dynamic analysis of the response of the structure following an earthquake. The triaxial time-history recorders are put into operation by a seismic trigger unit located in the reactor building. A seismic switch unit that is similar to the seismic trigger unit is also provided. The trip point of the seismic switch unit is set at the maximum acceleration corresponding to the OBE, and it provides immediate Control Room annunciation that the OBE has been exceeded requiring declaration of an Alert (Ref. 1)

This also annunciates the seismic activity alarm H1i3.P851 .S1.2-5 Minimum Seismic Earthquake Exceeded. (Ref. 2, 3, 4)

The National Earthquake Information Center (NEIC) can confirm seismic activity in the vicinity of the CGS. The NEIC can be contacted by calling (303) 273-8500. Select option #1 and inform the analyst you wish to confirm recent seismic activity in the vicinity of Columbia Generating Station. Provide the analyst with the following CGS coordinates: 460 28' 18" (46.4720) north latitude, 1190 19' 58" (119.3330) west longitude (Ref. 5).

Alternatively go to the USGS NEIC website:

http://earthquake. usgs.gov/eqcenter/

Click on 'Northwest' from the earthquake map index. The maps are updated within 5 min. of a measured earthquake.

This event escalates to an Alert under EAL HA1.1 if the earthquake exceeds Operating Basis Earthquake (OBE) levels.

CGS Basis Reference(s):

1. CGS FSAR Section 3.7.4 Seismic Instrumentation

2. ISP-SEIS-M201 Seismic Systems Channel Check

3. PPM 4.8513.1.2-5 Minimum Seismic Earthquake Exceeded

4. ABN-EARTHQUAKE Earthquake

5. FSAR Section 2.1.1 Specification of Location Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 116 of 274 Category: H - Hazards & Other Conditions Affecting Plant Safety Subcategory: 1 - Natural & Destructive Phenomena Initiating Condition: Natural or destructive phenomena affecting the Protected Area EAL:

HU1.2 Unusual Event Tornado striking within PROTECTED AREA boundary or sustained high winds GT 70 mph Mode Applicability:

All NEI 99-01 Basis:

This EAL is categorized on the basis of the occurrence of an event of sufficient magnitude to be of concern to plant operators.

This EAL is based on a tornado striking (touching down) or high winds within the Protected Area.

Escalation of this emergency classification level, if appropriate, would be based on visible damage, or by other in plant conditions, via HA1.2.

CGS Basis:

If damage is confirmed visually or by other in-plant indications, the event may be escalated to an Alert under EAL HA1.2.

Columbia Generating Station is designed to be safely shutdown following extreme high winds or a tornado. However, some of the systems and structures are not analyzed to function after extreme conditions (SW spray trees, CW cooling towers). Therefore, enhanced plant monitoring is warranted during severe wind conditions.

Wind advisories are issued by the National Warning Alert System or Hanford Meteorology. The on-site 15 minute average wind speed at 33' above grade elevation is indicated at MET-WSR-4 (Red Pen) on H13-P823 (Board L) (Ref. 3).

The design basis wind velocity for all Category 1 structures at CGS is 100 mph based on a 10 second average (Ref. 1). However, the Control Room wind speed recorder was modified to provide 15 min.

averages. The EAL threshold reflects a de-rated design wind speed of 70 mph to compensate for gust factor based on the extended wind speed indicator averaging (Ref. 2).

A tornado striking (touching down) within the Protected Area warrants declaration of an Unusual Event regardless of the measured wind speed at the meteorological tower. A tornado is defined as a violently rotating column of air in contact with the ground and extending from the base of a thunderstorm.

Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 117 of 274 Wind speed, wind direction and temperature readings are provided instantaneously and in 15 minute averages. The term "sustained" is intended to mean the 15 minute average wind speed indicator.

The Protected Area boundary is defined in Civil Drawing C897 Columbia Generating Station Site Plan (Ref. 4).

CGS Basis Reference(s):

1. FSAR Section 3.1.1 Wind Loadings

2. CGS Calculation CALC CE-02-93-16 Evaluate PMR/BDC 98-0131-OA change from 5 Min. to 15 min. averaging of 33 ft. elev. met twr. wind speeds for UE and Alert declarations

3. ABN-WIND Tornado/High Wind

4. Civil Drawing C897 Columbia Generating Station Site Plan Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 118 of 274 Category: H - Hazards & Other Conditions Affecting Plant Safety Subcategory: 1 - Natural & Destructive Phenomena Initiating Condition: Natural or destructive phenomena affecting the Protected Area EAL:

HU1.3 Unusual Event Turbine failure resulting in EITHER:

Casing penetration OR Damage to turbine or generator seals Mode Applicability:

All NEI 99-01 Basis:

This EAL is categorized on the basis of the occurrence of an event of sufficient magnitude to be of concern to plant operators.

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

Of major concern is the potential for leakage of combustible fluids (lubricating oils) and gases (hydrogen cooling) to the plant environs. Actual fires and flammable gas build up are appropriately classified via HU2.1 and HU3.1.

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

Escalation of this emergency classification level, if appropriate, would be to HA1.3 based on damage done by projectiles generated by the failure or by the radiological releases.

CGS Basis:

None CGS Basis Reference(s):

None Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 119 of 274 Category: H - Hazards & Other Conditions Affecting Plant Safety Subcategory: 1 - Natural & Destructive Phenomena Initiating Condition: Natural or destructive phenomena affecting the Protected Area EAL:

HU1.4 Unusual Event Internal flooding that has the potential to affect safe shutdown systems or components required by Technical Specifications for the current operating mode in any Table H-1 plant structure.

Table H-i Plant Structures Containing Safe Table H-1 Plant Structures Containing Safe Shutdown Systems or Components

• Reactor Building

• Turbine Building

• Radwaste/Control Building

• Standby Service Water Pump Houses

• Diesel Generator Building

• Diesel Fuel Oil Storage Area Mode Applicability:

All NEI 99-01 Basis:

This EAL is categorized on the basis of the occurrence of an event of sufficient magnitude to be of concern to plant operators.

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

Escalation of this emergency classification level, if appropriate, would be based visible damage via HA1.4, or by other plant conditions.

CGS Basis:

The Table H-1 Plant Structures Containing Safe Shutdown Systems or Components include those structures that contain any Class 1, 2 or 3 SSC (Ref. 1).

CGS Basis Reference(s):

Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 120 of 274

1. FSAR Table 3.2-1 Equipment Classification Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft I - Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 121 of 274 Category: H - Hazards & Other Conditions Affecting Plant Safety Subcategory: 1 - Natural & Destructive Phenomena Initiating Condition: Natural or destructive phenomena affecting the Protected Area EAL:

HU1.5 Unusual Event Range FIRE impeding access to the site OR Volcanic ash fallout requiring plant shutdown Mode Applicability:

All NEI 99-01 Basis:

This EAL is categorized on the basis of the occurrence of an event of sufficient magnitude to be of concern to plant operators.

This EAL addresses other site specific phenomena that can also be precursors of more serious events.

CGS Basis:

Columbia Generating Station is located on a dry land steppe. Range fires routinely occur in this type of environment. An example of an event that may reduce the level of plant safety is a range fire that affects the ability of plant personnel to respond if required to report to the site (Ref. 1).

The potential for volcanic eruption exists in the Pacific Northwest. Heavy ash fall, such as that experienced at certain locations following the eruption of Mt. St. Helens in 1980, could affect operation of plant equipment if precautionary measures are not taken. The design basis ash fall is projected for a 20 hour2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> duration. Plant shutdown may be warranted, based on several individual criteria specified in ABN-ASH (Ref. 2). This threshold is met when ABN-ASH requires plant shutdown.

Either a range fire impeding site access or volcanic ash fallout requiring plant shutdown pose a potential degradation to the level of safety of the plant.

CGS Basis Reference(s):

1. ABN-FIRE Fire

2. ABN-ASH Ash Fall Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft I Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 122 of 274 Category: H - Hazards & Other Conditions Affecting Plant Safety Subcategory: Natural & Destructive Phenomena Initiating Condition: Natural or destructive phenomena affecting vital areas EAL:

HA1.1 Alert Seismic event GT Operating Basis Earthquake (0.125 g) as indicated by H13.P851.S1.5-1 (OPERATING BASIS EARTHQUAKE EXCEEDED) activated AND Earthquake confirmed by any of the following:

• Earthquake felt in plant

• National Earthquake Information Center (Note 4)

• Control Room indication of degraded performance of safe shutdown systems NOTE 4: The NEIC can be contacted by calling (303) 273-8500. Select option #1 and inform the analyst you wish to confirm recent seismic activity in the vicinity of Columbia Generating Station. Provide the analyst with the following CGS coordinates: 460 28' 18" (46.4720) north latitude, 1190 19' 58" (119.3330) west longitude.

Mode Applicability:

All NEI 99-01 Basis:

Escalation of this emergency classification level, if appropriate, would be based on System Malfunction EALs.

Seismic events of this magnitude can result in a vital area being subjected to forces beyond design limits, and thus damage may be assumed to have occurred to plant safety systems.

The National Earthquake Information Center can confirm if an earthquake has occurred in the area of the plant.

CGS Basis:

Ground motion acceleration of 0.125g is the Operating Basis Earthquake for CGS (Ref. 1,4).

CGS seismic instrumentation consists of a Kinemetrics SMA-3 Strong Motion Accelerograph and associated sensors that are equipped with seismic triggers set to initiate recording at an acceleration equal to or exceeding 0.01 g (Ref 2). Three time-history triaxial acceleration sensors are provided.

These sensors transmit electrical signals to the main Control Room where they are recorded on Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 123 of 274 magnetic tape. A playback unit is provided in the main Control Room immediately below the magnetic tape recorders. The operator can obtain a visual record on paper tape by withdrawing the magnetic cassette from the recorder and inserting it into the playback unit. One time-history triaxial acceleration sensor is located in the basement of the reactor building on the foundation. Another is located at a higher elevation on reactor building floor el. 522 ft. The time-history records are provided to facilitate the dynamic analysis of the response of the structure following an earthquake. The triaxial time-history recorders are put into operation by a seismic trigger unit located in the reactor building. A seismic switch unit that is similar to the seismic trigger unit is also provided. The trip point of the seismic switch unit is set at the maximum acceleration corresponding to the OBE, and it provides immediate Control Room annunciation (Hi3.P851.S1.5-1 Operating Basis Earthquake Exceeded) that the OBE has been exceeded (Ref. 1, 3).

As defined in the EPRI-sponsored Guidelines for Nuclear Plant Response to an Earthquake, dated October 1989, a "felt earthquake" is: An earthquake of sufficient intensity such that: (a) the vibratory ground motion is felt at the nuclear plant site and recognized as an earthquake based on a consensus of Control Room operators on duty at the time, and (b) for plants with operable seismic instrumentation, the seismic switches of the plant are activated.

An earthquake felt "in plant" pertains to an earthquake felt within the CGS power block.

The National Earthquake Information Center (NEIC) can confirm seismic activity in the vicinity of the CGS. The NEIC can be contacted by calling (303) 273-8500. Select option #1 and inform the analyst you wish to confirm recent seismic activity in the vicinity of Columbia Generating Station. Provide the analyst with the following CGS coordinates: 460 28' 18" (46.4720) north latitude, 1190 19' 58" (119.3330) west longitude (Ref. 5). Alternatively go to the USGS NEIC website:

http://earthquakeusgs.gov/eqcenter/

Click on 'Northwest' from the earthquake map index. The maps are updated within 5 min. of a measured earthquake.

A time interval is not associated with this EAL and degraded safe shutdown system performance may not be immediately apparent. Assessment of degraded safe shutdown system performance includes examination of systems in standby status as well as those in operation. When a safe shutdown system is in operation, its performance can be directly observed and compared to its design capability (e.g., rated flow is required but cannot be achieved). When an operating safe shutdown system cannot fulfill its design function, its performance is degraded. When a safe shutdown system is in standby, its performance capability may not be readily determined. One or more of the following can provide indirect indication of its performance capability:

" Electrical faults on power supplies

" Normally closed breakers in tripped position

" System annunciators activated

" Busy lights lit

" Insufficient system pressure from water leg pumps

" Elevated area temperatures or radiation levels Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft I Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 124 of 274

• Increased sump pump operation in areas in which the system is located CGS Basis Reference(s):

1. FSAR Section 3.7.4 Seismic Instrumentation

2. ISP-SEIS-M201 Seismic Systems Channel Check

3. PPM 4.851.S1.5-1 Operating Basis Earthquake Exceeded

4. ABN-EARTHQUAKE Earthquake

5. FSAR Section 2.1.1 Specification of Location Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 125 of 274 Category: H - Hazards & Other Conditions Affecting Plant Safety Subcategory: 1 - Natural & Destructive Phenomena Initiating Condition: Natural or destructive phenomena affecting vital areas EAL:

HA1.2 Alert Tornado striking or sustained high winds GT 70 mph resulting in EITHER:

VISIBLE DAMAGE to any Table H-1 plant structure containing safe shutdown systems or components OR Control Room indication of degraded performance of those safe shutdown systems Table H-1 Plant Structures Containing Safe Shutdown Systems or Components

• Reactor Building

" Turbine Building

• Radwaste/Control Building

• Standby Service Water Pump Houses

" Diesel Generator Building

• Diesel Fuel Oil Storage Area Mode Applicability:

All NEI 99-01 Basis:

These EALs escalate from HU1.2 in that the occurrence of the event has resulted in visible damage to plant structures or areas containing equipment necessary for a safe shutdown, or has caused damage to the safety systems in those structures evidenced by control room indications of degraded system response or performance. The occurrence of visible damage and/or degraded system response is intended to discriminate against lesser events. The initial report should not be interpreted as mandating a lengthy damage assessment prior to classification. No attempt is made in this EAL to assess the actual magnitude of the damage. The significance here is not that a particular system or structure was damaged, but rather, that the event was of sufficient magnitude to cause this degradation.

Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 126 of 274 Escalation of this emergency classification level, if appropriate, would be based on System Malfunction EALs.

This EAL is based on a tornado striking (touching down) or high winds that have caused visible damage to structures containing functions or systems required for safe shutdown of the plant.

CGS Basis:

Columbia Generating Station is designed to be safely shutdown following extreme high winds or a tornado. However, some of the systems and structures are not analyzed to function after extreme conditions (SW spray trees, CW cooling towers). Therefore, enhanced plant monitoring is warranted during severe wind conditions.

Wind advisories are issued by the National Warning Alert System or Hanford Meteorology. The on-site 15 minute average wind speed at 33' above grade elevation is indicated at MET-WSR-4 (Red Pen) on H13-P823 (Board L) (Ref. 3).

The design basis wind velocity for all Category 1 structures at CGS is 100 mph based on a 10 second average (Ref. 1). However, the Control Room wind speed recorder was modified to provide 15 min.

averages. The EAL threshold reflects a de-rated design wind speed of 70 mph to compensate for gust factor based on the extended wind speed indicator averaging (Ref. 2).

A tornado striking (touching down) within the Protected Area resulting in visible damage warrants declaration of an Alert regardless of the measured wind speed at the meteorological tower. A tornado is defined as a violently rotating column of air in contact with the ground and extending from the base of a thunderstorm.

Wind speed, wind direction and temperature readings are provided instantaneously and in 15 minute averages. The term "sustained" is intended to mean the 15 minute average wind speed indicator.

The Protected Area boundary is defined in Civil Drawing C897 Columbia Generating Station Site Plan (Ref. 3).

Visible Damage is defined as:

Damage to equipment or structure that is readily observable without measurements, testing, or analysis. Damage is sufficient to cause concern regarding the continued operability or reliability of affected safety structure, system, or component. Example damage includes: deformation due to heat or impact, denting, penetration, rupture, cracking, paint blistering. Surface blemishes (e.g.,

paint chipping, scratches) should not be included.

The Table H-1 Plant Structures Containing Safe Shutdown Systems or Components include those structures that contain any Class 1, 2 or 3 SSC (Ref. 5). The Alert classification is appropriate if relevant plant parameters indicate that the performance of safety systems in the affected vital areas has been degraded. No attempt should be made to fully inventory the actual magnitude of the damage or quantify the degradation of safety system performance prior to declaration of an Alert under this threshold. The declaration of an Alert and the activation of the TSC provide the Emergency Director with the resources needed to perform detailed damage assessments.

A time interval is not associated with this EAL and degraded safe shutdown system performance may not be immediately apparent. Assessment of degraded safe shutdown system performance includes examination of systems in standby status as well as those in operation. When a safe shutdown Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 127 of 274 system is in operation, its performance can be directly observed and compared to its design capability (e.g., rated flow is required but cannot be achieved). When an operating safe shutdown system cannot fulfill its design function, its performance is degraded. When a safe shutdown system is in standby, its performance capability may not be readily determined. One or more of the following can provide indirect indication of its performance capability:

• Electrical faults on power supplies

• Normally closed breakers in tripped position

• System annunciators activated

• BISI lights lit

• Insufficient system pressure from water leg pumps

• Elevated area temperatures or radiation levels

• Increased sump pump operation in areas in which the system is located CGS Basis Reference(s):

1. FSAR Section 3.1.1 Wind Loadings

2. CGS Calculation CALC CE-02-93-16 Evaluate PMR/BDC 98-0131 -OA change from 5 min. to 15 min. averaging of 33 ft. elev. met twr. wind speeds for UE and Alert declarations

3. ABN-WIND Tornado/High Wind

4. Civil Drawing C897 Columbia Generating Station Site Plan

5. FSAR Table 3.2-1 Equipment Classification Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 128 of 274 Category: H - Hazards & Other Conditions Affecting Plant Safety Subcategory: 1 - Natural & Destructive Phenomena Initiating Condition: Natural or destructive phenomena affecting vital areas EAL:

HA1.3 Alert Turbine failure-generated projectiles resulting in EITHER:

VISIBLE DAMAGE to or penetration of any Table H-1 plant structure containing safe shutdown systems or components OR Control Room indication of degraded performance of those safe shutdown systems Table H-1 Plant Structures Containing Safe Shutdown Systems or Components

• Reactor Building

• Turbine Building

• Radwaste/Control Building

• Standby Service Water Pump Houses

• Diesel Generator Building

• Diesel Fuel Oil Storage Area Mode Applicability:

All NEI 99-01 Basis:

This EALs escalate from HU1.3 in that the occurrence of the event has resulted in visible damage to plant structures or areas containing equipment necessary for a safe shutdown, or has caused damage to the safety systems in those structures evidenced by control room indications of degraded system response or performance. The occurrence of visible damage and/or degraded system response is intended to discriminate against lesser events. The initial report should not be interpreted as mandating a lengthy damage assessment prior to classification. No attempt is made in this EAL to assess the actual magnitude of the damage. The significance here is not that a particular system or structure was damaged, but rather, that the event was of sufficient magnitude to cause this degradation.

Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 129 of 274 Escalation of this emergency classification level, if appropriate, would be based on System Malfunction EALs.

This EAL addresses the threat to safety related equipment imposed by projectiles generated by main turbine rotating component failures. Therefore, this EAL is consistent with the definition of an Alert in that the potential exists for actual or substantial potential degradation of the level of safety of the plant.

CGS Basis:

The turbine generator stores large amounts of rotational kinetic energy in its rotor. In the unlikely event of a major mechanical failure, this energy may be transformed into both rotational and translational energy of rotor fragments. These fragments may impact the surrounding stationary parts. If the energy-absorbing capability of these stationary turbine generator parts is insufficient, external missiles will be released. These ejected missiles may impact various plant structures, including those housing safety related equipment.

In the event of missile ejection, the probability of a strike on a plant region is a function of the energy and-direction of an ejected missile and of the orientation of the turbine with respect to the plant region.

Per the CGS FSAR it is concluded that the probability of damage to safety-related systems by turbine missiles is extremely low, due to (a) the protection provided by reinforced-concrete structural barriers, (b) the calculated probability of turbine missile generation, and (c) periodic testing and inspection of turbine overspeed protection systems with associated corrective action as required (Ref. 1).

Visible Damage is defined as:

Damage to equipment or structure that is readily observable without measurements, testing, or analysis. Damage is sufficient to cause concern regarding the continued operability or reliability of affected safety structure, system, or component. Example damage includes: deformation due to heat or impact, denting, penetration, rupture, cracking, paint blistering. Surface blemishes (e.g.,

paint chipping, scratches) should not be included.

A time interval is not associated with this EAL and degraded safe shutdown system performance may not be immediately apparent. Assessment of degraded safe shutdown system performance includes examination of systems in standby status as well as those in operation. When a safe shutdown system is in operation, its performance can be directly observed and compared to its design capability (e.g., rated flow is required but cannot be achieved). When an operating safe shutdown system cannot fulfill its design function, its performance is degraded. When a safe shutdown system is in standby, its performance capability may not be readily determined. One or more of the following can provide indirect indication of its performance capability:

• Electrical faults on power supplies

" Normally closed breakers in tripped position

" System annunciators activated

" Busy lights lit

• Insufficient system pressure from water leg pumps

• Elevated area temperatures or radiation levels

• Increased sump pump operation in areas in which the system is located Attachment 7.1, EAL Bases

Number: 13.1.1A I Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 130 of 274 CGS Basis Reference(s):

1. FSAR Section 3.5.1.3 Turbine Missiles Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 131 of 274 Category: H - Hazards & Other Conditions Affecting Plant Safety Subcategory: 1 - Natural & Destructive Phenomena Initiating Condition: Natural or destructive phenomena affecting vital areas EAL:

HA1.4 Alert Internal flooding in any Table H-1 plant structure resulting in EITHER:

An electrical shock hazard that precludes access to operate or monitor safe shutdown systems or components OR Control Room indication of degraded performance of those safe shutdown systems Table H-1 Plant Structures Containing Safe Shutdown Systems or Components

• Reactor Building

• Turbine Building

• Radwaste/Control Building

• Standby Service Water Pump Houses

• Diesel Generator Building

• Diesel Fuel Oil Storage Area Mode Applicability:

All NEI 99-01 Basis:

This EAL escalates from HU1.4 in that the occurrence of the event has resulted in an electrical shock hazard precluding access to plant structures containing safe shutdown systems or components or damage to the safety systems or components in those structures as evidenced by control room indications of degraded system response or performance. The lack of access or occurrence of degraded system response is intended to discriminate against lesser events. The initial report should not be interpreted as mandating a lengthy assessment prior to classification. No attempt is made in this EAL to assess the actual magnitude of any damage. The significance here is not that a particular system or structure was damaged, but rather, that the event was of sufficient magnitude to cause this lack of access or performance degradation.

Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 132 of 274 Escalation of this emergency classification level, if appropriate, would be based on System Malfunction EALs.

This EAL addresses the effect of internal flooding caused by events such as component failures, equipment misalignment, or outage activity mishaps. It is based on the degraded performance of systems, or has created industrial safety hazards (e.g., electrical shock) that preclude necessary access to operate or monitor safety equipment. The inability to access, operate or monitor safety equipment represents an actual or substantial potential degradation of the level of safety of the plant.

Flooding as used in this EAL describes a condition where water is entering the room faster than installed equipment is capable of removal, resulting in a rise of water level within the room.

Classification of this EAL should not be delayed while corrective actions are being taken to isolate the water source.

CGS Basis:

The Table H-1 Plant Structures Containing Safe Shutdown Systems or Components include those structures that contain any Class 1, 2 or 3 SSC (Ref. 1).

A time interval is not associated with this EAL and degraded safe shutdown system performance may not be immediately apparent. Assessment of degraded safe shutdown system performance includes examination of systems in standby status as well as those in operation. When a safe shutdown system is in operation, its performance can be directly observed and compared to its design capability (e.g., rated flow is required but cannot be achieved). When an operating safe shutdown system cannot fulfill its design function, its performance is degraded. When a safe shutdown system is in standby, its performance capability may not be readily determined. One or more of the following can provide indirect indication of its performance capability:

" Electrical faults on power supplies

• Normally closed breakers in tripped position

• System annunciators activated

• Busy lights lit

• Insufficient system pressure from water leg pumps

" Elevated area temperatures or radiation levels

• Increased sump pump operation in areas in which the system is located CGS Basis Reference(s):

1. FSAR Table 3.2-1 Equipment Classification Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 133 of 274 Category: H - Hazards & Other Conditions Affecting Plant Safety Subcategory: 1 - Natural & Destructive Phenomena Initiating Condition: Natural or destructive phenomena affecting vital areas EAL:

HA1.5 Alert Vehicle crash resulting in EITHER:

VISIBLE DAMAGE to any Table H-1 plant structure containing safe shutdown systems or components OR Control Room indication of degraded performance of those safe shutdown systems Table H-1 Plant Structures Containing Safe Shutdown Systems or Components

" Reactor Building

• Turbine Building

" Radwaste/Control Building

" Standby Service Water Pump Houses

• Diesel Generator Building

• Diesel Fuel Oil Storage Area Mode Applicability:

All NEI 99-01 Basis:

The occurrence of visible damage and/or degraded system response is intended to discriminate against lesser events. The initial report should not be interpreted as mandating a lengthy damage assessment prior to classification. No attempt is made in this EAL to assess the actual magnitude of the damage. The significance here is not that a particular system or structure was damaged, but rather, that the event was of sufficient magnitude to cause this degradation.

Escalation of this emergency classification level, if appropriate, would be based on System Malfunction EALs.

Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 134 of 274 This EAL addresses vehicle crashes within the Protected Area that results in visible damage to vital areas or indication of damage to safety structures, systems, or components containing functions and systems required for safe shutdown of the plant.

CGS Basis:

If the vehicle crash is determined to be hostile in nature, the event is classified under EAL HS4.1.

Visible Damage is defined as:

Damage to equipment or structure that is readily observable without measurements, testing, or analysis. Damage is sufficient to cause concern regarding the continued operability or reliability of affected safety structure, system, or component. Example damage includes: deformation due to heat or impact, denting, penetration, rupture, cracking, paint blistering. Surface blemishes (e.g.,

paint chipping, scratches) should not be included.

The Table H-1 Plant Structures Containing Safe Shutdown Systems or Components include those structures that contain any Class 1, 2 or 3 SSC (Ref. 1).

A time interval is not associated with this EAL and degraded safe shutdown system performance may not be immediately apparent. Assessment of degraded safe shutdown system performance includes examination of systems in standby status as well as those in operation. When a safe shutdown system is in operation, its performance can be directly observed and compared to its design capability (e.g., rated flow is required but cannot be achieved). When an operating safe shutdown system cannot fulfill its design function, its performance is degraded. When a safe shutdown system is in standby, its performance capability may not be readily determined. One or more of the following can provide indirect indication of its performance capability:

" Electrical faults on power supplies

• Normally closed breakers in tripped position

• System annunciators activated

• Busy lights lit

• Insufficient system pressure from water leg pumps

• Elevated area temperatures or radiation levels

• Increased sump pump operation in areas in which the system is located CGS Basis Reference(s):

1. FSAR Table 3.2-1 Equipment Classification Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 135 of 274 Category: H - Hazards & Other Conditions Affecting Plant Safety Subcategory: 2 - Fire or Explosion Initiating Condition: Fire within the Protected Area not extinguished within 15 minutes of detection or explosion within the Protected Area EAL:

HU2.1 Unusual Event FIRE not extinguished within 15 min. (Note 3) of Control Room notification or verification of a Control Room fire alarm in any Table H-1 plant structure NOTE 3: The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the condition will likely exceed the applicable time.

Table H-i Plant Structures Containing Safe Shutdown Systems or Components

• Reactor Building

• Turbine Building

" Radwaste/Control Building

• Standby Service Water Pump Houses

" Diesel Generator Building

" Diesel Fuel Oil Storage Area Mode Applicability:

All NEI 99-01 Basis:

This EAL addresses the magnitude and extent of fires that may be potentially significant precursors of damage to safety systems. It addresses the fire, and not the degradation in performance of affected systems that may result.

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

The 15 minute time period begins with a credible notification that a fire is occurring, or indication of a fire detection system alarm/actuation. Verification of a fire detection system alarm/actuation includes actions that can be taken within the control room or other nearby site specific location to ensure that it Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 136 of 274 is not spurious. An alarm is assumed to be an indication of a fire unless it is disproved within the 15 minute period by personnel dispatched to the scene. In other words, a personnel report from the scene may be used to disprove a sensor alarm if received within 15 minutes of the alarm, but shall not be required to verify the alarm.

The intent of this 15 minute duration is to size the fire and to discriminate against small fires that are readily extinguished (e.g., smoldering waste paper basket).

Escalation of this emergency classification level, if appropriate, would be based on HA2.1.

CGS Basis:

The 15 minute requirement begins with a credible notification that a fire is occurring, or receipt of a 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 15 minute time limit or a classification must be made. If a fire is verified to be occurring by field report, the 15 minute time limit is from the original receipt of the fire detection alarm.

The Table H-1 Plant Structures Containing Safe Shutdown Systems or Components include those structures that contain any Class 1, 2 or 3 SSC (Ref. 1).

CGS Basis Reference(s):

1. FSAR Table 3.2-1 Equipment Classification Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 137 of 274 Category: H - Hazards & Other Conditions Affecting Plant Safety Subcategory: 2 - Fire or Explosion Initiating Condition: Fire within the Protected Area not extinguished within 15 minutes of detection or explosion within the Protected Area EAL:

HU2.2 Unusual Event EXPLOSION within the PROTECTED AREA Mode Applicability:

All NEI 99-01 Basis:

This EAL addresses the magnitude and extent of explosions that may be potentially significant precursors of damage to safety systems. It addresses the explosion, and not the degradation in performance of affected systems that may result.

This EAL addresses only those explosions of sufficient force to damage permanent structures or equipment within the Protected Area.

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

The Emergency director also needs to consider any security aspects of the explosion, if applicable.

Escalation of this emergency classification level, if appropriate, would be based on HA2.1.

CGS Basis:

If damage is confirmed visually or by other in-plant indications, the event may be escalated to an Alert under EAL HA2.1.

The Protected Area boundary is defined in Civil Drawing C897 Columbia Generating Station Site Plan (Ref. 1).

As used here, an explosion is a rapid, violent, unconfined combustion or catastrophic failure of pressurized/energized equipment that imparts energy of sufficient force to potentially damage permanent structures, systems, or components. The occurrence of the explosion with reports of evidence of damage (e.g., deformation, scorching, etc.) is sufficient for declaration.

If the explosion is determined to be hostile in nature, the event is classified under security based EALs.

CGS Basis Reference(s):

1. Civil Drawing C897 Columbia Generating Station Site Plan Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 138 of 274 Category: H - Hazards & Other Conditions Affecting Plant Safety Subcategory: 2 - Fire or Explosion Initiating Condition: Fire or explosion affecting the operability of plant safety systems required to establish or maintain safe shutdown EAL:

HA2.1 Alert FIRE or EXPLOSION resulting in EITHER:

VISIBLE DAMAGE to any Table H-1 plant structure containing safe shutdown systems or components OR Control Room indication of degraded performance of those safe shutdown systems Table H-1 Plant Structures Containing Safe Shutdown Systems or Components

• Reactor Building

• Turbine Building

" Radwaste/Control Building

• Standby Service Water Pump Houses

• Diesel Generator Building

• Diesel Fuel Oil Storage Area Mode Applicability:

All NEI 99-01 Basis:

Visible damage is used to identify the magnitude of the fire or explosion and to discriminate against minor fires and explosions.

The reference to structures containing safety systems or components is included to discriminate against fires or explosions in areas having a low probability of affecting safe operation. The significance here is not that a safety system was degraded but the fact that the fire or explosion was large enough to cause damage to these systems.

Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 139 of 274 The use of visible damage should not be interpreted as mandating a lengthy damage assessment prior to classification. The declaration of an Alert and the activation of the Technical Support Center will provide the Emergency Director with the resources needed to perform detailed damage assessments.

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

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

CGS Basis:

Fire, as used in this EAL, means 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.

An explosion is a rapid, violent, unconfined combustion or a catastrophic failure of pressurized equipment that potentially imparts significant energy to nearby structures and materials.

Visible Damage is defined as:

Damage to equipment or structure that is readily observable without measurements, testing, or analysis. Damage is sufficient to cause concern regarding the continued operability or reliability of affected safety structure, system, or component. Example damage includes: deformation due to heat or impact, denting, penetration, rupture, cracking, paint blistering. Surface blemishes (e.g.,

paint chipping, scratches) should not be included.

The Table H-1 Plant Structures Containing Safe Shutdown Systems or Components include those structures that contain any Class 1, 2 or 3 SSC (Ref. 1).

A time interval is not associated with this EAL and degraded safe shutdown system performance may not be immediately apparent. Assessment of degraded safe shutdown system performance includes examination of systems in standby status as well as those in operation. When a safe shutdown system is in operation, its performance can be directly observed and compared to its design capability (e.g., rated flow is required but cannot be achieved). When an operating safe shutdown system cannot fulfill its design function, its performance is degraded. When a safe shutdown system is in standby, its performance capability may not be readily determined. One or more of the following can provide indirect indication of its performance capability:

• Electrical faults on power supplies

• Normally closed breakers in tripped position

• System annunciators activated

• Busy lights lit

" Insufficient system pressure from water leg pumps

• Elevated area temperatures or radiation levels

• Increased sump pump operation in areas in which the system is located Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 140 of 274 CGS Basis Reference(s):

1. FSAR Table 3.2-1 Equipment Classification Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev:. Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 141 of 274 Category: H - Hazards & Other Conditions Affecting Plant Safety Subcategory: 3- Hazardous Gas Initiating Condition: Release of toxic, corrosive, asphyxiant or flammable gases deemed detrimental to normal plant operations EAL:

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

All NEI 99-01 Basis:

This EAL is based on the release of toxic, corrosive, asphyxiant or flammable gases of sufficient quantity to affect normal plant operations.

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

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

An asphyxiant is a gas capable of reducing the level of oxygen in the body to dangerous levels. Most commonly, asphyxiants work by merely displacing air in an enclosed environment. This reduces the concentration of oxygen below the normal level of around 19%, which can lead to breathing difficulties, unconsciousness or even death.

Escalation of this emergency classification level, if appropriate, would be based on HA3.1.

CGS Basis:

As used in this EAL, affecting normal plant operations means that activities at the plant site associated with routine testing, maintenance, or equipment operations, in accordance with normal operating or administrative procedures have been impacted. The EAL criterion is not met until an evaluation of the affect on normal plant operations has been completed. Entry into abnormal or emergency operating procedures, or deviation from normal security or radiological controls posture, is a departure from normal plant operations and thus would be considered to have been affected.

The release may have originated within the Site Boundary, or it may have originated offsite and subsequently drifted onto the Site Boundary. Offsite events (e.g., tanker truck accident releasing toxic gases, etc.) resulting in the plant being within the evacuation area should also be considered in this EAL because of the adverse affect on normal plant operations.

FSAR Section 6.4.4.2 (Ref. 1) provides a summary of the analysis performed for potential toxic chemical releases in and around Columbia Generating Station. The worst case releases are the Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 142 of 274 potential rupture of a chlorine tanker railroad car on the railroad tracks north of the plant and a sodium release at FFTF. Both of these events would result in adequate time for the Control Room HVAC system to be placed in the recirculation mode of operation. The quantities and types of other chemicals stored at Columbia Generating Station should not result in a threat to Control Room habitability should a spill occur (Ref. 2).

A toxic gas is considered to be any gas that is dangerous to life or limb by reason of inhalation or skin contact. A halon discharge in the Control Room is not considered a toxic gas in this situation. The only Halon system at CGS is in the Control Room PGCC troughs under most of the Control Room panels. There are penetration seals in the PGCC to contain the Halon gas. The removable floor panels could slightly leak but Halon is heavier than air and there should not be much leakage.

Should the release affect plant vital areas, escalation to an Alert would be based on EAL HA3.1.

Should an explosion or fire occur due to flammable gas within an affected plant area, an Alert may be appropriate based on EAL HA2.1.

CGS Basis Reference(s):

1. FSAR Section 6.4.4.2

2. ABN-HAZMAT Hazardous Material Spills/Releases Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 143 of 274 Category: H - Hazards & Other Conditions Affecting Plant Safety Subcategory: 3 - Hazardous Gas Initiating Condition: Release of toxic, corrosive, asphyxiant or flammable gases deemed detrimental to normal plant operations EAL:

HU3.2 Unusual Event Recommendation by local, county, state or DOE officials to evacuate or shelter site personnel based on offsite event Mode Applicability:

All NEI 99-01 Basis:

This EAL is based on the release of toxic, corrosive, asphyxiant or flammable gases of sufficient quantity to affect normal plant operations.

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

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

An asphyxiant is a gas capable of reducing the level of oxygen in the body to dangerous levels. Most commonly, asphyxiants work by merely displacing air in an enclosed environment. This reduces the concentration of oxygen below the normal level of around 19%, which can lead to breathing difficulties, unconsciousness or even death.

Escalation of this emergency classification level, if appropriate, would be based on HA3.1.

CGS Basis:

This EAL is based on the existence of an uncontrolled release originating off site and local, county, state or DOE officials have reported the need for evacuation or sheltering of site personnel. Offsite events (e.g., tanker truck accident releasing toxic gases, etc.) are considered in this EAL because they may adversely affect normal plant operations. The EAL criterion is not met until an evaluation of the affect on normal plant operations has been completed.

DOE officials may determine the evacuation area for offsite spills by using the Department of Transportation (DOT) Evacuation Tables for Selected Hazardous Materials in the DOT Emergency Response Guide for Hazardous Materials.

FSAR Section 6.4.4.2 (1) provides a summary of the analysis performed for potential toxic chemical releases in and around Columbia Generating Station. The worst case releases are the potential rupture of a chlorine tanker railroad car on the railroad tracks north of the plant and a sodium release at FFTF. Both of these events would result in adequate time for the Control Room HVAC system to be placed in the recirculation mode of operation. The quantities and types of other chemicals stored at Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 144 of 274 Columbia Generating Station should not result in a threat to Control Room habitability should a spill occur ( 2).

Should the release affect plant vital areas, escalation to an Alert would be based on EAL HA3.1 or HA3.2. Should an explosion or fire occur due to flammable gas within an affected plant area, an Alert may be appropriate based on EAL HA2.1.

CGS Basis Reference(s):

1. FSAR Section 6.4.4.2

2. ABN-HAZMAT Hazardous Material Spills/Releases Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 145 of 274 Category: H - Hazards & Other Conditions Affecting Plant Safety Subcategory: 3 - Hazardous Gas Initiating Condition: Access to a vital area is prohibited due to toxic, corrosive, asphyxiant or flammable gases which jeopardize operation of operable equipment required to maintain safe operations or safely shutdown the reactor EAL:

HA3.1 Alert Access to any Table H-1 plant structure is prohibited due to toxic, corrosive, asphyxiant, or flammable gases which jeopardize operation of safe shutdown systems (Note 5)

NOTE 5: If the equipment in the stated area was already inoperable, or out of service, before the event occurred, then this EAL should not be declared as.:it will have no adverse impact on the ability of the plant to safely operate or safely shutdown beyond that already allowed by Technical Specifications at the time of the event.

Table H-1 Plant Structures Containing Safe Shutdown Systems or Components

• Reactor Building

• Turbine Building

• Radwaste/Control Building

• Standby Service Water Pump Houses

" Diesel Generator Building 0 Diesel Fuel Oil Storage Area Mode Applicability:

All NEI 99-01 Basis:

Gases in a vital area can affect the ability to safely operate or safely shutdown the reactor.

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

Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 146 of 274 Declaration should not be delayed for confirmation from atmospheric testing if the atmosphere poses an immediate threat to life and health or an immediate threat of severe exposure to gases. This could be based upon documented analysis, indication of personal ill effects from exposure, or operating experience with the hazards.

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

An asphyxiant is a gas capable of reducing the level of oxygen in the body to dangerous levels. Most commonly, asphyxiants work by merely displacing air in an enclosed environment. This reduces the concentration of oxygen below the normal level of around 19%, which can lead to breathing difficulties, unconsciousness or even death.

An uncontrolled release of flammable gasses within a facility structure has the potential to affect safe operation of the plant by limiting either operator or equipment operations due to the potential for ignition and resulting equipment damage/personnel injury. Flammable gasses, such as hydrogen and acetylene, are routinely used to maintain plant systems (hydrogen) or to repair equipment/components (acetylene - used in welding). This EAL assumes concentrations of flammable gasses which can ignite/support combustion.

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

CGS Basis:

This EAL is based on gases that have entered a plant structure in concentrations that could be unsafe for plant personnel and, therefore, preclude access to equipment necessary for the safe operation of the plant. Table H-1 safe shutdown structures contain systems that are operated to establish or maintain safe shutdown. The Table H-1 Plant Structures Containing Safe Shutdown Systems or Components include those structures that contain any Class 1, 2 or 3 SSC (1).

Flammable gasses, such as hydrogen and acetylene, are routinely used to maintain plant systems (hydrogen) or to repair equipment/components (acetylene - used in welding). This EAL addresses concentrations at which gases can ignite/support combustion. An uncontrolled release of flammable gasses within a facility structure has the potential to affect safe operation of the plant by limiting either operator or equipment operations due to the potential for ignition and resulting equipment damage/personnel injury.

The only halon system at CGS is located in the Control Room PGCC troughs under most of the Control Room panels. Penetration seals in the PGCC are designed to contain any discharged halon gas. The removable floor panels may be susceptible to minor leakage but halon is heavier than air and should not rise appreciably above the floor level. In response to various halon-related operating experience reports, Fire Protection Engineering claims an inadvertent halon discharge will not appreciably affect Control Room operations because of this halon system design. Halon discharge in the Control Room, therefore, should not be immediately hazardous to the Control Room staff.

This EAL does not apply to routine inerting of the Primary Containment.

Attachment 7.1, EAL Bases

Number: 13.1.1A IUse Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 147 of 274 CGS Basis Reference(s):

1. FSAR Table 3.2-1 Equipment Classification Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 148 of 274 Category: H - Hazards & Other Conditions Affecting Plant Safety Subcategory: 4 - Security Initiating Condition: Confirmed security condition or threat which indicates a potential degradation in the level of safety of the plant EAL:

HU4.1 Unusual Event A SECURITY CONDITION that does not involve a HOSTILE ACTION as reported by the Security Sergeant or Security Lieutenant OR A credible site-specific security threat notification OR A validated notification from NRC providing information of a site-specific aircraft threat Mode Applicability:

All NEI 99-01 Basis:

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

Security events which do not represent a potential degradation in the level of safety of the plant are reported under 10 CFR 73.71 or in some cases under 10 CFR 50.72. Security events assessed as hostile actions are classifiable under HA4.1, HS4.1 and HG1.1.

A higher initial classification could be made based upon the nature and timing of the security threat and potential consequences. The Emergency Director shall consider upgrading the emergency response status and emergency classification level in accordance with the CGS Physical Security Plan.

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

This threshold is based on the CGS Physical Security Plan. The CGS Physical Security Plan is based on guidance provided by NEI 03-12.

2nd Condition This threshold is included to ensure that appropriate notifications for the security threat are made in a timely manner. This includes information of a credible threat. Only the plant to which the specific threat is made need declare the Unusual Event.

Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft I - Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 149 of 274 The determination of "credible" is made through use of information found in the CGS Physical Security Plan.

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

This EAL is met when a plant receives information regarding an aircraft threat from NRC. Validation is performed by calling the NRC or by other approved methods of authentication. Only the plant to which the specific threat is made need declare the Unusual Event.

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

Escalation to Alert emergency classification level would be via HA4.1 would be appropriate if the threat involves an airliner within 30 minutes of the plant.

CGS Basis:

The Security Shift Supervision is defined as either the Security Lieutenant or the Security Sergeant (Ref. 1).

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.

Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 150 of 274 The following table lists the Security Events that meet the criteria of a Security Condition:

Security Security Event' Condition1 Hostile Action2 1 V V" 2 V V 3 V N/A 4 N/A N/A 5 V V 6 V V 7 V V 8 V V 9 N/A N/A 10 V V 11 V V 12 V V 13 V V 14 V V 15 V V 16 N/A N/A 17 1/ V 18 V V 19 V V 20 V N/A 1The Security Event titles are considered sensitive security information and are therefore not included in this document. For Security Event titles refer to the CGS Physical security Plan.

2 Each security event is categorized as a "Security Condition" until such time as the assessment determines that the actions are indeed validated and underway, at which time the categorization will be escalated to "Hostile Action." The escalation to "Hostile Action" may cause the Security Event being evaluated to gravitate to another Security Event where the "Hostile Action" occurring is better aligned.

Hostile Action: An act toward a nuclear power plant or its personnel that includes the use of violent force to destroy equipment, take hostages, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, projectiles, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. Hostile Action should not be construed to include acts of civil disobedience or felonious acts that are not part Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 151 of 274 of a concerted attack on the Nuclear Power Plant. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the Security Defined Owner Controlled Area).

This EAL is based on the CGS Physical Security Plan (Ref.1). As defined in the Physical Security Plan, "credible" is information received from a source determined to be reliable (such as Law Enforcement, government agency, etc.) or has been verified to be true.

Note that the ISFSI Protected Area is an area separate from the Protected Area surrounding the power block.

CGS Basis Reference(s):

1. CGS Physical Security Plan Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 152 of 274 Category: H - Hazards & Other Conditions Affecting Plant Safety Subcategory: 4 - Security Initiating Condition: Hostile action within the Security Defined Owner Controlled Area (SDOCA) or airborne attack threat EAL:

HA4.1 Alert A HOSTILE ACTION is occurring or has occurred within the SDOCA as reported by the Security Sergeant or Security Lieutenant OR A validated notification from NRC of a site-specific airliner attack threat LT 30 min. away Mode Apolicability:

All NEI 99-01 Basis:

Note: Timely and accurate communication between security shift supervision and the Control Room is crucial for the implementation of effective security EALs.

These EALs address the contingency for a very rapid progression of events, such as that experienced on September 11, 2001. They are not premised solely on the potential for a radiological release.

Rather the issue includes the need for rapid assistance due to the possibility for significant and indeterminate damage from additional air, land or water attack elements.

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

1st Condition This EAL addresses the potential for a very rapid progression of events due to a hostile action. It is not intended to address incidents that are accidental events or acts of civil disobedience, such as small aircraft impact, hunters, or physical disputes between employees within the Security Defined Owner Controlled Area (SDOCA). Those events are adequately addressed by other EALs.

Note that this EAL is applicable for any hostile action occurring, or that has occurred, in the SDOCA.

This includes ISFSI's that may be outside the Protected Area but still within the SDOCA.

Although nuclear plant security officers are well trained and prepared to protect against hostile action, it is appropriate for OROs to be notified and encouraged to begin activation (if they do not normally) to be better prepared should it be necessary to consider further actions.

Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 153 of 274 If not previously notified by the NRC that the airborne hostile action was intentional, then it would be expected, although not certain, that notification by an appropriate Federal agency would follow. In this case, appropriate federal agency is intended to be NORAD, FBI, FAA or NRC. However, the declaration should not be unduly delayed awaiting Federal notification.

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

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

This EAL is met when a plant receives information regarding an airliner attack threat from NRC and the airliner is within 30 minutes of the plant. Only the plant to which the specific threat is made need declare the Alert.

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

CGS Basis:

The Security Shift Supervision is defined as either the Security Lieutenant or the Security Sergeant (Ref.1).

Hostile Action: An act toward a nuclear power plant or its personnel that includes the use of violent force to destroy equipment, take hostages, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, projectiles, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. Hostile Action should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the nuclear power plant. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the Security Defined Owner Controlled Area).

Note that the ISFSI Protected Area is an area separate from the Protected Area surrounding the power block.

CGS Basis Reference(s):

1. CGS Physical Security Plan Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 154 of 274 Category: H - Hazards & Other Conditions Affecting Plant Safety Subcategory: 4 - Security Initiating Condition: Hostile action within the Protected Area EAL:

HS4.1 Site Area Emergency Security Sergeant or Security Lieutenant reports a HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA Mode Applicability:

All NEI 99-01 Basis:

This condition represents an escalated threat to plant safety above that contained in the Alert in that a hostile force has progressed from the Security Defined Owner Controlled Area (SDOCA) to the Protected Area.

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

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

This EAL addresses the potential for a very rapid progression of events due to a hostile action. It is not intended to address incidents that are accidental events or acts of civil disobedience, such as small aircraft impact, hunters, or physical disputes between employees within the Protected Area.

Those events are adequately addressed by other EALs.

Although nuclear plant security officers are well trained and prepared to protect against hostile action, it is appropriate for OROs to be notified and encouraged to begin preparations for public protective actions (if they do not normally) to be better prepared should it be necessary to consider further actions.

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

Escalation of this emergency classification level, if appropriate, would be based on actual plant status after impact or progression of attack.

Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 155 of 274 CGS Basis:

The Security Shift Supervision is defined as either the Security Lieutenant or the Security Sergeant.

These individuals 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 CGS Physical Security Plan (Safeguards) information. (Ref. 1)

Note that the ISFSI Protected Area is an area separate from the Protected Area surrounding the power block.

CGS Basis Reference(s):

1. CGS Physical Security Plan Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 156 of 274 Category: H - Hazards & Other Conditions Affecting Plant Safety Subcategory: 4 - Security Initiating Condition: Hostile action resulting in loss of physical control of the facility EAL:

HG4.1 General Emergency A HOSTILE ACTION has occurred such that plant personnel are unable to operate equipment required to maintain safety functions (i.e., reactivity control, RPV water level, or decay heat removal)

OR A hostile action has caused failure of Spent Fuel Cooling Systems and IMMINENT fuel damage is likely Mode Applicability:

All NEI 99-01 Basis:

1st Condition This EAL encompasses conditions under which a hostile action has resulted in a loss of physical control of vital areas (containing vital equipment or controls of vital equipment) required to maintain safety functions and control of that equipment cannot be transferred to and operated from another location.

Typically, these safety functions are reactivity control (ability to shut down the reactor and keep it shutdown) reactor water level (ability to cool the core), and decay heat removal (ability to maintain a heat sink).

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

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

2nd Condition This EAL addresses failure of spent fuel cooling systems as a result of hostile action if imminent fuel damage is likely, such as when a freshly off-loaded reactor core is in the spent fuel pool.

Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 157 of 274 CGS Basis:

Note that the ISFSI Protected Area is an area separate from the Protected Area surrounding the power block.

Failure of spent fuel cooling systems may be indicated by (1, 2):

• Reduction in fuel pool level

• Elevated fuel pool temperatures

• H1 3.P626.FPC1.1-2 (FUEL POOL TEMPERATURE HIGH) activated

• H13.P626.FPC1.2-2 (FUEL POOL LOW LEVEL HIGH/LOW) activated

• H1 3.P626.FPC1.1-1 (SKIMMER SURGE TANK-A LEVEL HIGH/HIGH) activated Fuel damage should not occur even if the pool boils provided the fuel remains submerged. Cladding damage occurs about two hours after a fuel assembly is uncovered. "Imminent" fuel damage exists, therefore, when fuel assembly submergence cannot be assured. (Ref. 1)

The FPC system is designed to limit the temperature of the Spent Fuel Pool to 125 0 F with the design heat load in the pool and both trains of FPC in operation. If one train of FPC is lost under these conditions and heat exchanger cooling water is 90°F or less, the peak fuel pool temperature will be 155 0 F or less. It will take at least 22 hours2.546296e-4 days <br />0.00611 hours <br />3.637566e-5 weeks <br />8.371e-6 months <br /> for the pool to heat up from 125 0 F to 155 0 F (FSAR fuel pool temperature limit). Following a full core off-load and a total loss of cooling, the Spent Fuel Pool will heat up at a rate of about 1 00 F/hr at 15 days after shutdown. The decay heat load of a full core off-load will exceed the FPC system heat removal capacity for about 6 months. For the spent fuel pool with the gate installed, the maximum expected level reduction rate is 1.4 in/min with an 1150 gpm leak. (Ref.1)

CGS Basis Reference(s):

1. ABN-FPC-LOSS Loss of Fuel Pool Cooling

2. PPM 4.626.FPC1 ANNUNCIATOR RESPONSE, 626.FPC1 Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 158 of 274 Category: H - Hazards & Other Conditions Affecting Plant Safety Subcategory: 5- Control Room Evacuation Initiating Condition: Control Room evacuation has been initiated EAL:

HA5.1 Alert Control Room evacuation has been initiated Mode Applicability:

All NEI 99-01 Basis:

With the control room evacuated, additional support, monitoring and direction through the Technical Support Center and/or other emergency response facilities may be necessary.

Inability to establish plant control from outside the control room will escalate this event to a Site Area Emergency.

CGS Basis:

The Shift Manager (SM) determines if the Control Room is inoperable 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 (Ref.1).

Inability to establish plant control from outside the Control Room (i.e., actions in ABN-CR-EVAC completed up to the requirement for emergency RPV depressurization from the Remote Shutdown Panel RSP) escalates this event to a Site Area Emergency per EAL HS5.1.

CGS Basis Reference(s):

1. ABN-CR-EVAC Control Room evacuation and Remote Cooldown Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 159 of 274 Category: H - Hazards & Other Conditions Affecting Plant Safety Subcategory: 5- Control Room Evacuation Initiating Condition: Control Room evacuation has been initiated and plant control cannot be established EAL:

HS5.1 Site Area Emergency Control Room evacuation has been initiated AND Control of the plant cannot be established within 15 min.

Mode Applicability:

All NEI 99-01 Basis:

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

The intent of the EAL is to establish control of important plant equipment and knowledge of important plant parameters in a timely manner. Primary emphasis should be placed on those components and instruments that supply protection for and information about safety functions. These safety functions are reactivity control (ability to shutdown the reactor and maintain it shutdown), reactor water level (ability to cool the core), and decay heat removal (ability to maintain a heat sink).

The determination of whether or not control is established at the remote shutdown panel is based on Emergency Director judgment. The Emergency Director is expected to make a reasonable, informed judgment within the allocated 15 min. that the licensee has control of the plant from the remote shutdown panel.

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

CGS Basis:

The Shift Manager determines if the Control Room is inoperable 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 (Ref. 1).

Control of the plant is established when actions in ABN-CR-EVAC have been completed up to the requirement for emergency RPV depressurization from the RSP (Ref. 1).

CGS Basis Reference(s):

1. ABN-CR-EVAC Control Room evacuation and Remote Cooldown Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 160 of 274 Category: H - Hazards & Other Conditions Affecting Plant Safety Subcategory: 6 - Judgment Initiating Condition: Other conditions exist which in the judgment of the Emergency Director warrant declaration of an Unusual Event EAL:

HU6.1 Unusual Event Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or have occurred which indicate EITHER:

A potential degradation of the level of safety of the plant OR A security threat to facility protection has been initiated.

No releases of radioactive material requiring off-site response or monitoring are expected unless further degradation of safety systems occurs Mode Applicability:

All NEI 99-01 Basis:

This EAL addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the Unusual Event emergency classification level.

CGS Basis:

The Emergency Director is the designated onsite individual having the responsibility and authority for implementing the CGS Emergency Plan. The Shift Manager (SM) initially acts in the capacity of the Emergency Director and takes actions as outlined in the Emergency Plan implementing procedures.

If required by the emergency classification or if deemed appropriate by the Emergency Director, 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 either the TSC Manager or ultimately the EOF Manager 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 (Ref. 1).

CGS Basis Reference(sl:

1. CGS Emergency Plan section 2.1 Organizational Concepts Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 161 of 274 Category: H - Hazards & Other Conditions Affecting Plant Safety Subcategory: 6 - Judgment Initiating Condition: Other conditions exist which in the judgment of the Emergency Director warrant declaration of an Alert EAL:

HA6.1 Alert Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or have occurred which involve EITHER:

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 NEI 99-01 Basis:

This EAL addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the Alert emergency classification level.

CGS Basis:

The Emergency Director is the designated onsite individual having the responsibility and authority for implementing the CGS Emergency Plan. The Shift Manager (SM) initially acts in the capacity of the Emergency Director and takes actions as outlined in the Emergency Plan implementing procedures.

If required by the emergency classification or if deemed appropriate by the Emergency Director, 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 either the TSC Manager or ultimately the EOF Manager 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 (Ref. 1).

CGS Basis Reference(s):

1. CGS Emergency Plan section 2.1 Organizational Concepts Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 162 of 274 Category: H - Hazards & Other Conditions Affecting Plant Safety Subcategory: 6 - Judgment Initiating Condition: Other conditions exist which in the judgment of the Emergency Director warrant declaration of a Site Area Emergency EAL:

HS6.1 Site Area Emergency Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or have occurred which involve EITHER:

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 NEI 99-01 Basis:

This EAL addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the emergency classification level description for Site Area Emergency.

CGS Basis:

The Emergency Director is the designated onsite individual having the responsibility and authority for implementing the CGS Emergency Plan. The Shift Manager (SM) initially acts in the capacity of the Emergency Director and takes actions as outlined in the Emergency Plan implementing procedures.

If required by the emergency classification or if deemed appropriate by the Emergency Director, 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 either the TSC Manager or ultimately the EOF Manager 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 (Ref. 1).

CGS Basis Reference(s):

1. CGS Emergency Plan section 2.1 Organizational Concepts Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 163 of 274 Category: H - Hazards & Other Conditions Affecting Plant Safety Subcategory: 6 - Judgment Initiating Condition: Other conditions exist which in the judgment of the Emergency Director warrant declaration of a General Emergency EAL:

HG6.1 General Emergency Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or have occurred which involve EITHER:

Actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity OR HOSTILE ACTION that results in an actual loss of physical control of the facility.

Releases can be reasonably expected to exceed EPA Protective Action Guideline exposure levels off-site for more than the immediate site area Mode Applicability:

All NEI 99-01 Basis:

This EAL addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the emergency classification level description for General Emergency.

CGS Basis:

The Emergency Director is the designated onsite individual having the responsibility and authority for implementing the CGS Emergency Plan. The Shift Manager (SM) initially acts in the capacity of the Emergency Director and takes actions as outlined in the Emergency Plan implementing procedures.

If required by the emergency classification or if deemed appropriate by the Emergency Director, 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 either the TSC Manager or ultimately the EOF Manager 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 (Ref. 1).

CGS Basis Reference(s):

1. CGS Emergency Plan Section 2.1 Organizational Concepts Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 164 of 274 Category: H - Hazards & Other Conditions Affecting Plant Safety Subcategory: 7 - 618-11 Area Initiating Condition: 618-11 Waste Burial Ground fire or explosion EAL:

HU7.1 Unusual Event Report of an explosion and/or fire involving or suspected to involve the waste buried within the 618-11 Waste Burial Ground Mode Applicability:

All NEI 99-01 Basis:

None.

CGS Basis:

This Initiating Condition and its associated EAL is based upon a report of a release from an abnormal event at the 618-11 Waste Burial Ground that could potentially be detrimental to the health and safety of CGS personnel and visitors within the CGS exclusion area.

OR Recommendation by the 618-11 Waste Burial Ground project officials for evacuation or shelter of CGS site personnel based on a 618-11 site event.

UE declaration ensures a heightened awareness to potentially hazardous events at 618-11. No Energy Northwest actions are prescribed. Washington Closure Hanford (DOE contractor) will request Energy Northwest personnel respond as prescribed in the 618-11 Emergency Plan by calling the Control Room (@2222) and providing the Shift Manager with response information. These are only recommendations, final Protective Action Decisions for Energy Northwest personnel and visitors are the responsibility of Energy Northwest.

(Ref. 1).

CGS Basis Reference(s):

1. Ammendment 218, Columbia Generating Station - Issuance Of Amendment Re: Change To Emergency Plan For U.S. Department Of Energy 618-11 Waste Burial Ground Remediation (Tac No.

Me3863)

2. CGS Emergency Plan Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 165 of 274 Category: H - Hazards & Other Conditions Affecting Plant Safety Subcategory: 7 - 618-11 Area Initiating Condition: 618-11 Waste Burial Ground release EAL:

HA7.1 Alert Report of a release resulting from an abnormal event at the 618-11 Waste Burial Ground that could potentially be detrimental to the health and safety of CGS personnel and visitors within the CGS exclusion area OR Recommendation by the 618-11 Waste Burial Ground project officials for evacuation or shelter of CGS site personnel based on a 618-11 event Mode Applicability:

All NEI 99-01 Basis:

None.

CGS Basis:

This Initiating Condition and its associated EAL is based upon a report of an explosion and/or fire involving or suspected to involve the waste buried within the 618-11 Waste Burial Ground.

An Alert activates the Energy Northwest ERO. This ensures there would be no delay in ERO activation is an unrelated emergency occurred at CGS which required ERO response. The ERO would activate and assemble on-site if possible, or off-site if directed, to be prepared to respond to an event at CGS. The ERO activation is NOT required to mitigate the fire or explosion at 618-11.

Washington Closure Hanford and other DOE contractors are solely responsible to mitigate accidents at 618-11 in accordance with DOE procedures.

(Ref. 1).

CGS Basis Reference(s):

1. Ammendment 218, Columbia Generating Station - Issuance Of Amendment Re: Change To Emergency Plan For U.S. Department Of Energy 618-11 Waste Burial Ground Remediation (Tac No.

Me3863)

2. CGS Emergency Plan Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 166 of 274 Category S - System Malfunction EAL Group: Hot Conditions (RCS temperature GT 2000 F); EALs in this cateqory are applicable only in one or more hot operating modes.

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

The events of this category pertain to the following subcategories:

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

2. Loss of DC Power Loss of plant vital DC 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 125 VDC power sources.

3. ATWS / Criticality Events related to failure of the Reactor Protection System (RPS) to initiate and complete reactor scrams. In the plant licensing basis, postulated failures of the RPS to complete a reactor scram comprise a specific set of analyzed events referred to as Anticipated Transient Without Scram (ATWS) events. For EAL classification however, ATWS is intended to mean any scram 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. Inadvertent criticalities pose potential personnel safety hazards as well being indicative of losses of reactivity control.

4. Inability to Reach or Maintain Shutdown Conditions System malfunctions may lead to loss of capability to remove heat removal the reactor core and RCS.

Two EALs fall into this subcategory. They are related to the failure of the plant to be brought to the required plant operating condition required by technical specifications if a limiting condition for operation (LCO) is not met and loss of heat removal capability.

Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 167 of 274

5. Instrumentation Certain events that degrade plant operator ability to effectively assess plant conditions within the plant warrant emergency classification. Loss of annunciators or indicators is in this subcategory.

6. Communications Certain events that degrade plant operator ability to effectively communicate with essential personnel within or external to the plant warrant emergency classification.

7. Fuel Clad Degradation 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.

8. RCS Leakaae The RPV provides a volume for the coolant that covers the reactor core. The RPV 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 are utilized to indicate potential pipe cracks that may propagate to an extent threatening fuel clad, RCS and Containment integrity.

Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 168 of 274 Category: S - System Malfunction Subcategory: 1 - Loss of AC Power Initiating Condition: Loss of all offsite AC power to emergency buses for 15 minutes or longer EAL:

SU1.1 Unusual Event Loss of all offsite AC power, Table S-3, to emergency buses SM-7 and SM-8 for GE 15 min.

(Note 3)

NOTE 3: The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the condition will likely exceed the applicable time.

Table S-3 AC Power Sources (Hot C6nditions)

Offsite

• Startup Transformer TR-S

• Backup Transformer TR-B Onsite

• Main Generator through Main Transformers to the Normal Transformers TR-N I/N2

" DG1

• DG2

" DG3 with SM-4 crosstie to SM-7 or SM-8 (EAL SGI.1 goj")

Mode Applicability:

1 - Power Operations, 2 - Startup, 3 - Hot Shutdown NEI 99-01 Basis:

Prolonged loss of off-site AC power reduces required redundancy and potentially degrades the level of safety of the plant by rendering the plant more vulnerable to a complete loss of AC power to emergency busses.

Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 169 of 274 Fifteen minutes was selected as a threshold to exclude transient or momentary losses of off-site power.

CGS Basis:

Table S-3 provides the list of AC power sources available to power emergency buses in hot conditions. (Ref. 1, 2)

Station Startup and shutdown 230KV power comes from the Ashe substation, located just north of the plant, through Startup transformer TR-S. The startup transformer usually supplies station auxiliary loads when the main generator is not available. Station Normal power, 25KV, is supplied from the main generator via Normal Auxiliary transformers TR-N1 and TR-N2. When the main generator is ready, all station auxiliary loads are transferred from the Startup transformer to the Normal Auxiliary transformers. Station Backup 115KV power from the Benton Substation feeder, located about 4 miles southeast of the plant, can be supplied to emergency buses SM-7 and SM-8. (Ref. 3, 4)

The buses addressed in this EAL are 4160V SM-7 and SM-8 (Ref. 5, 6).

• Bus SM-7 is normally energized from TR-N1 thru SM-1 when the main generator is operating and from TR-S thru SM-1 when the main generator is not operating. SM-7 can receive backup power from TR-B or emergency power from DG1.

• Bus SM-8 is normally energized from TR-N1 thru SM-3 when the main generator is operating and from TR-S thru SM-3 when the main generator is not operating. SM-8 can receive backup power from TR-B or emergency power from DG2.

Credit is not taken in this EAL for SM-4 crosstie capability because:

• Establishing the crosstie to SM-7 or SM-8 is assumed to require more than 15 minutes.

• Prolonged loss of offsite AC power reduces the required redundancy and potentially degrades the level of safety by rendering the plant more vulnerable to a complete loss of AC power (Station Blackout). (Ref. 7, 8)

• The HPCS pump does not have any long-term decay heat removal systems, in particular, wetwell cooling Each emergency bus has multiple sources of power. In order of preference, they are (Ref. 5, 6):

1. Normal Transformers: TR-N1 (SM-7 and/or SM-8)

2. Startup Transformers: TR-S (SM-7 and/or SM-8)

3. Backup Transformers: TR B (SM-7 and/or SM-8)

4. Emergency Diesel Generators: DG 1 (SM-7), DG 2 (SM-8)

Credit is not taken in this EAL for emergency diesel generators because they are onsite sources of power to SM-7 and SM-8.

It is possible to remove startup power from service and continue to supply the plant during shutdown conditions by backfeeding 500 KV power from Ashe Substation through the Main Transformers, the Normal Transformers and associated "N" breakers. This involves disconnecting the Main Generator from the Isolated Phase conductors (25 KV system) and overriding various interlocks.

Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 170 of 274 C GS Basis Reference(s):

1. FSAR Figure 8.1-2.1 Main One-Line Diagram - Main Buses

2. FSAR Figure 8.1-2.2 Main One-Line Diagram - Emergency Buses

3. FSAR Section 8.2

4. 01-53 Offsite Power

5. FSAR Section 8.3

6. ABN-ELEC-LOOP Loss Of All Off-Site Electrical Power

7. FSAR 8A.1

8. PPM 5.6.1 Station Blackout (SBO)

Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 171 of 274 Category: S - System Malfunction Subcategory: 1 - Loss of AC Power Initiating Condition: AC power capability to emergency buses reduced to a single power source for 15 minutes or longer such that any additional single failure would result in complete loss of AC power to emergency buses EAL:

SA1.1 Alert AC power capability, Table S-3, to emergency buses SM-7 and SM-8 reduced to a single power source for GE 15 min. (Note 3) such that any additional single power source failure would result in a loss of all emergency bus AC power NOTE 3: The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the condition will likely exceed the applicable time.

Table S-3: AC. Power $ources (Hot Condition.s) ,'. -...

Offsite

• Startup Transformer TR-S

" Backup Transformer TR-B Onsite

• Main Generator through Main Transformers to the Normal Transformers TR-N 1/N2

• DG1

" DG2

" DG3 with SM-4 crosstie to SM-7 or SM-8 (EAL SGI.1 ng "

Mode Applicability:

1 - Power Operations, 2 - Startup, 3 - Hot Shutdown Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 172 of 274 NEI 99-01 Basis:

The condition indicated by this EAL is the degradation of the off-site and on-site AC power systems such that any additional single failure would result in a complete loss of AC power to emergency buses. This condition could occur due to a loss of off-site power with a concurrent failure of all but one emergency generator to supply power to its emergency busses. Another related condition could be the loss of all off-site power and loss of on-site emergency generators with only one train of emergency busses being backfed from the unit main generator, or the loss of on-site emergency generators with only one train of emergency busses being backfed from off-site power. The subsequent loss of this single power source would escalate the event to a Site Area Emergency in accordance with SS1.1.

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

CGS Basis:

Table S-3 provides the list of AC power sources available to power emergency buses in hot conditions. (Ref. 1, 2).

Station Startup and shutdown 230KV power comes from the Ashe substation, located just north of the plant, through Startup transformer TR-S. The startup transformer usually supplies station auxiliary loads when the main generator is not available. Station Normal power, 25KV, is supplied from the main generator via Normal Auxiliary transformers TR-N1 and TR-N2. When the main generator is ready, all station auxiliary loads are transferred from the Startup transformer to the Normal Auxiliary transformers. Station Backup 115KV power from the Benton Substation feeder, located about 4 miles southeast of the plant, can be supplied to emergency buses SM-7 and SM-8. ( 3, 4)

The buses addressed in this EAL are 4160V SM-7 and SM-8 ( 5, 6).

• Bus SM-7 is normally energized from TR-N1 thru SM-1 when the main generator is operating and from TR-S thru SM-1 when the main generator is not operating. SM-7 can receive backup power from TR-B or emergency power from DG1.

• Bus SM-8 is normally energized from TR-N1 thru SM-3 when the main generator is operating and from TR-S thru SM-3 when the main generator is not operating. SM-8 can receive backup power from TR-B or emergency power from DG2.

Credit is not taken in this EAL for SM-4/DG3 crosstie capability because:

• Establishing the crosstie to SM-7 or SM-8 is assumed to require more than 15 minutes.

• Prolonged loss of offsite AC power reduces the required redundancy and potentially degrades the level of safety by rendering the plant more vulnerable to a complete loss of AC power (Station Blackout). (Ref. 7, 8)

• The HPCS pump does not have any long-term decay heat removal systems, in particular, wetwell cooling Power to buses SM-7 and SM-8 may come from either its respective standby diesel generator or from the Switch Yard through the Normal, Startup or Backup Transformers. Regardless of the source of power, failure of the remaining power source would result, at least temporarily, in a station blackout.

The determining factor of whether or not to classify then becomes the amount of time required to restore power to an emergency bus.

Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 173 of 274 Each emergency bus has multiple sources of power. In order of preference, they are ( 5, 6):

1. Normal Transformers: TR-N1 (SM-7 and/or SM-8)

2. Startup Transformers: TR-S (SM-7 and/or SM-8)

3. Backup Transformers: TR B (SM-7 and/or SM-8)

4. Emergency Diesel Generators: DG 1 (SM-7), DG 2 (SM-8)

It is possible to remove startup power from service and continue to supply the plant during shutdown conditions by backfeeding 500 KV power from Ashe Substation through the Main Transformers, the Normal Transformers and associated "N" breakers. This involves disconnecting the Main Generator from the Isolated Phase conductors (25 KV system) and overriding various interlocks.

This hot condition EAL is equivalent to the cold condition EAL CU1.1.

CGS Basis Reference(s):

1. FSAR Figure 8.1-2.1 Main One-Line Diagram - Main Buses

2. FSAR Figure 8.1-2.2 Main One-Line Diagram - Emergency Buses

3. FSAR Section 8.2

4. 01-53 Offsite Power

5. FSAR Section 8.3

6. ABN-ELEC-LOOP Loss Of All Off-Site Electrical Power

7. FSAR Section 8A.1

8. PPM 5.6.1 Station Blackout (SBO)

Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 174 of 274 Category: S - System Malfunction Subcategory: 1 - Loss of AC Power Initiating Condition: Loss of all offsite power and all onsite AC power to emergency buses for 15 minutes or longer EAL:

SS1.1 Site Area Emergency Loss of all offsite and all onsite AC power, Table S-3, to emergency buses SM-7 and SM-8 for GE 15 min. (Note 3)

NOTE 3: The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the condition will likely exceed the applicable time.

Tablel*"- AC Power Sources (HotConditions)

Offsite

• Startup Transformer TR-S

• Backup Transformer TR-B Onsite

• Main Generator through Main Transformers to the Normal Transformers TR-N 1/N2

• DG1

• DG2

• DG3 with SM-4 crosstie to SM-7 or SM-8 (EAL SGI.1 gniy)

Mode Applicability:

1 - Power Operations, 2 - Startup, 3 - Hot Shutdown Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 175 of 274 NEI 99-01 Basis:

Loss of all AC power to emergency busses compromises all plant safety systems requiring electric power including RHR, ECCS, Containment Heat Removal and the Ultimate Heat Sink. Prolonged loss of all AC power to emergency busses will lead to loss of Fuel Clad, RCS, and Containment, thus this event can escalate to a General Emergency.

Fifteen minutes was selected as a threshold to exclude transient or momentary losses of off-site power.

Escalation to General Emergency is via Fission Product Barrier Degradation or EAL SG1.1, "Prolonged Loss of All Off-site Power and Prolonged Loss of All On-site AC Power."

CGS Basis:

Table S-3 provides the list of AC power sources available to power emergency buses in hot conditions. (Ref. 1, 2)

Station Startup and shutdown 230KV power comes from the Ashe substation, located just north of the plant, through Startup transformer TR-S. The startup transformer usually supplies station auxiliary loads when the main generator is not available. Station Normal power, 25KV, is supplied from the main generator via Normal Auxiliary transformers TR-N1 and TR-N2. When the main generator is ready, all station auxiliary loads are transferred from the Startup transformer to the Normal Auxiliary transformers. Station Backup 115KV power from the Benton Substation feeder, located about 4 miles southeast of the plant, can be supplied to emergency buses SM-7 and SM-8. (Ref. 3, 4)

The buses addressed in this EAL are 4160V SM-7 and SM-8 (Ref. 5, 6).

" Bus SM-7 is normally energized from TR-N1 thru SM-1 when the main generator is operating and from TR-S thru SM-1 when the main generator is not operating. SM-7 can receive backup power from TR-B or emergency power from DGI.

" Bus SM-8 is normally energized from TR-N1 thru SM-3 when the main generator is operating and from TR-S thru SM-3 when the main generator is not operating. SM-8 can receive backup power from TR-B or emergency power from DG2.

Credit is not taken in this EAL for SM-4/DG3 crosstie capability because:

• Establishing the crosstie to SM-7 or SM-8 is assumed to require more than 15 minutes.

• Prolonged loss of offsite AC power reduces the required redundancy and potentially degrades the level of safety by rendering the plant more vulnerable to a complete loss of AC power (Station Blackout). (Ref. 7, 8)

• The HPCS pump does not have any long-term decay heat removal systems, in particular, wetwell cooling Power to buses SM-7 and SM-8 may come from either its respective standby diesel generator or from the Switch Yard through the Normal, Startup or Backup Transformers. Regardless of the source of power, failure of the remaining power source would result, at least temporarily, in a station blackout.

The determining factor of whether or not to classify then becomes the amount of time required to restore power to an emergency bus.

Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 176 of 274 Each emergency bus has multiple sources of power. In order of preference, they are (Ref. 5, 6):

1. Normal Transformers: TR-N1 (SM-7 and/or SM-8)

2. Startup Transformers: TR-S (SM-7 and/or SM-8)

3. Backup Transformers: TR B (SM-7 and/or SM-8)

4. Emergency Diesel Generators: DG 1 (SM-7), DG 2 (SM-8)

It is possible to remove startup power from service and continue to supply the plant during shutdown conditions by backfeeding 500 KV power from Ashe Substation through the Main Transformers, the Normal Transformers and associated "N" breakers. This involves disconnecting the Main Generator from the Isolated Phase conductors (25 KV system) and overriding various interlocks.

This hot condition EAL is equivalent to the cold condition EAL CA1.1.

CGS Basis Reference(s):

1. FSAR Figure 8.1-2.1 Main One-Line Diagram - Main Buses

2. FSAR Figure 8.1-2.2 Main One-Line Diagram - Emergency Buses

3. FSAR Section 8.2

4. 01-53 Offsite Power

5. FSAR Section 8.3

6. ABN-ELEC-LOOP Loss Of All Off-Site Electrical Power

7. FSAR Section 8A.1

8. PPM 5.6.1 Station Blackout (SBO)

Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft I - Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 177 of 274 Category: S -System Malfunction Subcategory: 1 - Loss of AC Power Initiating Condition: Prolonged loss of all offsite and all onsite AC power to emergency buses EAL:

SG1.1 General Emergency Loss of all offsite and all onsite AC power, Table S-3, to emergency buses SM-7 and SM-8 AND EITHER:

In the judgment of the Emergency Director, AC power to SM-7 or SM-8 is not likely to be restored within 4 hrs OR RPV level cannot be restored and maintained above -161 in. or cannot be determined Table S-,3 AC Power Sources" (HotConditions)

Offsite

• Startup Transformer TR-S

• Backup Transformer TR-B Onsite

• Main Generator through Main Transformers to the Normal Transformers TR-N 1/N2

• DG1

• DG2 I

DG3 with SM-4 crosstie to SM-7 or SM-8 (EAL SG1.1 only)

Mode Applicability:

1 - Power Operations, 2 - Startup, 3 - Hot Shutdown Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 178 of 274 NEI-9901 Basis:

Loss of all AC power to emergency busses compromises all plant safety systems requiring electric power including RHR, ECCS, Containment Heat Removal and the Ultimate Heat Sink. Prolonged loss of all AC power to emergency busses will lead to loss of fuel clad, RCS, and containment, thus warranting declaration of a General Emergency.

This EAL is specified to assure that in the unlikely event of a prolonged station blackout, timely recognition of the seriousness of the event occurs and that declaration of a General Emergency occurs as early as is appropriate, based on a reasonable assessment of the event trajectory.

The likelihood of restoring at least one emergency bus should be based on a realistic appraisal of the situation since a delay in an upgrade decision based on only a chance of mitigating the event could result in a loss of valuable time in preparing and implementing public protective actions.

In addition, under these conditions, fission product barrier monitoring capability may be degraded.

CGS Basis:

Table S-3 provides the list of AC power sources available to power emergency buses in hot conditions. (Ref. 1, 2)

Station Startup and shutdown 230KV power comes from the Ashe substation, located just north of the plant, through Startup transformer TR-S. The startup transformer usually supplies station auxiliary loads when the main generator is not available. Station Normal power, 25KV, is supplied from the main generator via Normal Auxiliary transformers TR-N1 and TR-N2. When the main generator is ready, all station auxiliary loads are transferred from the Startup transformer to the Normal Auxiliary transformers. Station Backup 115KV power from the Benton Substation feeder, located about 4 miles southeast of the plant, can be supplied to emergency buses SM-7 and SM-8. (Ref. 3, 4)

The buses addressed in this EAL are 4160V SM-7 and SM-8 (Ref. 5, 6).

• Bus SM-7 is normally energized from TR-N1 thru SM-1 when the main generator is operating and from TR-S thru SM-1 when the main generator is not operating. SM-7 can receive backup power from TR-B or emergency power from DG1.

• Bus SM-8 is normally energized from TR-N1 thru SM-3 when the main generator is operating and from TR-S thru SM-3 when the main generator is not operating. SM-8 can receive backup power from TR-B or emergency power from DG2.

Power to buses SM-7 and SM-8 may come from either its respective standby diesel generator or from the Switch Yard through the Normal, Startup or Backup Transformers. Regardless of the source of power, failure of the remaining power source would result, at least temporarily, in a station blackout.

The determining factor of whether or not to classify then becomes the amount of time required to restore power to an emergency bus.

Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Minor Rev: Draft Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 179 of 274 Each emergency bus has multiple sources of power. In order of preference, they are (Ref. 5, 6):

1. Normal Transformers: TR-N1 (SM-7 and/or SM-8)

2. Startup Transformers: TR-S (SM-7 and/or SM-8)

3. Backup Transformers: TR B (SM-7 and/or SM-8)

4. Emergency Diesel Generators: DG 1 (SM-7), DG 2 (SM-8)

Credit may be taken in this EAL for DG 3 crosstie capability provided a reasonable expectation exists that AC power can be restored to either SM-7 or SM-8 from DG3 and SM-4 within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

Four hours is the station blackout coping time (Ref 7, 8).

It is possible to remove startup power from service and continue to supply the plant during shutdown conditions by backfeeding 500 KV power from Ashe Substation through the Main Transformers, the Normal Transformers and associated "N" breakers. This involves disconnecting the Main Generator from the Isolated Phase conductors (25 KV system) and overriding various interlocks.

The second condition of this EAL should be answered by the Emergency Director by making a realistic assessment of the time required to complete any necessary repairs. This EAL requires the Emergency Director to classify the event as soon as his assessment indicates that necessary repairs will take longer than four hours rather than waiting for the time duration to expire.

Indication of continuing core cooling degradation must be based on fission product barrier monitoring with particular emphasis on Emergency Director judgment as it relates to imminent Loss or Potential Loss of fission product barriers and degraded ability to monitor fission product barriers. Indication of continuing core cooling degradation is manifested by an RPV level that cannot be restored and maintained above the top of active fuel, -161 in., or cannot be determined ( Ref. 9, 10). When RPV level is at or above the top of active fuel, the core is completely submerged. Core submergence is the most desirable means of core cooling. When RPV level is below the top of active fuel, the uncovered portion of the core must be cooled by less reliable means (i.e., steam cooling or spray cooling). If core uncovery is threatened, the EOPs specify alternate, more extreme, RPV level control measures in order to restore and maintain adequate core cooling. Since core uncovery begins if RPV level drops to -161 in., the level is indicative of a challenge to core cooling and the Fuel Clad barrier.

When RPV level cannot be determined, EOPs require RPV flooding strategies. RPV water level indication provides the primary means of knowing if adequate core cooling is being maintained. When all means of determining RPV level are unavailable, the fuel clad barrier is threatened and reliance on alternate means of assuring adequate core cooling must be attempted. The instructions in PPM 5.1.4 and PPM 5.1.6 specify these means, which include emergency depressurization of the RPV and injection into the RPV at a rate needed to flood to the elevation of the main steam lines or hold RPV pressure above the Minimum Steam Cooling Pressure (in ATWS events) (Ref. 11, 12). If RPV level cannot be determined with respect to the top of active fuel, a potential loss of the fuel clad barrier exists.

Consistent with the EOP definition of "cannot be restored and maintained," the determination that RPV level cannot be restored and maintained above the top of active fuel may be made at, before, or after RPV level actually decreases to this point. If RPV level dropped below the top of active fuel and subsequently restored above the top of active fuel before the Emergency Director determines that restoration of power to SM-7 or SM-8 is not likely, this EAL is not met.

Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 180 of 274 CGS Basis Reference(s):

1. FSAR Figure 8.1-2.1 Main One-Line Diagram - Main Buses

2. FSAR Figure 8.1-2.2 Main One-Line Diagram - Emergency Buses

3. FSAR Section 8.2

4. 01-53 Offsite Power

5. FSAR Section 8.3

6. ABN-ELEC-LOOP Loss Of All Off-Site Electrical Power

7. FSAR Section 8A.1

8. PPM 5.6.1 Station Blackout (SBO)

9. Calculation NE-02-03-05 Attachment 3 Note 8

10. PPM 5.1.1 RPV Control

11. PPM 5.1.4 RPV Flooding

12. PPM 5.1.6 RPV Flooding - ATWS Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 181 of 274 Category: S - System Malfunction Subcategory: 2 - Loss of DC Power Initiating Condition: Loss of all DC power for 15 minutes or longer EAL:

SS2.1 Site Area Emergency LT 108 VDC bus voltage indications on both 125 VDC buses DP-S1-1 and DP-S1-2 for GT 15 min.

(Note 3)

NOTE 3: The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the condition will likely exceed the applicable time.

Mode Applicability:

1 - Power Operations, 2 - Startup, 3 - Hot Shutdown NEI 99-01 Basis:

Loss of all DC power compromises ability to monitor and control plant safety functions. Prolonged loss of all DC power will cause core uncovering and loss of containment integrity when there is significant decay heat and sensible heat in the reactor system.

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

Escalation to a General Emergency would occur by Abnormal Rad Levels/Radiological Effluent, Fission Product Barrier Degradation.

CGS Basis:

The 125 VDC system is illustrated in Figure S-2 (Ref. 1, 2).

The 125 VDC Class 1 E DC power system consists of three electrically independent and separate distribution systems (S1-1, S1-2, and S1-HPCS). S1-HPCS is not included in this EAL. Each DC distribution system has a battery and a battery charger that are normally connected to the bus such that these two sources of power are operating in parallel. The charger is normally supplying system electrical loads with the battery on a float charge. Each battery has the necessary amp-hour discharge capacity to sustain system loads for a minimum of two hours. This capacity is specifically for a loss of power to the charger coincident with a design basis accident. The batteries have capacity to carry design load at 60°F without decreasing battery voltage below 1.81 volts/cell (or 108 VDC, Ref. 3) with loss of output from the battery chargers during the specified period. Battery capacity is sufficient to provide starting currents while operating at full load. (Ref. 4)

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

Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 182 of 274 CGS Basis Reference(s):

1. E505-1 DC One Line Diagram

2. E505-2 DC One Line Diagram

3. Calculation No. 2.05.01 Battery Sizing, Voltage Drop, and Charger Studies for Div. 1 & 2 Systems

4. FSAR Section 8.3.2 Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 183 of 274 Figure S-2:125 VDC System (Ref. 1, 2)

MC-S1I-1D-DP-Si-I A NSSS BOARD- DP-SI-2A LNSSS BOARD INVERTER IN-3A INVERTER IN-2A INVERTER IN-3B -

DP-SI-1 D INVERTER IN-2B REMOTE S/D PANEL - DP-SI-2D MC-S2-1ACONTPWR DIV. 2 CRIT SWGR

& REMOTE S/D PANEL DP-S1-1E -

DIG #1 DIST. PANEL DP-S1-2E DP-SI-IF DG #2 DIST. PANEL DIV. 1 CRIT SWGR ALT-REMOTE S/D PANEL MC-S2-1ACONTPWR Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 184 of 274 Category: S - System Malfunction Subcategory: 3 - ATWS / Criticality Initiating Condition: Inadvertent criticality EAL:

SU3.1 Unusual Event UNPLANNED sustained positive period observed on nuclear instrumentation Mode Applicability:

3 - Hot Shutdown NEI 99-01 Basis:

This EAL addresses inadvertent criticality events. This EAL indicates a potential degradation of the level of safety of the plant, warranting a UE classification. This EAL excludes inadvertent criticalities that occur during planned reactivity changes associated with reactor startups (e.g., criticality earlier than estimated).

Escalation would be by the Fission Product Barrier Table, as appropriate to the operating mode at the time of the event.

CGS Basis:

This condition may be identified by (Ref. 1):

" Period meters A, B, C, and D on Main Control Room panel H13-P603 and H13-P633

• Amber period lights for each of the four channels on Main Control Room panel H13-P603 and H13-P633

• Annunciator P603-A7 SRM PERIOD FAST which is actuated at +60 seconds by any one of the four SRM channels (2)

CGS Basis Reference(s):

1. FSAR Section 7.7.1.7

2. PPM 4.603.A7 Annunciator Response 5-6, SOURCE RANGE MONITORS FAST PERIOD Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 185 of 274 Category: S - System Malfunction Subcategory: 3 - ATWS / Criticality Initiating Condition: Automatic scram fails to shut down the reactor and the manual actions taken from the reactor control console are successful in shutting down the reactor EAL:

SA3.1 Alert An automatic scram failed to shut down the reactor AND Manual actions taken at the reactor control console (mode switch in shutdown, manual push buttons or ARI) successfully shut down the reactor as indicated by reactor power LE 5% (APRM downscale)

Mode Applicability:

1 - Power Operations, 2 - Startup NEI 99-01 Basis:

Manual scram actions taken at the reactor control console are any set of actions by the reactor operator(s) which causes or should cause control rods to be rapidly inserted into the core and shuts down the reactor.

This condition indicates failure of the automatic protection system to scram the reactor. This condition is more than a potential degradation of a safety system in that a front line automatic protection system did not function in response to a plant transient. Thus the plant safety has been compromised because design limits of the fuel may have been exceeded. An Alert is indicated because conditions may exist that lead to potential loss of fuel clad or RCS and because of the failure of the Reactor Protection System to automatically shutdown the plant. '

If manual actions taken at the reactor control console fail to shutdown the reactor, the event would escalate to a Site Area Emergency.

Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 186 of 274 CGS Basis:

The first condition of this EAL identifies the need to cease critical reactor operations by actuation of the automatic Reactor Protection System (RPS) scram function. A reactor scram is automatically initiated by the Reactor Protection System (RPS) when certain continuously monitored parameters exceed predetermined setpoints. A reactor scram may be the result of manual or automatic action in response to any of the following conditions (Ref. 1):

S. .....

Condition [ .2............

l .. ...

s.. .P Trip

.. UX....!..............................................................................................................................

I...

IRM Flux Hi 120/125 scale

.R........

M. p . ................................................................................................

o. ................................................................

APRM Flux Hi Setdown 115%

APRM Flow Biased Simulated Thermal .58Wr+59%

Powe.............................................................................................................

r ...............................

.5...................M

.... ax ) .................... .. ... .............................................

APRM Neutron Flux Hi 118%

APRM Inop OPRM Trip OPRM Enabled AND Counts greater than or equal to 14 AND RPV. .L.vel. .. PBA greater than or equal to 1.11 PV Low Level 13 in R P... .V

. . s ~ q...........

.m ..........

_1.o_6 .. s g ...........

RPV D . .Hi Pressure

............... ..*.i.......

e..!...........

P.... .. 1060 psig I._._6.... S .I..................................................................

D.ry.

.. . .v. g. re. .... s.. u.. 9...68.p...........

6 p.ig ............................

..Sc r am .Disch .Vo lu m e .High .Level . ...7 7'"......

e.e.Iev..

e................................

.......................... e......................................

Governor Valve Fast Closure Emergency Trip Header less than 1250 psig Following a successful reactor scram, 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 period. 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-scram response from an automatic reactor scram 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. A successful scram has therefore occurred when there is sufficient rod insertion from the trip of RPS to bring the reactor power to or below the APRM downscale trip setpoint of 5%. For the purposes of this EAL, a successful automatic initiation of ARI that reduces reactor power to or below 5% is a not a successful automatic scram. If automatic actuation of ARI has occurred and caused reactor shutdown, the automatic RPS scram must have failed. ARI is a backup means of inserting control rods in the unlikely event that an automatic RPS scram signal exists but the reactor continues to generate significant power. (Ref. 2, 3, 4, 5, 6)

Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 187 of 274 For the purposes of emergency classification at the Alert level, successful manual scram actions are those which can be quickly performed from the reactor control console (i.e., mode switch, manual scram pushbuttons, and manual ARI actuation). Reactor shutdown achieved by use of the alternate control rod insertion methods of PPM 5.5.11 does not constitute a successful manual scram (Ref. 7).

Following any automatic RPS scram signal EOP flowcharts (Ref. 5, 6) prescribe insertion of redundant manual scram signals to back up the automatic RPS scram function and ensure reactor shutdown is achieved. Even if the first subsequent manual scram signal inserts all control rods to the full-in position immediately after the initial failure of the automatic scram, the lowest level of classification that must be declared is an Alert.

If the operator determines the reactor must be scrammed before one of the RPS setpoints is reached, procedures require that the Mode Switch first be placed in the shutdown position. Although manipulation of the Mode Switch is a manual action, the RPS logic trains are actuated as with an automatic RPS-initiated scram. If reactor power remains above the APRM downscale trip setpoint after the Mode Switch is placed in shutdown, RPS has failed and, as a minimum, an Alert emergency declaration is required. If subsequent actuation of the reactor scram pushbuttons and manual initiation of ARI do not reduce reactor power to or below the APRM downscale trip setpoint, a Site Area Emergency declaration is required under EAL SS3.1.

In the event that the operator identifies a reactor scram is imminent and initiates a successful manual reactor scram before the automatic scram setpoint is reached, no declaration is required. The successful manual scram of the reactor before it reaches its automatic scram setpoint or reactor scram signals caused by instrumentation channel failures do not lead to a potential fission product barrier loss. If manual reactor scram actions fail to reduce reactor power to or below 5% (Ref. 1, 5),

the event escalates to the Site Area Emergency under EAL SS3.1.

By procedure, operator actions include the initiation of an immediate manual scram following receipt of an automatic scram signal. If there are no clear indications that the automatic scram failed (such as a time delay following indications that a scram setpoint was exceeded), it may be difficult to determine if the reactor was shut down because of automatic scram or manual actions. If a subsequent review of the scram actuation indications reveals that the automatic scram did not cause the reactor to be shut down, then consideration should be given to evaluating the fuel for potential damage, and the reporting requirements of 50.72 should be considered for the transient event.

By definition, an operating mode change occurs when the Mode Switch is moved from the startup or run position to the shutdown position. The plant operating mode that existed at the time the event occurs (i.e., Power Operations or Startup), however, requires emergency classification of at least an Alert. The operating mode change associated with movement of the Mode Switch, by itself, does not justify failure to declare an emergency for ATWS events.

CGS Basis Reference(s):

1. Technical Specifications Table 3.3.1.1-1

2. FSAR Section 7.2.1.1.9

3. FSAR Section 7.2.1.1.10

4. FSAR Section 7.4.1.6

5. PPM 5.1.1 RPV Control

6. PPM 5.1.2 RPV Control-ATWS

7. PPM 5.5.11 Alternate Control Rod Insertions Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 188 of 274 Category: S - System Malfunction Subcategory: 3 - ATWS / Criticality Initiating Condition: Automatic scram fails to shut down the reactor and manual actions taken from the reactor control console are not successful in shutting down the reactor.

EAL:

SS3.1 Site Area Emergency An automatic scram failed to shut down the reactor AND Manual actions taken at the reactor control console (mode switch in shutdown, manual push buttons and ARI) do not shut down the reactor as indicated by reactor power GT 5% (APRM downscale)

Mode Aonlicability:

1 - Power Operations, 2 - Startup NEI 99-01 Basis:

Under these conditions, the reactor is producing more heat than the maximum decay heat load for which the safety systems are designed and efforts to bring the reactor subcritical are unsuccessful. A Site Area Emergency is warranted because conditions exist that lead to imminent loss or potential loss of both fuel clad and RCS.

Manual scram actions taken at the reactor control console are any set of actions by the reactor operator(s) at which causes or should cause control rods to be rapidly inserted into the core and shuts down the reactor.

Manual scram actions are not considered successful if action away from the reactor control console is required to scram (trip) the reactor. This EAL is still applicable even if actions taken away from the reactor control console are successful in shutting the reactor down because the design limits of the fuel may have been exceeded or because of the gross failure of the Reactor Protection System to shut down the plant.

Escalation of this event to a General Emergency would be due to a prolonged condition leading to an extreme challenge to either core-cooling or heat removal.

CGS Basis:

This EAL addresses any automatic reactor scram signal followed by a manual scram 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.

For the purposes of emergency classification at the Site Area Emergency level, successful manual scram actions are those which can be quickly performed from the reactor control console (i.e., Mode Switch, manual scram pushbuttons and ARI actuation). Reactor shutdown achieved by use of the Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 189 of 274 alternate control rod insertion methods of PPM 5.5.11 does not constitute a successful manual scram (Ref. 1).

The APRM downscale trip setpoint (5%) is a minimum reading on the power range scale that indicates power production (Ref. 2, 3). It also approximates the decay heat which the shutdown systems were designed to remove and is indicative of a condition requiring immediate response to prevent subsequent core damage. At or below the APRM downscale trip setpoint, plant response will be similar to that observed during a normal shutdown. Nuclear instrumentation (APRM/IRM) indications or other reactor parameters (e.g., number of open SRVs, number of open main turbine bypass valves, main steam flow, RPV pressure and wetwell temperature trend, etc.) can be used to determine if reactor power is greater than 5% power (Ref. 4).

By definition, an operating mode change occurs when the Mode Switch is moved from the startup or run position to the shutdown position. The plant operating mode that existed at the time the event occurs (i.e., Power Operations or Startup), however, requires emergency classification of at least an Alert. The operating mode change associated with movement of the Mode Switch, by itself, does not justify failure to declare an emergency for ATWS events.

Escalation of this event to a General Emergency would be under EAL SG3.1 or Emergency Director judgment.

CGS Basis Reference(s):

1. PPM 5.5.11 Alternate Control Rod Insertions

2. Technical Specifications Table 3.3.1.1-1

3. PPM 5.1.1 RPV Control

4. PPM 5.0.10 Flowchart Training Manual Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 190 of 274 Category: S - System Malfunction Subcategory: 3 - ATWS / Criticality Initiating Condition: Automatic scram and all manual actions fail to shut down the reactor and indication of an extreme challenge to the ability to cool the core exists EAL:

SG3.1 General Emergency An automatic scram failed to shut down the reactor AND All manual actions do not shut down the reactor as indicated by reactor power GT 5% (APRM downscale)

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

RPV level cannot be restored and maintained above -183 in. or cannot be determined OR WW temperature and RPV pressure cannot be maintained below the HCTL (EOP Figure C)

Mode Applicability:

1 - Power Operations, 2 - Startup NEI 99-01 Basis:

Under these conditions, the reactor is producing more heat than the maximum decay heat load for which the safety systems are designed and efforts to bring the reactor subcritical are unsuccessful.

In the event either of these challenges exists at a time that the reactor has not been brought below the power associated with the safety system design a core melt sequence exists. In this situation, core degradation can occur rapidly. For this reason, the General Emergency declaration is intended to be anticipatory of the fission product barrier table declaration to permit maximum off-site intervention time.

CGS Basis:

This EAL addresses the following:

• Any automatic reactor scram signal followed by a manual scram 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 (EAL SS3.1), and

• Indications that either core cooling is extremely challenged or heat removal is extremely challenged.

Reactor shutdown achieved by use of the alternate control rod insertion methods of PPM 5.5.11 is also credited as a successful manual scram provided reactor power can be reduced below the APRM Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 191 of 274 downscale trip setpoint before indications of an extreme challenge to either core cooling or heat removal exist. (Ref. 1)

The APRM downscale trip setpoint (5%) is a minimum reading on the power range scale that indicates power production (Ref. 2, 3). It also approximates the decay heat which the shutdown systems were designed to remove and is indicative of a condition requiring immediate response to prevent subsequent core damage. At or below the APRM downscale trip setpoint, plant response will be similar to that observed during a normal shutdown. Nuclear instrumentation (APRM/IRM) indications or other reactor parameters (e.g., number of open SRVs, number of open main turbine bypass valves, main steam flow, RPV pressure and wetwell temperature trend, etc.) can be used to determine if reactor power is greater than 5% power (Ref. 4).

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.

By definition, an operating mode change occurs when the Mode Switch is moved from the startup or run position to the shutdown position. The plant operating mode that existed at the time the event occurs (i.e., Power Operations or Startup), however, requires emergency classification of at least an Alert. The operating mode change associated with movement of the Mode Switch, by itself, does not justify failure to declare an emergency for ATWS events.

Indication that core cooling is extremely challenged is manifested by:

RPV level cannot be restored and maintained above -183 in. (Ref. 5, 6, 7). The Minimum Steam Cooling RPV Water Level (MSCRWL) is the lowest RPV water level at which the covered portion of the reactor core will generate sufficient steam to preclude any clad temperature in the uncovered portion of the core from exceeding 1500°F (Ref. 8). Consistent with the EOP definition of "cannot be restored and maintained," the determination that RPV level cannot be restored and maintained above the MSCRWL may be made at, before, or after RPV level actually decreases to this point.

RPV level cannot be determined. When RPV level cannot be determined, EOPs require entry to flowchart PPM 5.1.6, Reactor Flooding - ATWS (Ref. 9). RPV level indication provides the primary means of knowing if adequate core cooling is being maintained. When all means of determining RPV level are unavailable, the fuel clad barrier is threatened and reliance on alternate means of assuring adequate core cooling must be attempted. The instructions in PPM 5.1.6 specify these means, which include emergency depressurization of the RPV and injection into the RPV at a rate needed to flood to the elevation of the main steam lines or hold RPV pressure above the Minimum Steam Cooling Pressure. (Ref. 8)

Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 192 of 274 The Heat Capacity Temperature Limit (HCTL) is the highest wetwell temperature at which emergency RPV depressurization will not raise:

" Wetwell temperature above the maximum temperature capability of the wetwell and equipment within the wetwell which may be required to operate when the RPV is pressurized (275°F), or

• Wetwell pressure above the Primary Containment Pressure Limit (PCPL), while the rate of energy transfer from the RPV to the containment is greater than the capacity of the containment vent.

The HCTL is used to preclude failure of the containment or equipment necessary for safe shutdown by assuring that RPV blowdown does not cause containment pressure to exceed the PCPL. (Ref. 10)

The HCTL is given in EOP flowchart Figure C. This threshold is met when EOP flowchart PPM 5.2.1 Step WT-5 is reached (Ref. 11). This condition addresses loss of functions required for hot shutdown with the reactor at pressure and temperature.

CGS Basis Reference(s):

1. PPM 5.5.11 Alternate Control Rod Insertions

2. Technical Specifications Table 3.3.1.1-1

3. PPM 5.1.1 RPV Control

4. PPM 5.0.10 Flowchart Training Manual

5. PPM 5.1.1 RPV Control

6. PPM 5.1.2 RPV Control - ATWS

7. Calculation NE-02-03-06 Attachment 10 RPV Variables 8 PPM 5.0.10 Flowchart Training Manual

9. PPM 5.1.6 RPV Flooding - ATWS

10. Calculation NE-02-03-06 Attachment 5

11. PPM 5.2.1 Primary Containment Control Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 193 of 274 Category: S - System Malfunction Subcategory: 4 - Inability to Reach or Maintain Shutdown Conditions Initiating Condition: Inability to reach required shutdown within Technical Specification limits EAL:

SU4.1 Unusual Event Plant is not brought to required operating mode within Technical Specifications LCO action statement time Mode Applicability:

1 - Power Operations, 2 - Startup, 3 - Hot Shutdown NEI 99-01 Basis:

Limiting Conditions of Operation (LCOs) require the plant to be brought to a required operating mode when the Technical Specification required configuration cannot be restored. Depending on the circumstances, this may or may not be an emergency or precursor to a more severe condition. In any case, the initiation of plant shutdown required by the site Technical Specifications requires a four hour report under 10 CFR 50.72 (b) Non-emergency events. The plant is within its safety envelope when being shut down within the allowable action statement time in the Technical Specifications. An immediate UE is required when the plant is not brought to the required operating mode within the allowable action statement time in the Technical Specifications. Declaration of a UE is based on the time at which the LCO-specified action statement time period elapses under the site Technical Specifications and is not related to how long a condition may have existed.

CGS Basis:

None CGS Basis Reference(sl:

1. Columbia Generating Station Technical Specifications Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 194 of 274 Category: S - System Malfunction Subcategory: 5 - Instrumentation Initiating Condition: Unplanned loss of safety system annunciation or indication in the Control Room for 15 minutes or longer EAL:

SU5.1 Unusual Event UNPLANNED loss of GT approximately 75% of the Control Room safety system annunciators or indicators on the following panels for GE 15 min. (Note 3):

" P601

" P602

• P603

" Bd. C NOTE 3: The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the condition will likely exceed the applicable time.

Mode Applicability:

1 - Power Operations, 2 - Startup, 3 - Hot Shutdown NEI 99-01 Basis:

This IC and its associated EAL are intended to recognize the difficulty associated with monitoring changing plant conditions without the use of a major portion of the annunciation or indication equipment.

Recognition of the availability of computer based indication equipment is considered.

"Planned" loss of annunciators or indicators includes scheduled maintenance and testing activities.

Quantification is arbitrary, however, it is estimated that if approximately 75% of the safety system annunciators or indicators are lost, there is an increased risk that a degraded plant condition could go undetected. It is not intended that plant personnel perform a detailed count of the instrumentation lost but use the value as a judgment threshold for determining the severity of the plant conditions.

It is further recognized that most plant designs provide redundant safety system indication powered from separate uninterruptible power supplies. While failure of a large portion of annunciators is more likely than a failure of a large portion of indications, the concern is included in this EAL due to difficulty associated with assessment of plant conditions. The loss of specific, or several, safety system indicators should remain a function of that specific system or component operability status. This will be addressed by the specific Technical Specification. The initiation of a Technical Specification imposed plant shutdown related to the instrument loss will be reported via 10CFR 50.72. If the Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 195 of 274 shutdown is not in compliance with the Technical Specification action, the UE is based on SU4.1 "Inability to Reach Required Shutdown Within Technical Specification Limits."

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

This UE will be escalated to an Alert based on a concurrent loss of compensatory indications.

CGS Basis:

Safety system annunciation and indication considered in this EAL is found on Main Control Room Panels P601, P602, P603, and Bd. C. The other annunciators and indicators are important to plant operation but are not important to safety (Ref. 1-6). If the loss of these annunciators or indicators is accompanied by either the unavailability of compensatory indications (i.e., Process Computers and Graphic Display System) or the existence of a significant transient, the event may escalate to an Alert under EAL SA5.1.

The judgment of the Shift Manager should be used as the threshold for determining the severity of the plant conditions.

CGS Basis Reference(s):

1. EWD-49E-050 Electrical Wiring Diagram Annunciator Engraving NSSS Control Room Panels, P601

2. EWD-49E-051 Electrical Wiring Diagram Annunciator Engraving NSSS Control Room Panels, P602

3. EWD-49E-052 Electrical Wiring Diagram Annunciator Engraving NSSS Control Room Panels, P603

4. EWD-49E-060 Electrical Wiring Diagram Annunciator Engraving BOP Control Room Panels, Bd C1, C2, C3

5. EWD-49E-052 Electrical Wiring Diagram Annunciator Engraving BOP Control Room Panels, Bd C4, C5

6. A521-1 Control Room Plan Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 196 of 274 Category: S - System Malfunction Subcategory: 5 - Instrumentation Initiating Condition: Unplanned loss of safety system annunciation or indication in the Control Room with either (1) a significant transient in progress, or (2) compensatory indicators unavailable EAL:

SA5.1 Alert UNPLANNED loss of GT approximately 75% of the Control Room safety system annunciators or indicators on the following panels for GE 15 min. (Note 3):

• P601

• P602

• P603

" Bd. C AND EITHER:

A SIGNIFICANT TRANSIENT is in progress, Table S-1 OR Compensatory indications are unavailable (Process Computers and GDS)

NOTE 3 The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the condition will likely exceed the applicable time.

Table S-1 Significant Transients

" Reactor scram

" Runback GT 25% thermal power

" Electrical load rejection GT 25% full electrical load

" ECCS injection

" Thermal power oscillations GT 10%

Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 197 of 274 Mode Applicability:

1 - Power Operations, 2 - Startup, 3 - Hot Shutdown NEI 99-01 Basis:

This EAL is intended to recognize the difficulty associated with monitoring changing plant conditions without the use of a major portion of the annunciation or indication equipment during a significant transient.

"Planned" loss of annunciators or indicators includes scheduled maintenance and testing activities.

Quantification is arbitrary, however, it is estimated that if approximately 75% of the safety system annunciators or indicators are lost, there is an increased risk that a degraded plant condition could go undetected. It is not intended that plant personnel perform a detailed count of the instrumentation lost but use the value as a judgment threshold for determining the severity of the plant conditions. It is also not intended that the Shift Manager be tasked with making a judgment decision as to whether additional personnel are required to provide increased monitoring of system operation.

It is further recognized that most plant designs provide redundant safety system indication powered from separate uninterruptible power supplies. While failure of a large portion of annunciators is more likely than a failure of a large portion of indications, the concern is included in this EAL due to difficulty associated with assessment of plant conditions. The loss of specific, or several, safety system indicators should remain a function of that specific system or component operability status. This will be addressed by the specific Technical Specification. The initiation of a Technical Specification imposed plant shutdown related to the instrument loss will be reported via 10 CFR 50.72. If the shutdown is not in compliance with the Technical Specification action, the UE is based on SU4.1 "Inability to Reach Required Shutdown Within Technical Specification Limits."

"Compensatory indications" in this context includes computer based information such as the process computer system and GDS. If both a major portion of the annunciation system and all computer monitoring are unavailable, the Alert is required.

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

This Alert will be escalated to a Site Area Emergency if the operating crew cannot monitor the transient in progress due to a concurrent loss of compensatory indications with a significant transient in progress during the loss of annunciation or indication.

CGS Basis:

The Process Computer System and Graphic Display System (GDS, which displays SPDS required information) serve as a redundant compensatory indicators which may be utilized in lieu of normal Main Control Room annunciators and indicators (Ref. 1-5). Safety system annunciation and indication considered in this EAL is found on Main Control Room Panels P601, P602, P603, and Bd. C. The other annunciators and indicators are important to plant operation but are not important to safety (Ref. 6-11).

The judgment of the Shift Manager should be used as the threshold for determining the severity of the plant conditions.

Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 198 of 274 Significant transients are listed in Table S-1 and include response to automatic or manually initiated functions such as scrams, runbacks involving greater than 25% thermal power change, electrical load rejections of greater than 25% full electrical load, ECCS injections, or thermal power oscillations of 10%

or greater.

If the operating crew cannot monitor the transient in progress, the Alert escalates to a Site Area Emergency under EAL SS5.1.

CGS Basis Reference(s):

1. FSAR Section 7.7.1

2. FSAR Section 7.7.1.22

3. ABN-COMPUTER

4. SOP-COMPUTER-OPS Plant Process Computer (PPC)

5. SOP-GDS-OPS Graphics Display System

6. EWD-49E-050 Electrical Wiring Diagram Annunciator Engraving NSSS Control Room Panels, P601

7. EWD-49E-051 Electrical Wiring Diagram Annunciator Engraving NSSS Control Room Panels, P602

8. EWD-49E-052 Electrical Wiring Diagram Annunciator Engraving NSSS Control Room Panels, P603

9. EWD-49E-060 Electrical Wiring Diagram Annunciator Engraving BOP Control Room Panels, Bd C1, C2, C3

10. EWD-49E-052 Electrical Wiring Diagram Annunciator Engraving BOP Control Room Panels, Bd C4, C5

11. A521 -1 Control Room Plan Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 199 of 274 Category: S - System Malfunction Subcategory: 5 - Instrumentation Initiating Condition: Inability to monitor a significant transient in progress EAL:

SS5.1 Site Area Emergency Loss of GT approximately 75% of the Control Room safety system annunciators or indicators on the following panels for GE 15 min. (Note 3):

• P601

" P602

" P603

" Bd. C AND A SIGNIFICANT TRANSIENT is in progress, Table S-1 AND Compensatory indications are unavailable (Process Computers and GDS)

NOTE 3: The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the condition will likely exceed the applicable time.

TableS-I Sig nificant Transients

• Reactor scram

• Runback GT 25% thermal power

• Electrical load rejection GT 25% full electrical load

" ECCS injection

" Thermal power oscillations GT 10%

Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 200 of 274 Mode Applicability:

1 - Power Operations, 2 - Startup, 3 - Hot Shutdown NEI 99-01 Basis:

This EAL is intended to recognize the threat to plant safety associated with the complete loss of capability of the control room staff to monitor plant response to a significant transient.

"Planned" and "unplanned" actions are not differentiated since the loss of instrumentation of this magnitude is of such significance during a transient that the cause of the loss is not an ameliorating factor.

Quantification is arbitrary, however, it is estimated that if approximately 75% of the safety system annunciators or indicators are lost, there is an increased risk that a degraded plant condition could go undetected. It is not intended that plant personnel perform a detailed count of the instrumentation lost but use the value as a judgment threshold for determining the severity of the plant conditions. It is also not intended that the Shift Manager be tasked with making a judgment decision as to whether additional personnel are required to provide increased monitoring of system operation.

It is further recognized that most plant designs provide redundant safety system indication powered from separate uninterruptible power supplies. While failure of a large portion of annunciators is more likely than a failure of a large portion of indications, the concern is included in this EAL due to difficulty associated with assessment of plant conditions. The loss of specific, or several, safety system indicators should remain a function of that specific system or component operability status. This will be addressed by the specific Technical Specification. The initiation of a Technical Specification imposed plant shutdown related to the instrument loss will be reported via 10 CFR 50.72. If the shutdown is not in compliance with the Technical Specification action, the UE is based on SU4.1 "Inability to Reach Required Shutdown Within Technical Specification Limits."

A Site Area Emergency is considered to exist if the control room staff cannot monitor safety functions needed for protection of the public while a significant transient is in progress.

Site specific indications needed to monitor safety functions necessary for protection of the public must include control room indications, computer generated indications and dedicated annunciation capability.

"Compensatory indications" in this context includes computer based information such as the process computer system and GDS. This should include all computer systems available for this use depending on specific plant design and subsequent retrofits.

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

Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 201 of 274 CGS Basis:

The Process Computer System and Graphic Display System (GDS, which displays SPDS required information) serve as a redundant compensatory indicators which may be utilized in lieu of normal Main Control Room annunciators and indicators (Ref. 1-5). Safety system annunciation and indication considered in this EAL is found on Main Control Room Panels P601, P602, P603, and Bd. C. The other annunciators and indicators are important to plant operation but are not important to safety (Ref. 6-11).

Significant transients are listed in Table S-1 and include response to automatic or manually initiated functions such as trips, runbacks involving greater than 25% thermal power change, electrical load rejections of greater than 25% full electrical load, ECCS injections, or thermal power oscillations of 10% or greater.

CGS Basis Reference(s):

1. FSAR Section 7.7.1

2. FSAR Section 7.7.1.22

3. ABN-COMPUTER

4. SOP-COMPUTER-OPS Plant Process Computer (PPC)

5. SOP-GDS-OPS Graphics Display System

6. EWD-49E-050 Electrical Wiring Diagram Annunciator Engraving NSSS Control Room Panels, P601

7. EWD-49E-051 Electrical Wiring Diagram Annunciator Engraving NSSS Control Room Panels, P602

8. EWD-49E-052 Electrical Wiring Diagram Annunciator Engraving NSSS Control Room Panels, P603

9. EWD-49E-060 Electrical Wiring Diagram Annunciator Engraving BOP Control Room Panels, Bd C1, C2, C3

10. EWD-49E-052 Electrical Wiring Diagram Annunciator Engraving BOP Control Room Panels, Bd C4, C5

11. A521 -1 Control Room Plan Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 202 of .274 Category: S - System Malfunction Subcategory: 6 - Communications Initiating Condition: Loss of all onsite or offsite communications capabilities EAL:

SU6.1 Unusual Event Loss of all Table S-2 onsite communications capability affecting the ability to perform routine operations OR Loss of all Table S-2 off site communications affecting the ability to perform offsite notifications Table S-2 Communications Systems System (internal) Offsite Onsite (external)

Plant Public Address (PA) System X Plant Telephone System X Plant Radio System Operations and X Security Channels Off site calling capability from the X Control Room via direct telephone and fax lines Long distance calling capability on X the commercial phone system Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 203 of 274 Mode Applicability:

1 - Power Operations, 2 - Startup, 3 - Hot Shutdown NEI 99-01 Basis:

The purpose of this IC and its associated EALs is to recognize a loss of communications capability that either defeats the plant operations staff ability to perform routine tasks necessary for plant operations or the ability to communicate issues with off-site authorities.

The availability of one method of ordinary off-site communications is sufficient to inform federal, state, and local authorities of plant problems. This EAL is intended to be used only when extraordinary means (e.g., relaying of information from non-routine radio transmissions, individuals being sent to off-site locations, etc.) are being used to make communications possible.

CGS Basis:

Onsite and offsite communications include one or more of the systems listed in Table S-2 (Ref. 1, 2).

Public Address (PA) System The public address system provides a way of contacting personnel in the various buildings of the plant and locations of the site that might be inaccessible using other means of communication. The building-wide alarm system alerts (via the public address system speakers) operating personnel to fire hazards and other trouble conditions for which plant management finds it necessary to alert plant personnel.

Plant Telephone System This system consists of interconnections to the public telephone network (and trunks to the PBX) with individual direct lines that provide inward and outward dialing access to most plant locations.

Plant Radio System Operations and Security Channels The radio communications system is used for communications with personnel involved in maintenance and security in and around the plant complex by means of hand-held portable radio units, mobile radio units, and paging receivers. The telephone link to BPA provides a direct communication link to the BPA Dittmer Control Center. The radio communications system provides a communications link for security and emergency communications to local law enforcement agencies and emergency control centers.

Offsite calling capability from the Control Room via direct telephone and fax lines This communications method includes following dedicated phone networks that are available for emergency communications in addition to the normal Energy Northwest phone network:

• Energy Northwest Emergency Center Network

• Response Agency Network

• NRC Emergency Notification System Various locations such as the Control Room, Technical Support Center, Emergency Operations Facility, Joint Information Center, Department of Energy-RL, Washington State Emergency Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 204 of 274 Operations Center, Oregon State Emergency Coordination Center and the Benton and Franklin County Emergency Operations Centers have facsimile transceivers. The facsimile transceivers enable the transmission and receipt of printed material. The facsimile system which connects the Energy Northwest emergency centers with the county and state emergency centers uses dedicated phone lines.

Long distance calling capability on the commercial phone system The Energy Northwest Richland phone system is a computer based, software controlled telephone exchange (Computerized Branch Exchange). It is equipped with redundant computerized processor units and is served by an uninterruptible power supply. The direct-dial private telephone system provides communication between the Hanford Energy Northwest facilities. The phone system is arranged such that plant telephones can reach other Energy Northwest facilities by direct-dialing and without the need of an operator.

This EAL is the hot condition equivalent of the cold condition EAL CU5.1.

CGS Basis Reference(s):

1. Emergency Plan Section 6.6

2. FSAR Section 9.5.2 Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 205 of 274 Category: S - System Malfunction Subcategory: 7 - Fuel Clad Degradation Initiating Condition: Fuel clad degradation EAL:

SU7.1 Unusual Event Fuel clad degradation GT Technical Specifications as indicated by H13.P602.A5.3-3 (SJAE CONDSR OUTLET RAD HI HI) activated Mode Applicability:

1 - Power Operations, 2 - Startup, 3 - Hot Shutdown NEI 99-01 Basis:

This EAL is included because it is a precursor of more serious conditions and, as result, is considered to be a potential degradation of the level of safety of the plant.

Escalation of this EAL to the Alert level is via the Fission Product Barriers.

This threshold addresses site-specific radiation monitor readings that provide indication of a degradation of fuel clad integrity.

CGS Basis:

The Steam Jet Air Ejector radiation monitor setpoint provides reasonable assurance that the total body exposure to an individual at the exclusion area boundary will not exceed a small fraction of the limits of 10CFR100 in the event of an inadvertent release via the condenser air ejector. (Ref. 1)

SJAE CONDSR OUTLET RAD HI HI monitor and alarm, OG-RIS-612 (GE 2300 mR/hr), reflects the offgas effluent and, therefore, may be one of the first indicators of degrading fuel conditions. The alarm is confirmed by verification of greater than the current alarm setpoint on Recorder OG-RIS-612 on Panel P604 or high offgas pre-treatment air activity (determined by sample results) greater than limits specified in Technical Specification.

If OG-RIS-612 and OG-RR-604 are reading off-scale high, the alarm may be confirmed by a significant increase in the Main Steam Line radiation monitors (MS-RIS-610A-D) on H13-P606 and H13-P633. (Ref. 2)

CGS Basis Reference(s):

1. Technical Specifications 3.7.5

2. PPM 4.602.A5 ANNUNCIATOR RESPONSE, P602 ANNUNCIATOR A5 3-3 Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 206 of 274 Category: S - System Malfunction Subcategory: 7- Fuel Clad Degradation Initiating Condition: Fuel clad degradation EAL:

SU7.2 Unusual Event Coolant activity GT 0.2 pCi/gm dose equivalent 1-131 Mode Applicability:

1 - Power Operations, 2 - Startup, 3 - Hot Shutdown NEI 99-01 Basis:

This EAL is included because it is a precursor of more serious conditions and, as result, is considered to be a potential degradation of the level of safety of the plant.

Escalation of this EAL to the Alert level is via the Fission Product Barriers.

This threshold addresses coolant samples exceeding coolant technical specifications for transient iodine spiking limits.

CGS Basis:

The limits on the specific activity of the primary coolant ensure that the 2-hour thyroid and whole body doses at the site boundary, resulting from an Main Steam Line Break (MSLB) outside containment during steady state operation, will not exceed the dose guidelines of 10 CFR 50.67.

CGS Basis Reference(s):

1. Technical Specifications 3.4.8 Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 207 of 274 Category: S - System Malfunction Subcategory: 8- RCS Leakage Initiating Condition: RCS leakage EAL:

SU8.1 Unusual Event Unidentified or pressure boundary leakage GT 10 gpm or upscale high indicated on recorder EDR-FRS-623 Pen 1 (P632) (non-RCC)

OR Identified leakage GT 25 gpm indicated on recorder EDR-FRS-623 Pen 2 (P632) (Note 6)

NOTE 6: See Table F-i, Fission Product Barrier Matrix, for possible escalation above the Unusual Event due to RCS Leakage Mode Applicability:

1 - Power Operations, 2 - Startup, 3 - Hot Shutdown NEI 99-01 Basis:

This EAL is included as a UE because it may be a precursor of more serious conditions and, as result, is considered to be a potential degradation of the level of safety of the plant. The 10 gpm value for the unidentified or pressure boundary leakage was selected as it is observable with normal control room indications. Lesser values must generally be determined through time-consuming surveillance tests (e.g., mass balances).

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

The EAL for identified leakage is set at a higher value due to the lesser significance of identified leakage in comparison to unidentified or pressure boundary leakage. In either case, escalation of this EAL to the Alert level is via Fission Product Barrier Degradation EALs.

CGS Basis:

Unidentified leakage is defined to be all leakage into the drywell that is not identified leakage.

Pressure boundary leakage is defined to be leakage through a nonisolable fault in a RCS component body, pipe wall, or vessel wall.

Identified leakage is defined to be leakage into the drywell such as that from pump seals or valve packing, that is captured and conducted to a sump or collecting tank; or leakage into the drywell 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. (Ref. 1)

Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 208 of 274 The Leak Detection (LD) system is designed to monitor leakage from the reactor coolant pressure boundary and to initiate alarms and/or isolate this leakage when limits are exceeded. Systems, or parts of systems, that are in direct communication with the reactor vessel (form part of the primary coolant pressure boundary) are provided with leakage detection systems. (Ref. 2-8)

Leaks within the drywell are detected by monitoring for abnormally high:

• Pressure or temperature inside the drywell

• Fill up rates of equipment and floor drain sumps

• Cooling water temperature differences across drywell coolers

• Levels of fission products in the drywell atmosphere The LD system provides leakage indication and alarms for the following items inside primary containment:

• Containment area temperatures

• RPV head seal leak off

• Valve stem leakage RRC-V-60A & B

• Drywell equipment and floor drain sumps Drain flow from the drywell equipment and floor drain sumps is monitored and recorded (EDR-FRS-623) on P632 and this recorder energizes "high flow" alarms on P601. The drywell floor drain sump (unidentified leakage) gravity drains to the reactor building floor drain sump and the drywell equipment drain sump (identified leakage) gravity drains to the reactor building equipment drain sump.

Drywell equipment sump drain flow in excess of 25 gpm or floor drain sump flow in excess of 5 gpm is annunciated on P601 (A3-6.6 or 6.5) as "LEAK DET DRYWELL EQUIP DRAIN FLOW HIGH" or "LEAK DET DRYWELL FLOOR DRAIN FLOW HIGH" respectively.

There is one reactor building equipment drain sump (EDR-SUMP-R5) and four reactor building floor drain sumps (FDR-SUMP-R1 through R4). The reactor building equipment drain sump (EDR-SUMP-R5) accepts leakage from expected equipment leakage points in the reactor building in addition to receiving all of the flow from the drywell equipment drain sump. The four reactor building floor drain sumps collect the general leakage throughout the building. Reactor building floor drain sump FDR-SUMP-R3 also receives flow from the drywell floor drain sump. All reactor building sumps are equipped with two timers, a sump fill timer and a sump pump out timer. The location of these sumps are: R1 in the RHR A room, R2 in the RHR B room, R3 in the HPCS room, R4 in the RHR C room and R5 in the CRD room.

Sump fill timers are installed to alert the Control Room if it takes an abnormally short time to fill a sump (the time between sum pump runs). The timers are set for the amount of time it normally takes between the time the sump pump stops (the sump is empty) until it starts again (the sump is full). This time is different for each sump due to the different areas that drain into each sump; sump size Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 209 of 274 variations and sump pump flow rates. If the sump pump starts before the timer times out it causes an alarm in the Control Room on P601, otherwise it simply times out.

The alarm appears on Panel 601 (A2 or A3) in the main Control Room. The alarm reads "LEAK DET REACTOR BLDG EQUIP SUMP LEAKAGE HIGH" or "LEAK DET REACTOR BLDG FLOOR SUMP R1 (R2, R3, or R4) LEAKAGE HIGH".

The Reactor Building sumps also have high level and temperature alarms on P601.

RCC is not considered part of RCS leakage for this EAL.

For classification under this EAL, RCS leakage includes a broken SRV tailpipe that is discharging into the drywell or wetwell airspace. Once the SRV is closed, however, this RCS leakage path is considered isolated.

Escalation of this EAL to the Alert level is via Category F, Fission Product Barrier Degradation, EAL FA1.1. Note 6 has been added to remind the EAL-user to review Table F-1 for possible escalation to higher emergency classifications.

CGS Basis Reference(s):

1. Technical Specification 1.1

2. Technical Specifications 3.4.7

3. FSAR Section 5.2.5

4. FSAR Section 7.6.1

5. ABN-LEAKAGE Reactor Coolant Leakage

6. SOP-EDR-OPS Equipment Drain System Operation

7. SOP-FDR-OPS Floor Drain System Operation

8. PPM 10.27.35 Leakage Surveillance And Prevention Program Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 210 of 274 Categqory E - ISFSI EAL Group: ANY (EALs in this category are applicable to any plant condition, hot or cold.)

An independent spent fuel storage installation (ISFSI) is a complex that is designed and constructed for the interim storage of spent nuclear fuel and other radioactive materials associated with spent fuel storage. A significant amount of the radioactive material contained within a cask/canister must escape its packaging and enter the biosphere for there to be a significant environmental effect resulting from an accident involving the dry storage of spent nuclear fuel. Formal offsite planning is not required because the postulated worst-case accident involving an ISFSI has insignificant consequences to the public health and safety.

An Unusual Event is declared on the basis of the occurrence of an event of sufficient magnitude that a loaded cask confinement boundary is damaged or violated. This includes classification based on a loaded fuel storage cask/canister confinement boundary loss leading to the degradation of the fuel during storage or posing an operational safety problem with respect to its removal from storage.

A hostile security event that leads to a potential loss in the level of safety of the ISFSI is a classifiable event under Security category EAL HA4.1.

Minor surface damage that does not affect storage cask/canister boundary is excluded from the scope of these EALs.

Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 211 of 274 Category: E - ISFSI Sub-category: None Initiating Condition: Damage to a loaded cask confinement boundary EAL:

EU1.1 Unusual Event Damage to a loaded Multi Purpose Container (MPC) CONFINEMENT BOUNDARY Mode Applicability:

Mode Not Applicable NEI 99-01 Basis:

An Unusual Event in this EAL is categorized on the basis of the occurrence of an event of sufficient magnitude that a loaded MPC confinement boundary is damaged or violated. This includes classification based on a loaded fuel storage cask confinement boundary loss leading to the degradation of the fuel during storage or posing an operational safety problem with respect to its removal from storage.

CGS Basis:

The CGS ISFSI utilizes the HI-STORM 100 Cask System for dry spent fuel storage.

The HI-STORM 100 Cask System (the cask) consists of the following components: (1) interchangeable multi-purpose canisters (MPCs), which contain the fuel; (2) a storage overpack (HI-STORM), which contains the MPC during storage; and (3) a transfer cask (HI-TRAC), which contains the MPC during loading, unloading, and transfer operations. The cask stores up to 68 BWR fuel assemblies (Ref. 1, 2).

The MPC is the confinement system for the stored fuel. It is a welded, cylindrical canister with a honeycombed fuel basket, a baseplate, a lid, a closure ring, and the canister shell. The canister shell, baseplate, lid, vent and drain port cover plates, and closure ring are the main confinement boundary components. The honeycombed basket, which is equipped with neutron absorbers, provides criticality control (Ref. 1, 2).

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 Multi Purpose Canister (MPC) allowing leakage from the MPC confinement boundary.

Abnormal radiation levels outside the MPC would be an indicator that the cask confinement boundary has been damaged.

CGS Basis Reference(s):

1. Certificate of Compliance for Spent Fuel Storage Casks No. 1014 Amendment 2

2. Certificate of Compliance No. 1014 Appendix A Technical Specifications for the HI-STORM 100 Cask System Section 1.1 Definitions Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 212 of 274 Category F - Fission Product Barrier Degradation EAL Group: Hot Conditions (RCS temperature GT 2000 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): Zirconium tubes which house the ceramic uranium oxide pellets along with the end plugs which are welded into each end of the fuel rods comprise the FC barrier..

B. Reactor Coolant System (RCS): The reactor vessel shell, vessel head, CRD housings, vessel nozzles and penetrations, and all primary systems directly connected to the RPV up to the outermost primary containment isolation valve comprise the RCS barrier.

C. Containment (PC): The drywell, the suppression chamber/pool, their respective interconnecting paths, and other connections up to and including the outermost containment isolation valves comprise the PC barrier.

The EALs in this category require evaluation of the loss and potential loss thresholds listed in the fission product barrier matrix of Table F-1 (Attachment 7.2). "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:

Unusual Event:

Any loss or any potential loss of PrimaryContainment Alert:

Any loss or any potential loss of either Fuel Clad or RCS Site Area Emergency:

Loss or potential loss of any two barriers General Emergency:

Loss of any two barriersand loss or potential loss of third barrier Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 213 of 274 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 Primary Containment Barrier. Unusual Event ICs associated with RCS and Fuel Clad Barriers are addressed under System Malfunction ICs.

• At the Site Area Emergency level, there must be some ability to dynamically assess how far present conditions are from the threshold for a General Emergency. For example, if Fuel Clad and RCS Barrier "Loss" EALs existed, that, in addition to offsite dose assessments, would require continual assessments of radioactive inventory and containment integrity.

Alternatively, if both Fuel Clad and RCS Barrier "Potential Loss" EALs existed, the Emergency Director would have more assurance that there was no immediate need to escalate to a General Emergency.

• The ability to escalate the emergency classification as an event deteriorates must be maintained. For example, RCS leakage steadily increasing would represent an increasing risk to public health and safety.

" The Primary Containment Barrier should not be declared lost or potentially lost based on exceeding Technical Specification action statement criteria, unless there is an event in progress requiring mitigation by the Primary Containment barrier. When no event is in progress (Loss or Potential Loss of either Fuel Clad and/or RCS) the Primary Containment Barrier status is addressed by Technical Specifications.

Determine which combinations of the three barriers are lost or have a potential loss and use FU1.1, FA1.1, FS1.1 and FG1.1 to classify the event. Also an event or multiple events could occur which result in the conclusion that exceeding the loss or potential loss thresholds is imminent. In this imminent loss situation, use judgment and classify as if the thresholds are exceeded.

Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 214 of 274 Category: Fission Product Barrier Degradation Subcategory: N/A Initiating Condition: Any loss or any potential loss of Containment EAL:

FU1.1 Unusual Event Any loss or any potential loss of Containment (Table F-i)

Mode Applicability:

1 - Power Operations, 2 - Startup, 3 - Hot Shutdown NEI 99-01 Basis:

None CGS Basis:

Fuel Clad, RCS and Containment (PC) comprise the fission product barriers. Table F-1 (Attachment 2) lists the fission product barrier thresholds, bases and references.

Fuel Clad and RCS barriers are weighted more heavily than the Containment barrier. Unlike the Fuel Clad and RCS barriers, the loss of either of which results in an Alert (EAL FA1.1), loss of the Containment barrier in and of itself does not result in the relocation of radioactive materials or the potential for degradation of core cooling capability. However, loss or potential loss of the Containment barrier in combination with the loss or potential loss of either the Fuel Clad or RCS barrier results in declaration of a Site Area Emergency under EAL FS1.1.

CGS Basis Reference(s):

None Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 215 of 274 Category: Fission Product Barrier Degradation Subcategory: N/A Initiating Condition: Any loss or any potential loss of either Fuel Clad or RCS EAL:

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

Mode Applicability:

1 - Power Operations, 2 - Startup, 3 - Hot Shutdown NEI 99-01 Basis:

None CGS Basis:

Fuel Clad, RCS and Containment (PC) comprise the fission product barriers. Table F-1 (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 FS1.

CGS Basis Reference(s):

None Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 216 of 274 Category: Fission Product Barrier Degradation Subcategory: N/A Initiating Condition: Loss or potential loss of any two barriers EAL:

FS1.1 Site Area Emergency Loss or potential loss of any two barriers (Table F-i)

Mode Applicability:

1 - Power Operations, 2 - Startup, 3 - Hot Shutdown NEI 99-01 Basis:

None CGS Basis:

Fuel Clad, RCS and Containment (PC) comprise the fission product barriers. Table F-1 (Attachment 7.2) 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 Director would have greater assurance that escalation to a General Emergency is less imminent.

CGS Basis Reference(s):

None Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 217 of 274 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:

1 - Power Operations, 2 - Startup, 3 - Hot Shutdown NEI 99-01 Basis:

None CGS Basis:

Fuel Clad, RCS and Containment (PC) comprise the fission product barriers. Table F-1 (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 CGS Basis Reference(s):

None Attachment 7.1, EAL Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 218 of 274 FISSION PRODUCT BARRIER LOSS/POTENTIAL LOSS MATRIX AND BASES Introduction Table F-1 lists the threshold conditions that define the Loss and Potential Loss of the three fission product barriers (Fuel Clad, Reactor Coolant System, and Containment). The table is structured so that each of the three barriers occupies adjacent columns. Each fission product barrier column is further divided into two columns; one for Loss thresholds and one for Potential Loss thresholds.

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

A. RPV Level B. PC Pressure / Temperature C. Isolation D. Rad E. Judgment Each category occupies a row in Table F-1 thus forming a matrix defined by the categories. The intersection of each row with each Loss/Potential Loss column forms a cell in which one or more fission product barrier thresholds appear. If NEI 99-01 does not define a threshold for a barrier Loss/Potential Loss, the word "None" is entered in the cell.

Thresholds are assigned sequential numbers within each 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.1 ," the third Containment barrier Potential Loss would be assigned "PC P-Loss B.3," etc.

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

Subdivision of Table F-1 by category facilitates association of plant conditions to the applicable fission product barrier Loss and Potential Loss thresholds. This structure promotes a systematic approach to assessing the classification status of the fission product barriers When equipped with knowledge of plant conditions related to the fission product barriers, the EAL-user first scans down the category column of Table F-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 Attachment 7.2, Fission Product Barrier Loss/Potential Loss Matrix and Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 219 of 274 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 primary containment radiation is sufficiently high, a Loss of the Fuel Clad and RCS barriers and a Potential Loss of the Containment barrier can occur. Barrier Losses and Potential Losses are then applied to the algorithms given in EALs FG1.1, FS1.1, FA1.1 and FU1.1 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 to category Loss followed by category Potential Loss beginning with Category A, then B,,

E.

Attachment 7.2, Fission Product Barrier Loss/Potential Loss Matrix and Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 220 of 274 Table F-1 %FissionProduct Barrier Matrix Fuel Clad Barrier Reactor Coolant System Barrier Containment Barrier Loss Potential Loss Loss Potential Loss Loss Potential Loss A. RPV Level 1. PC Flooding required (SAG entry) 1. RPV level cannot be 1. RPV level cannot be restored and 1. PC Flooding required (SAG entry) due to EITHER: restored and maintained maintained above -161 in. or 01111 due to EITHER:

RPV level cannot be restored and above -161 in. or cannot be be determined RPV level cannot be restored and maintained above -183 in. determined maintained above -183 in.

OR I None Nove OR RPV level cannot be determined RPV level cannot be determined and it is determined that and it is determined that substantial core damage is substantial core damage is occurring due to loss of core occurring due to loss ot core cooling cooling B. PC Pressure 2. PC pressure GT 1.68 psig due to 1. PC pressure rise followed by a rapid 2. PC pressure GT 45 psig and rising

/Temperature RCS leakage unexplained drop in PC pressure 3. Explosive mixture exists inside PC (H2 GE Nave None None 2. PC pressure response no=consistent with 6% end 02 GE 5%)

LOCA conditions  : 4. WW temperature end RPV pressure cannot be maintained below the HCTL (EOP Figure C)

C.Isolation 3. Release pathway exists outside 1. RCS leakage GT 30 gpm inside 3. Failure of containment isolation valves (LCS primary containment resulting from the drywell Table 1.6.1.3-1) in any one line to close isolation failure in an, ofthe following auto or manual initiation following (excluding normal 2. Unisolable primary system leakage AND process system tlorepathsfrom en outside primary containment en Direct downstream pathway outside primary unisolable system): indicated by exceeding EITHER: containment and to the environment exists

• Main steam line RB area temperature alarm

• RCIC steam line level (PPM 5.3.1 Table 23) 4. Intentional PC venting per EOPs to prevent Nve Non RWCU Feedwater OR RB area radiation alarm level failure Nave (PPM 5.3.1 Table 24) 5. Unisolable primary system leakage outside

4. Emergency RPV Depressurization r PC as indicated by exceeding EITHER:

is required RB area maximum safe operating temperature (PPM 5.3.1 Table 23)

OR RB area maximum safe operating radiation (PPM 5.3.1 Table 24)

D. Rad 2. Containment Radiation Monitor 5. Containment Radiation Monitor 5. Containment Radiation Monitor CMS-RIS-27E or CMS-RIS-27F CMS-RIS-27E or CMS-RIS-27F CMS-RIS-27E or CMS-RIS-27F reading reading GT 3,600 R/hr N reading GT 70 R/hr NvNeGT Noneone 14,000 R/thr

3. Primary coolant activity GT 300 pCVgm dose equivalent 1-i131 E. Judgment 4. Antvcondition in the opinion of the 2. AnUcondition in the opinion of 6. Ant condition in the opinion of the 3. Anv condition in the opinion of the 6. Any condition in the opinion of the 6. Any condition in the opinion of the Emergency Director that indicates the Emergency Director that Emergency Director that indicates Emergency Director that indicates Emergency Director that indicates loss of Emergency Director that indicates loss of the Fuel Clad barrier indicates potential loss of the loss of the RCS barrier potential loss of the RCS barrier the Containment barrier I potential loss of the Containment barrier Fuel Clad barder Attachment 7.2, Fission Product Barrier Loss/Potential Loss Matrix and Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 221 of 274 Barrier: Fuel Clad Category: A. RPV Level Degradation Threat: Loss Threshold:

1. PC Flooding required (SAG entry) due to EITHER:

RPV level cannot be restored and maintained above -183 in.

OR RPV level cannot be determined and it is determined that substantial core damage is occurring due to loss of core cooling NEI 99-01 Basis:

This site specific value corresponds to the level used in EOPs to indicate challenge of core cooling.

This is the minimum value to assure core cooling without further degradation of the clad.

CGS Basis:

EOP flowcharts (PPM 5.1.1, PPM 5.1.2, PPM 5.1.4 and PPM 5.1.6) specify the requirements for Primary Containment (PC) Flooding and entry to the SAGs when core cooling is severely challenged.

EOP flowchart symbols containing the phrase "PC FLD REQ'D" signal this requirement. These EOPs provide instructions to ensure adequate core cooling by maintaining RPV water level above prescribed limits or operating sufficient RPV injection sources when level cannot be determined. SAG entry is required when:

" RPV level cannot be restored and maintained above -183 in. (Ref. 1, 2, 3): The Minimum Steam Cooling RPV Water Level (MSCRWL) is the lowest RPV water level at which the covered portion of the reactor core will generate sufficient steam to preclude any clad temperature in the uncovered portion of the core from exceeding 1500°F (Ref. 4). Consistent with the EOP definition of "cannot be restored and maintained," the determination that RPV level cannot be restored and maintained above the MSCRWL may be made at, before, or after RPV level actually decreases to this point.

" RPV level cannot be determined and it is determined that substantial core damage is occurring due to loss of core cooling: When RPV level cannot be determined, EOPs require RPV flooding strategies. RPV level indication provides the primary means of knowing if adequate core cooling is being maintained. When all means of determining RPV level are unavailable, reliance on alternate means of assuring adequate core cooling must be attempted. The instructions in PPM 5.1.4 and PPM 5.1.6 specify these means, which include emergency depressurization of the RPV and injection into the RPV at a rate needed to flood to the elevation of the main steam lines or hold RPV pressure above the Minimum Steam Cooling Pressure (in ATWS events) (Ref. 5, 6). (Ref. 4)

Attachment 7.2, Fission Product Barrier Loss/Potential Loss Matrix and Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 222 of 274 This threshold is also a Potential Loss of the Containment barrier (PC P-Loss A.1). Since SAG entry occurs after core uncovery has occurred, a Loss of the RCS barrier exists (RCS Loss A.1). SAG entry, therefore, represents a Loss of two barriers and a Potential Loss of a third, which requires a General Emergency classification.

CGS Basis Reference(s):

1. PPM 5.1.1 RPV Control

2. PPM 5.1.2 RPV Control - ATWS

3. Calculation NE-02-03-06 Attachment 10 RPV Variables

4. PPM 5.0.10 Flowchart Training Manual

5. PPM 5.1.4 RPV Flooding

6. PPM 5.1.6 RPV Flooding - ATWS Attachment 7.2, Fission Product Barrier Loss/Potential Loss Matrix and Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 223 of 274 Barrier: Fuel Clad Category: A. RPV Level Degradation Threat: Potential Loss Threshold:

1. RPV level cannot be restored and maintained above -161 in. or cannot be determined NEI 99-01 Basis:

This threshold is the same as the RCS barrier Loss threshold A.1 and corresponds to the RPV water level at the top of the active fuel. Thus, this threshold indicates a Potential Loss of the Fuel Clad barrier and a Loss of RCS barrier that appropriately escalates the emergency classification level to a Site Area Emergency.

CGS Basis:

An RPV level instrument reading of -161 in. indicates RPV level is at the top of active fuel (Ref. 1, 2).

When RPV level is at or above the top of active fuel, the core is completely submerged. Core submergence is the most desirable means of core cooling. When RPV level is below the top of active fuel, the uncovered portion of the core must be cooled by less reliable means (i.e., steam cooling or spray cooling). If core uncovery is threatened, the EOPs specify alternate, more extreme, RPV level control measures in order to restore and maintain adequate core cooling. Since core uncovery begins if RPV level drops to -161 in., the level is indicative of a challenge to core cooling and the Fuel Clad barrier.

Consistent with the EOP definition of "cannot be restored and maintained," the determination that RPV level cannot be restored and maintained above the top of active fuel may be made at, before, or after RPV level actually decreases to this point.

When RPV level cannot be determined, EOPs require RPV flooding strategies. RPV water level indication provides the primary means of knowing if adequate core cooling is being maintained. When all means of determining RPV level are unavailable, the fuel clad barrier is threatened and reliance on alternate means of assuring adequate core cooling must be attempted. The instructions in PPM 5.1.4 and PPM 5.1.6 specify these means, which include emergency depressurization of the RPV and injection into the RPV at a rate needed to flood to the elevation of the main steam lines or hold RPV pressure above the Minimum Steam Cooling Pressure (in ATWS events) (Ref. 3, 4). If RPV level cannot be determined with respect to the top of active fuel, a potential loss of the fuel clad barrier exists.

Note that PPM 5.1.2 may require intentional uncovery of the core and control of RPV level between -

161 in. and the Minimum Steam Cooling RPV Water Level (MSCRWL) (Ref. 5). Under these conditions, a high-power ATWS event exists and requires at least a Site Area Emergency classification in accordance with the ATWS/Criticality EALs.

Attachment 7.2, Fission Product Barrier Loss/Potential Loss Matrix and Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 224 of 274 CGS Basis Reference(s):

1. Calculation NE-02-03-05 Attachment 3 Note 8

2. PPM 5.1.1 RPV Control

3. PPM 5.1.4 RPV Flooding

4. PPM 5.1.6 RPV Flooding - ATWS

5. PPM 5.1.2 RPV Control - ATWS Attachment 7.2, Fission Product Barrier Loss/Potential Loss Matrix and Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 225 of 274 Barrier: Fuel Clad Category: B. PC Pressure / Temperature Degradation Threat: Loss Threshold:

None Attachment 7.2, Fission Product Barrier Loss/Potential Loss Matrix and Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 226 of 274 Barrier: Fuel Clad Category: B. PC Pressure / Temperature Degradation Threat: Potential Loss Threshold:

None Attachment 7.2, Fission Product Barrier Loss/Potential Loss Matrix and Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 227 of 274 Barrier: Fuel Clad Category: C. Isolation Degradation Threat: Loss Threshold:

None Attachment 7.2, Fission Product Barrier Loss/Potential Loss Matrix and Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 228 of 274 Barrier: Fuel Clad Category: C. Isolation Degradation Threat: Potential Loss Threshold:

None Attachment 7.2, Fission Product Barrier Loss/Potential Loss Matrix and Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 229 of 274 Barrier: Fuel Clad Category: D. Rad Degradation Threat: Loss Threshold:

2. Containment Radiation Monitor CMS-RIS-27E or CMS-RIS-27F reading GT 3,600 R/hr NEI 99-01 Basis:

The 3,600 R/hr reading is a value which indicates the release of reactor coolant, with elevated activity indicative of fuel damage, into the drywell.

Reactor coolant concentrations of this magnitude are several times larger than the maximum concentrations (including iodine spiking) allowed within technical specifications and are therefore indicative of fuel damage.

This value is higher than that specified for RCS barrier Loss threshold D.5. Thus, this threshold indicates a loss of both Fuel Clad barrier and RCS barrier that appropriately escalates the emergency classification level to a Site Area Emergency.

There is no Potential Loss threshold associated with this item.

CGS Basis:

Four high range area radiation detectors (CMS-RE-27A, B, E and F) are installed to monitor the drywell. CMS-RE-27A and -27B are located in the bioshield wall at elevations 522' and 525', azimuth 600 and 2970, respectively. CMS-RE-27E and -27F are located inside containment at elevation 515',

azimuth 2900 and 51.50, respectively. The companion containment radiation monitors (CMS-RIS-27A, B, E and F) are located on RAD Boards 22 and 23 in the Main Control Room. (Ref. 1)

/

The threshold value was calculated assuming the instantaneous release and dispersal of the reactor coolant noble gas and iodine inventory associated with a concentration of 300 PCi/gm dose equivalent 1-131 (or approximately 5% clad failure) into the drywell atmosphere. Evaluation of detector location, geometry and anticipated response suggests CMS-RIS-27E or F will provide the desired response to a given radiation source in the drywell and are, therefore, identified as the preferred monitors for evaluating this Fuel Clad Loss threshold. (Ref. 2)

CGS Basis Reference(s):

1. TM-2117 TSG - Core Thermal Engineer, Attachment 4.2

2. Calculation NE-02-94-57 Attachment 7.2, Fission Product Barrier Loss/Potential Loss Matrix and Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 230 of 274 Barrier: Fuel Clad Category: D. Rad Degradation Threat: Loss Threshold:

3. Primary coolant activity GT 300 pCi/gm dose equivalent 1-131 NEI 99-01 Basis:

Assessment by the EAL Task Force indicates that 300 RCi/gm 1-131 equivalent coolant activity is well above that expected for iodine spikes and corresponds to less than 5% fuel clad damage. This amount of radioactivity indicates significant clad damage and thus the Fuel Clad Barrier is considered lost.

There is no Potential Loss threshold associated with this item.

CGS Basis:

None CGS Basis Reference(s):

None Attachment 7.2, Fission Product Barrier Loss/Potential Loss Matrix and Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 231 of 274 Barrier: Fuel Clad Category: D. Rad Degradation Threat: Potential Loss Threshold:

None Attachment 7.2, Fission Product Barrier Loss/Potential Loss Matrix and Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 232 of 274 Barrier: Fuel Clad Category: E. Judgment Degradation Threat: Loss Threshold:

4. Any condition in the opinion of the Emergency Director that indicates loss of the Fuel Clad barrier NEI 99-01 Basis:

This threshold addresses any other factors that are to be used by the Emergency Director in determining whether the Fuel Clad barrier is lost. In addition, the inability to monitor the barrier should also be incorporated in this threshold as a factor in Emergency Director judgment that the barrier may be considered lost.

CGS Basis:

The Emergency Director 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 ifthe degradation will likely occur within two hours based on a projection of current safety system performance. The term "imminent' refers to recognition of the inability to reach safety 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 sequences lead to degradation of all fission product barriers and likely entry to the EOPs. The Emergency Director should be mindful of the Loss of AC power (Station Blackout) and ATWS EALs to assure timely emergency classification declarations.

CGS Basis Reference(s):

None Attachment 7.2, Fission Product Barrier Loss/Potential Loss Matrix and Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 233 of 274 Barrier: Fuel Clad Category: E. Judgment Degradation Threat: Potential Loss Threshold:

NEI 99-01 Basis:

This threshold addresses any other factors that are to be used by the Emergency Director in determining whether the Fuel Clad barrier is potentially lost. In addition, the inability to monitor the barrier should also be incorporated in this threshold as a factor in Emergency Director judgment that the barrier may be considered potentially lost.

CGS Basis:

The Emergency Director 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 degradation exists if the degradation will likely occur within two hours based on a projection of current safety system performance. The term "imminent' refers to recognition of the inability to reach safety 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 seguences lead to degradation of all fission product barriers and likely entry to the EOPs. The Emergency Director should be mindful of the Loss of AC power (Station Blackout) and ATWS EALs to assure timely emergency classification declarations.

CGS Basis Reference(s):

None Attachment 7.2, Fission Product Barrier Loss/Potential Loss Matrix and Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 234 of 274 Barrier: Reactor Coolant System Category: A. RPV Level Degradation Threat: Loss Threshold:

1. RPV level cannot be restored and maintained above -161 in. or cannot be determined NEI 99-01 Basis:

The Loss threshold RPV water level of 161 in. corresponds to the level that is used in EOPs to indicate challenge of core cooling.

This threshold is the same as Fuel Clad Barrier Potential Loss threshold A.1 and corresponds to a challenge to core cooling. Thus, this threshold indicates a Loss of RCS barrier and Potential Loss of Fuel Clad barrier that appropriately escalates the emergency classification level to a Site Area Emergency.

There is no Potential Loss threshold associated with this item.

CGS Basis:

An RPV level instrument reading of -161 in. indicates RPV level is at the top of active fuel (Ref. 1).

The top of the active fuel is significantly lower than the normal operating RPV level control band. To reach this level, RPV inventory loss would have previously required isolation of the RCS and Containment (PC) barriers, and initiation of all ECCS. If RPV level cannot be maintained above the top of active fuel, ECCS and other sources of RPV injection have been ineffective or incapable of reversing the decreasing level trend. The cause of the loss of RPV inventory is therefore assumed to be a Loss of Coolant Accident (LOCA). By definition, a LOCA event is a Loss of the RCS barrier.

Consistent with the EOP definition of "cannot be restored and maintained," the determination that RPV level cannot be restored and maintained above the top of active fuel may be made at, before, or after RPV level actually decreases to this point.

When RPV level cannot be determined, EOPs require RPV flooding strategies. The RPV flooding instructions in PPM 5.1.4 and PPM 5.1.6 first specify emergency depressurization of the RPV (Ref. 2, 3), which is defined to be a Loss of the RCS barrier (RCS Loss C.4).

Note that PPM 5.1.2 may require intentional uncovery of the core and control of RPV level between -

161 in. and the Minimum Steam Cooling RPV Water Level (MSCRWL) (Ref. 4). Under these conditions, a high-power ATWS event exists and requires at least a Site Area Emergency classification in accordance with the System Malfunction - ATWS/Criticality EALs.

Attachment 7.2, Fission Product Barrier Loss/Potential Loss Matrix and Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 235 of 274 CGS Basis Reference(s):

1. Calculation NE-02-03-05 Attachment 3 Note 8

2. PPM 5.1.4 RPV Flooding

3. PPM 5.1.6 RPV Flooding - ATWS

4. PPM 5.1.2 RPV Control - ATWS Attachment 7.2, Fission Product Barrier Loss/Potential Loss Matrix and Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 236 of 274 Barrier: Reactor Coolant System Category: A. RPV Level Degradation Threat: Potential Loss Threshold:

None Attachment 7.2, Fission Product Barrier Loss/Potential Loss Matrix and Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 237 of 274 Barrier: Reactor Coolant System Category: B. PC Pressure / Temperature Degradation Threat: Loss Threshold:

2. PC pressure GT 1.68 psig due to RCS leakage NEI 99-01 Basis:

The primary containment pressure of 1.68 psig is based on the drywell high pressure set point which indicates a LOCA by automatically initiating the ECCS or equivalent makeup system.

There is no Potential Loss threshold associated with this item.

CGS Basis:

The drywell high pressure scram setpoint is an entry condition to the EOP flowcharts: PPM 5.1.1, RPV Control, and PPM 5.2.1, Primary Containment Control (Ref. 1, 2, 3). Normal primary containment (PC) pressure control functions such as operation of drywell cooling and venting through SGT are specified in PPM 5.2.1 in advance of less desirable but more effective functions such as operation of drywell or wetwell sprays.

In the CGS design basis, primary containment pressures above the drywell high pressure scram setpoint are assumed to be the result of a high-energy release into the containment for which normal pressure control systems are inadequate or incapable of reversing the increasing pressure trend.

Pressures of this magnitude, however, can be caused by non-LOCA events such as a loss of drywell cooling or inability to control primary containment vent/purge (Ref. 4).

The threshold phrase "...due to RCS leakage" focuses the barrier failure on the RCS instead of the non-LOCA malfunctions that may adversely affect primary containment pressure. Primary containment pressure greater than 1.68 psig with corollary indications (e.g., elevated drywell temperature, indications of loss of RCS inventory) should therefore be considered a Loss of the RCS barrier. Loss of drywell cooling that results in pressure greater than 1.68 psig should not be considered an RCS barrier loss.

CGS Basis Reference(s):

1. Technical Specifications Table 3.3.5.1-1

2. PPM 5.1.1 RPV Control

3. PPM 5.2.1 Primary Containment Control

4. FSAR Section 6.3 Attachment 7.2, Fission Product Barrier Loss/Potential Loss Matrix and Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 238 of 274 Barrier: Reactor Coolant System Category: B. PC Pressure/ Temperature Degradation Threat: Potential Loss Threshold:

None Attachment 7.2, Fission Product Barrier Loss/Potential Loss Matrix and Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 239 of 274 Barrier: Reactor Coolant System Category: C. Isolation Degradation Threat: Loss Threshold:

3. Release pathway exists outside primary containment resulting from isolation failure in any of the following (excluding normal process system flowpaths from an UNISOLABLE system):

• Main steam line

• RCIC steam line

• RWCU

• Feedwater NEI 99-01 Basis:

An unisolable MSL break is a breach of the RCS barrier. Thus, this threshold is included for consistency with the Alert emergency classification level.

Other large high-energy line breaks such as Feedwater, RWCU, or RCIC that are unisolable also represent a significant loss of the RCS barrier and should be considered as MSL breaks for purposes of classification.

CGS Basis:

The conditions of this threshold include required containment isolation failures allowing a flow path to the environment. A release pathway outside primary containment exists when flow is not prevented by downstream isolations. Emergency declaration under this threshold would not be required in the case of a failure of both isolation valves to close but no downstream flowpath exists. Similarly, if the emergency response requires the normal process flow of a system outside primary containment (e.g.,

EOP requirement to bypass MSIV low RPV water level interlocks and maintain the main condenser as a heat sink using main turbine bypass valves), the threshold is not met. The combination of these threshold conditions represent the loss of both the RCS and Primary Containment (see PC Loss C.3) barriers and justifies declaration of a Site Area Emergency (i.e., Loss or Potential Loss of any two barriers). (Ref. 1-4)

Even though RWCU and Feedwater systems do not contain steam, they are included in the list because an unisolable break could result in the high-pressure discharge of fluid that is flashed to steam from relatively large volume systems directly connected to the RCS.

Attachment 7.2, Fission Product Barrier Loss/Potential Loss Matrix and Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 240 of 274 CGS Basis Reference(s):

1. FSAR Section 5.4.5

2. FSAR Section 5.4.6

3. FSAR Section 5.4.8

4. FSAR Section 10.3 Attachment 7.2, Fission Product Barrier Loss/Potential Loss Matrix and Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 241 of 274 Barrier: Reactor Coolant System Category: C. Isolation Degradation Threat: Loss Threshold:

4. Emergency RPV Depressurization is required NEI 99-01 Basis:

Plant symptoms requiring Emergency RPV Depressurization per the EOP flowcharts are indicative of a loss of the RCS barrier. If Emergency RPV depressurization is required, the plant operators are directed to open safety relief valves (SRVs) and keep them open. Even though the RCS is being vented into the suppression pool, a loss of the RCS should be considered to exist due to the diminished effectiveness of the RCS pressure barrier to a release of fission products beyond its boundary.

CGS Basis:

Emergency RPV Depressurization is specified in the EOP flowcharts when symbols containing the phrase "EMERG DEPRESS REQ'D" are reached. The requirements for emergency RPV depressurization appear in the following EOP flowpaths:

• PPM 5.1.1, RPV Control: RPV Level and Steam Cooling flowpaths

• PPM 5.1.2, RPV Control - ATWS: RPV Level flowpath

• PPM 5.2.1, Primary Containment Control:

o Wetwell temperature o Drywell temperature o Primary containment pressure o Wetwell water level (two places)

• PPM 5.3.1, Secondary Containment Control

• PPM 5.4.1, Radioactivity Release Control Emergency RPV depressurization is also performed upon entry to the RPV Flooding flowcharts in PPM 5.1.4 and PPM 5.1.6.

Attachment 7.2, Fission Product Barrier Loss/Potential Loss Matrix and Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 242 of 274 CGS Basis Reference(s):

1. PPM 5.1.1 RPV Control

2. PPM 5.1.2 RPV Control - ATWS

3. PPM 5.1.4 RPV Flooding

4. PPM 5.1.6 RPV Flooding - ATWS

5. PPM 5.2.1 Primary Containment Control

6. PPM 5.3.1 Secondary Containment Control

7. PPM 5.4.1 Radioactivity Release Control Attachment 7.2, Fission Product Barrier Loss/Potential Loss Matrix and Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 243 of 274 Barrier: Reactor Coolant System Category: C. Isolation Degradation Threat: Potential Loss Threshold:

1. RCS leakage GE 30 gpm inside the drywell NEI 99-01 Basis:

This threshold is based on leakage is set at a level indicative of a small breach of the RCS but which is well within the makeup capability of normal and emergency high pressure systems. Core uncovery is not a significant concern for a 30 gpm leak, however, break propagation leading to significantly larger loss of inventory is possible.

If primary system leak rate information is unavailable, other indicators of RCS leakage should be used.

CGS Basis:

RCS leakage inside the drywell is normally determined by monitoring drywell equipment and floor drain sump pumpout rates. This method of monitoring leakage may be isolated as part of the drywell isolation, and thus may be unavailable. If primary system leak rate information is unavailable, other indicators of RCS leakage should be used (Ref. 1-7). Inventory loss events, such as a stuck open SRV, should not be considered when referring to "RCS leakage" because they are not indications of a break, which could propagate.

Total leakage is considered to be the total of both identified and unidentified leakage as measured on EDR-FRS-623 pen 1 (unidentified - Floor Drain Sump Fill Rate) and pen 2 (identified - Equipment Drain Sump Fill Rate). The maximum measurable identified leak rate (pen 2) in the Control Room at Columbia Generating Station is 30 gpm. (Ref. 8, 9)

For classification under this threshold, RCS leakage includes a broken SRV tailpipe that is discharging into the drywell or wetwell airspace. Once the SRV is closed, however, this RCS leakage path is considered isolated.

Attachment 7.2, Fission Product Barrier Loss/Potential Loss Matrix and Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 244 of 274 CGS Basis Reference(s):

1. Technical Specifications 3.4.7

2. FSAR Section 5.2.5

3. FSAR Section 7.6.1

4. ABN-LEAKAGE Reactor Coolant Leakage

5. SOP-EDR-OPS Equipment Drain System Operation

6. SOP-FDR-OPS Floor Drain System Operation

7. PPM 10.27.35 Leakage Surveillance and Prevention Program

8. PPM 4.601.A3 Annunciator Response, P601 Annunciator A3

9. Instrument Master Datasheet EDR-FRS-623 Attachment 7.2, Fission Product Barrier Loss/Potential Loss Matrix and Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 245 of 274 Barrier: Reactor Coolant System Category: C. Isolation Degradation Threat: Potential Loss Threshold:

2. UNISOLABLE primary system leakage outside primary containment as indicated by exceeding EITHER:

RB area temperature alarm level (PPM 5.3.1 Table 23)

OR RB area radiation alarm level (PPM 5.3.1 Table 24)

NEI 99-01 Basis:

Potential loss of RCS based on primary system leakage outside the primary containment is determined from PPM 5.3.1 temperature or area radiation Max Normal setpoints in the areas of the main steam line tunnel, main turbine generator, RCIC, etc., which indicate a direct path from the RCS to areas outside primary containment.

The indicators reaching the threshold barriers and confirmed to be caused by RCS leakage warrant an Alert classification. An unisolable leak which is indicated by a high alarm setpoint escalates to a Site Area Emergency when combined with Containment Barrier Loss threshold C.5 (after a containment isolation) and a General Emergency when the Fuel Clad Barrier criteria is also exceeded.

CGS Basis:

The presence of elevated general area temperatures or radiation levels in the Reactor Building (RB) may be indicative of unisolable primary system leakage outside the primary containment. Per EOP flowcharts, the RB alarm levels define the Maximum Normal Operating values because they signify the onset of abnormal system operation. When parameters reach this level, equipment failure or misoperation may be occurring. Elevated parameters may also adversely affect the ability to gain access to or operate equipment within the affected area. The locations into which the primary system discharge is of concern correspond to the areas addressed in PPM 5.3.1 Tables 23 and 24 (Ref. 1).

Area temperature alarms are provided by the leak detection and reactor building recirculation air (RRA) systems (Ref. 2)

The ARM alarm setpoints listed in Table 24 vary due to plant operating mode and Health Physics radiation surveys. A program is established to maintain the current setpoint values in PPM 4.602.A5 for annunciator window 3-1; thus, reference is made to the annunciator response procedure in Table 24. (Ref. 2)

Attachment 7.2, Fission Product Barrier Loss/Potential Loss Matrix and Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 246 of 274 In general, multiple indications should be used to determine if a primary system is discharging outside Primary Containment. For example, a high area radiation condition does not necessarily indicate that a primary system is discharging into the secondary containment since this may be caused by radiation shine from nearby steam lines or the movement of radioactive materials. Conversely, a high area radiation condition in conjunction with other indications (e.g. room flooding, high area temperatures, reports of steam in the secondary containment, an unexpected rise in feedwater flowrate, or unexpected main turbine control valve closure) may indicate that a primary system is discharging into the secondary containment.

Attachment 7.2, Fission Product Barrier Loss/Potential Loss Matrix and Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 247 of 274 PPM 5.3.1 Table 23 - RB Area Temps 23 1RB Area Temps MxmmSf Maximum Safe Area Instrument Alarm Operating Value (OF) (OF)

RWCU Pump 1A Room LD-TE-3A (B) 160 320 RWCU Pump IB Room LD-TE-3C (D) 160 320 Above RWCU Pump Rooms 'R57 LD-TE-24G (H) 160 320 RCIC Pump Room LD-TE-4A (B) 160 200 RWCU Pipe Routing Area " LD-TE-24C (D) 160 180 TIP Mezzanine (FT,3E LD-TE-24J (K) 160 212 RWCU Pipe Routing Area " LD-TE-24E (F) 160 340 RWCU Pipe Routing Area R*N LD-TE-24A (B) 160 165 Steam Tunnel LD-TE-31A (B,C.D) 164 320 RHR A Pump Room LD-TE-18C (D) 140 210 RHR B HX Room LO-TE-18L (M) 130 210 LD-TE-18J (K) 140 212 RHR A HX Room LD-TE-18E (F) 130 212 LD-TE-18G (H) 150 212 RHR B Pump Room LD-TE-18A (B) 140 212 Attachment 7.2, Fission Product Barrier Loss/Potential Loss Matrix and Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 248 of 274 PPM 5.3.1 Table 24 - RB Area Radiation 241 RB Area Radiation MxmmSf Maximum Safe Area Instrument Alarm Operating Value (mRilr) (mR/hr) (Rihr)

East CRD Area ARM-RIS-4 N/A West CRD Area ARM-RIS-5 104 N/A H2 Recomber Area ARM-RIS-6 104 N/A TIP Drive Area ARM-RISI7 104 N/A SGT Filter Area ARM-RIS-8 10- N4A RHR Pump A Room ARM-RIS-9 104 N/A RHR Pump B Room ARM-RIS- 10£1 104 N/A RHR Pump C Room ARM-RIS-II I 104 N/A RCIC Pump Room ARM-RIS-12 4 N/A HPCS Pump Room ARM-RIS-1 3 104 N/A CRD Pump Room ARM-RIS-23 104 N/A 471 Elev - West Area ARM-RIS-24 104 N/A 471 Elev - NE ARM-RIS-32 N/A 10 501 Elev - NW ARM-RIS-33 N/A 10 CGS Basis Reference(s):

1. PPM 5.3.1 Secondary Containment Control

2. PPM 5.0.10 Flowchart Training Manual Attachment 7.2, Fission Product Barrier Loss/Potential Loss Matrix and Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 249 of 274 Barrier: Reactor Coolant System Category: D. Rad Degradation Threat: Loss Threshold:

5. Containment Radiation Monitor CMS-RIS-27E or CMS-RIS-27F reading GT 70 R/hr NEI 99-01 Basis:

The 70 R/hr reading is a value which indicates the release of reactor coolant to the primary containment.

This reading will be less than that specified for Fuel Clad barrier Loss threshold D.2. Thus, this threshold would be indicative of a RCS leak only. If the radiation monitor reading increased to that value specified by Fuel Clad Barrier threshold, fuel damage would also be indicated.

There is no Potential Loss threshold associated with this item.

CGS Basis:

Four high range area radiation detectors (CMS-RE-27A, B, E and F) are installed in the drywell. CMS-RE-27A and -27B are located in the bioshield wall at elevations 522' and 525', azimuth 600 and 2970, respectively. CMS-RE-27E and -27F are located inside containment at elevation 515', azimuth 2900 and 51.50, respectively. The companion containment radiation monitors (CMS-RIS-27A, B, E and F) are located on RAD Boards 22 and 23 in the Main Control Room. (Ref. 1)

The threshold value was calculated assuming 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 drywell atmosphere. Evaluation of detector location, geometry and anticipated response suggests CMS-RIS-27E or F will provide the desired response to a given radiation source in the drywell and are, therefore, identified as the preferred monitors for evaluating this RCS Loss threshold. (Ref. 2)

CGS Basis Reference(s):

1. TM-2117 TSG - Core Thermal Engineer, Attachment 4.2

2. Calculation NE-02-94-57 Attachment 7.2, Fission Product Barrier Loss/Potential Loss Matrix and Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 250 of 274 Barrier: Reactor Coolant System Category: D. Rad Degradation Threat: Potential Loss Threshold:

None 1

Attachment 7.2, Fission Product Barrier Loss/Potential Loss Matrix and Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 251 of 274 Barrier: Reactor Coolant System Category: E. Judgment Degradation Threat: Loss Threshold:

6. Any condition in the opinion of the Emergency Director that indicates loss of the RCS barrier NEI 99-01 Basis:

This threshold addresses any other factors that are to be used by the Emergency Director in determining whether the RCS barrier is lost. In addition, the inability to monitor the barrier should also be incorporated in this threshold as a factor in Emergency Director judgment that the barrier may be considered lost.

CGS Basis:

The Emergency Director judgment threshold addresses any other factors relevant to determining if the RCS barrier is lost. Such a determination should include imminent barrier degradation, barrier monitoring capability and dominant accident sequences.

" Imminent barrier degradation exists ifthe degradation will likely occur within two hours based on a projection of current safety system performance. The term "imminent" refers to the recognition of the inability to reach safety 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 sequences lead to degradation of all fission product barriers and likely entry to the EOPs. The Emergency Director should be mindful of the Loss of AC power (Station Blackout) and ATWS EALs to assure timely emergency classification declarations.

CGS Basis Reference(s):

None Attachment 7.2, Fission Product Barrier Loss/Potential Loss Matrix and Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 252 of 274 Barrier: Reactor Coolant System Category: E. Judgment Degradation Threat: Potential Loss Threshold:

NEI 99-01 Basis:

This threshold addresses any other factors that are to be used by the Emergency Director in determining whether the RCS barrier is I potentially lost. In addition, the inability to monitor the barrier should also be incorporated in this threshold as a factor in Emergency Director judgment that the barrier may be considered potentially lost.

CGS Basis:

The Emergency Director 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 two hours based on a projection of current safety system performance. The term "imminent' refers to the inability to reach final safety acceptance criteria before completing 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 sequences lead to degradation of all fission product barriers and likely entry to the EOPs. The Emergency Director should be mindful of the Loss of AC power (Station Blackout) and ATWS EALs to assure timely emergency classification declarations.

CGS Basis Reference(s):

None Attachment 7.2, Fission Product Barrier Loss/Potential Loss Matrix and Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 253 of 274 Barrier: Containment Category: A. RPV Level Degradation Threat: Loss Threshold:

None Attachment 7.2, Fission Product Barrier Loss/Potential Loss Matrix and Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 254 of 274 Barrier: Containment Category: A. RPV Level Degradation Threat: Potential Loss Threshold:

1. PC Flooding required (SAG entry) due to EITHER:

RPV level cannot be restored and maintained above -183 in.

OR RPV level cannot be determined and it is determined that substantial core damage is occurring due to loss of core cooling NEI 99-01 Basis:

There is no Loss threshold associated with this item.

The potential loss requirement for Primary Containment Flooding indicates adequate core cooling cannot be established and maintained and that core melt is possible. Entry into Primary Containment Flooding procedures is a logical escalation in response to the inability to maintain adequate core cooling.

The condition in this potential loss threshold represents a potential core melt sequence which, if not corrected, could lead to vessel failure and increased potential for containment failure. In conjunction with Reactor Vessel water level "Loss" thresholds in the Fuel Clad and RCS barrier columns, this threshold will result in the declaration of a General Emergency -- loss of two barriers and the potential loss of a third.

CGS Basis:

EOP flowcharts (PPM 5.1.1, PPM 5.1.2, PPM 5.1.4 and PPM 5.1.6) specify the requirements for Primary Containment Flooding and entry to the SAGs when core cooling is severely challenged. EOP flowchart symbols containing the phrase "PC FLD REQ'D" signal this requirement. These EOPs provide instructions to ensure adequate core cooling by maintaining RPV water level above prescribed limits or operating sufficient RPV injection sources when level cannot be determined. SAG entry is required when:

RPV water level cannot be restored and maintained above -183 in. (Ref. 1, 2, 3): The Minimum Steam Cooling RPV Water Level (MSCRWL) is the lowest RPV water level at which the covered portion of the reactor core will generate sufficient steam to preclude any clad temperature in the uncovered portion of the core from exceeding 1500°F (Ref. 4). Consistent with the EOP definition of "cannot be restored and maintained," the determination that RPV level cannot be restored and maintained above the MSCRWL may be made at, before, or after RPV level actually decreases to this point.

Attachment 7.2, Fission Product Barrier Loss/Potential Loss Matrix and Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 255 of 274 RPV water level cannot be determined and it is determined that substantial core damage is occurring due to loss of core cooling: When RPV level cannot be determined, EOPs require RPV flooding strategies. RPV level indication provides the primary means of knowing if adequate core cooling is being maintained. When all means of determining RPV level are unavailable, reliance on alternate means of assuring adequate core cooling must be attempted. The instructions in PPM 5.1.4 and PPM 5.1.6 specify these means, which include emergency depressurization of the RPV and injection into the RPV at a rate needed to flood to the elevation of the main steam lines or hold RPV pressure above the Minimum Steam Cooling Pressure (in ATWS events) (Ref. 5, 6). Examples of possible methods of determining if substantial core damage has occurred include: ERO evaluation per TM-2117 Attachment 4 (core damage assessment), Elevated drywell radiation levels, H2 detected in the drywell or wetwell, elevated radiation levels on the SJAE or main steam line radiation monitors, and reactor coolant sampling. Existing damage and fuel failures resulting from power oscillations or reactivity transients, by themselves, are not indicative of a loss of adequate core cooling and do not constitute primary containment flooding requirements. (Ref. 4)

This threshold is also a Loss of the Fuel Clad barrier (FC Loss A.1). Since Primary Containment Flooding occurs after core uncovery has occurred a Loss of the RCS barrier exists (RCS Loss A.1).

Primary Containment Flooding, therefore, represents a Loss of two barriers and a Potential Loss of a third, which requires a General Emergency classification.

CGS Basis Reference(s):

1. PPM 5.1.1 RPV Control

2. PPM 5.1.2 RPV Control - ATWS

3. Calculation NE-02-03-06 Attachment 10 RPV Variables

4. PPM 5.0.10 Flowchart Training Manual

5. PPM 5.1.4 RPV Flooding

6. PPM 5.1.6 RPV Flooding - ATWS

7. TM-2117 Attachment 4 Core Damage Assessment Attachment 7.2, Fission Product Barrier Loss/Potential Loss Matrix and Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 256 of 274 Barrier: Containment Category: B. PC Pressure / Temperature Degradation Threat: Loss Threshold:

1. PC pressure rise followed by a rapid unexplained drop in PC pressure NEI 99-01 Basis:

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

This indicator relies on operator recognition of an unexpected response for the condition and therefore does not have a specific value associated with it. The unexpected response is important because it is the indicator for a containment bypass condition.

CGS Basis:

None CGS Basis Reference(s):

None Attachment 7.2, Fission Product Barrier Loss/Potential Loss Matrix and Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 257 of 274 Barrier: Containment Category: B. PC Pressure / Temperature Degradation Threat: Loss Threshold:

2. PC pressure response not consistent with LOCA conditions NEI 99-01 Basis:

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

This indicator relies on operator recognition of an unexpected response for the condition and therefore does not have a specific value associated with it. The unexpected response is important because it is the indicator for a containment bypass condition.

CGS Basis:

FSAR Section 6.2 provides a summary of primary containment pressure response for several postulated accident conditions resulting in the release of RCS inventory to the containment. These accidents include:

• Rupture of a recirculation line

• Rupture of a main steam line

• Intermediate size liquid line rupture

• Small size steam line rupture The containment response to the main steam line, intermediate liquid line and small size steam line breaks were bounded by the recirculation line break. (Ref. 1)

FSAR Figure 6.2-3 illustrates the containment pressure response due to a recirculation line break (Ref. 2). The maximum calculated drywell pressure is 37.4 psig (52.1 psia) and is well below the design allowable pressure of 45 psig. (Ref. 3, 4)

Due to conservatisms in LOCA analyses, actual pressure response is expected to be less than the analyzed response. For example, blowdown mass flowrate may be only 60-80% of the analyzed rate, initial containment pressure may be less than 2 psig, etc.

Attachment 7.2, Fission Product Barrier Loss/Potential Loss Matrix and Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 258 of 274 LOCA conditions are manifested on Control Room instrumentation by DW pressure rising with WW pressure following in a manner similar to that shown in FSAR Figure 6.2-3. A broken SRV tailpipe could infer this threshold if WW pressure is higher than DW pressure; however, if the SRV is closed, the condition would no longer exist.

CGS Basis Reference(s):

1. FSAR Section 6.2.1.1.3.3

2. FSAR Figure 6.2-3

3. FSAR Table 6.2-5

4. FSAR Table 6.2-1 Attachment 7.2, Fission Product Barrier Loss/Potential Loss Matrix and Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 259 of 274 FSAR Figure 6.2-3: Pressure Response for Recirculation Line Break (Initial Containment Pressure 2 psig) 75 50 3

97-a.

25 0

0 10 20 30 40 Time (Seconds)

Attachment 7.2, Fission Product Barrier Loss/Potential Loss Matrix and Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 260 of 274 Barrier: Containment Category: B. PC Pressure / Temperature Degradation Threat: Potential Loss Threshold:

2. PC pressure GT 45 psig and rising NEI 99-01 Basis:

The PC pressure of 45 psig is based on the primary containment design pressure.

CGS Basis:

If this threshold is exceeded, a challenge to the containment structure has occurred because assumptions used in the accident analysis are no longer valid and an unanalyzed condition exists (Ref. 1, 2). This constitutes a Potential Loss of the Containment barrier even if a containment breach has not occurred.

CGS Basis Reference(s):

1. FSAR Table 6.2-1

2. FSAR Section 6.2 Attachment 7.2, Fission Product Barrier Loss/Potential Loss Matrix and Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 261 of 274 Barrier: Containment Category: B. PC Pressure / Temperature Degradation Threat: Potential Loss Threshold:

3. Explosive mixture exists inside PC (H2 GE 6% and 02 GE 5%)

NEI 99-01 Basis:

BWRs specifically define the limits associated with explosive mixtures in terms of deflagration concentrations of hydrogen and oxygen.

CGS Basis:

Explosive (deflagration) mixtures in the primary containment are assumed to be elevated concentrations of hydrogen and oxygen. BWR industry evaluation of hydrogen generation for development of EOPs/SAGs indicates that any hydrogen concentration above minimum detectable is not to be expected within the short term. Post-LOCA hydrogen generation primarily caused by radiolysis is a slowly evolving, long-term condition. Hydrogen concentrations that rapidly develop are most likely caused by metal-water reaction. A metal-water reaction is indicative of an accident more severe than accidents considered in the plant design basis and would be indicative, therefore, of a potential threat to primary containment integrity. Hydrogen concentration of approximately 6% is considered the global deflagration concentration limit (Ref. 1).

Except for brief periods during plant startup and shutdown, oxygen concentration in the primary containment is maintained at insignificant levels by nitrogen inertion. The specified values for this Potential Loss threshold are the minimum global deflagration concentration limits (6% hydrogen and 5% oxygen, Ref. 2) and readily recognizable because 6% hydrogen is well above the EOP flowchart entry condition (Ref. 3). The minimum global deflagration hydrogen/oxygen concentrations (60/o/5%,

respectively) require intentional primary containment venting, which is defined to be a Loss of Containment (PC Loss C.4).

Atmosphere samples from a minimum of two locations inside the primary containment and one location in the suppression chamber are sequentially monitored for hydrogen and oxygen percentage levels by each of two redundant analyzer systems. The analyzers are single range (0 to 30%

hydrogen and 0 to 30% oxygen). Two redundant (divisional) recorders are provided in the Main Control Room CMS-02/H2R-1 (H13-P827) and CMS-02/H2R-2 (H13-P81 1). Hydrogen and oxygen concentrations can also be displayed on the plant computers (TDAS points X410 & X471 and 02 on TDAS points X395 and X456). The primary containment high hydrogen alarm is set at 3.56%. The minimum detectable primary containment hydrogen concentration is 0.6%. (Ref. 4, 5, 6, 7)

Attachment 7.2, Fission Product Barrier Loss/Potential Loss Matrix and Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 262 of 274 When the monitoring systems for hydrogen or oxygen become unavailable, the concentration of these gases must be determined by manual sample and analysis. This activity is a function of Chemistry using PPM 12.17.3 to sample and analyze the primary containment atmosphere. (Ref. 3)

CGS Basis Reference(s):

1. BWROG EPG/SAG Revision 2, Sections PC/G

2. PPM 5.7.1 RPV and Primary Containment Flooding SAG, Table 19

3. PPM 5.2.1 Primary Containment Control

4. FSAR Section 7.5.1.5.4

5. PPM 5.0.10 Flowchart Training Manual

6. PPM 4.814.J1 814.J1 Annunciator Panel Alarms, 2-2

7. PPM 4.814.J2 814.J2 Annunciator Panel Alarms, 2-2 Attachment 7.2, Fission Product Barrier Loss/Potential Loss Matrix and Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 263 of 274 Barrier: Containment Category: B. PC Pressure / Temperature Degradation Threat: Potential Loss Threshold:

4. WW temperature and RPV pressure cannot be maintained below the HCTL (EOP Figure C)

NEI 99-01 Basis:

The Heat Capacity Temperature Limit (HCTL) is the highest wetwell temperature from which Emergency RPV Depressurization will not raise:

" Wetwell temperature above the maximum temperature capability of the wetwell and equipment within the wetwell which may be required to operate when the RPV is pressurized (275 0 F),

OR

• Wetwell pressure above the Primary Containment Pressure Limit (PCPL),

while the rate of energy transfer from the RPV to the containment is greater than the capacity of the containment vent.

The HCTL is a function of RPV pressure and wetwell level. It is utilized to preclude failure of the containment and equipment in the containment necessary for the safe shutdown of the plant and therefore, the inability to maintain plant parameters below the limit constitutes a potential loss of containment.

CGS Basis:

The HCTL is given in EOP flowchart Figure C. This threshold is met when EOP flowchart PPM 5.2.1 Step WT-5 is reached (Ref. 1).

CGS Basis Reference(s):

1. PPM 5.2.1 Primary Containment Control Attachment 7.2, Fission Product Barrier Loss/Potential Loss Matrix and Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 264 of 274 Barrier: Containment Category: C. Isolation Degradation Threat: Loss Threshold:

3. Failure of containment isolation valves (LCS Table 1.6.1.3-1) in any one line to close following auto or manual initiation AND Direct downstream pathway outside primary containment and to the environment exists NEI 99-01 Basis:

These thresholds address incomplete containment isolation that allows direct release to the environment.

The use of the modifier "direct" in defining the release path discriminates against release paths through interfacing liquid systems. The existence of an in-line charcoal filter does not make a release path indirect since the filter is not effective at removing fission product noble gases. Typical filters have an efficiency of 95-99% removal of iodine. Given the magnitude of the core inventory of iodine, significant releases could still occur. In addition, since the fission product release would be driven by boiling in the reactor vessel, the high humidity in the release stream can be expected to render the filters ineffective in a short period.

CGS Basis:

This threshold addresses failure of open isolation devices which should close upon receipt of a manual or automatic containment isolation signal resulting in a significant radiological release pathway directly to the environment. The concern is the unisolable open pathway to the environment.

A failure of the ability to isolate any one line indicates a breach of primary containment integrity.

Technical Specifications LCS Table 1.6.1.3-1 provides a list of applicable isolation devices (Ref. 1).

As stated above, the adjective "Direct" modifies "pathway" to discriminate against release paths through interfacing liquid systems. Leakage into a closed system is to be considered only if the closed system is breached and thereby creates a significant pathway to the environment. Examples include unisolable Main steam line or RCIC steam line breaks, unisolable RWCU system breaks, and unisloable containment atmosphere vent paths. If the main condenser is available with an unisolable main steam line, there may be releases through the steam jet air ejectors and gland seal exhausters.

These pathways are monitored, however, and do not meet the intent of a nonisolable release path to the environment, These minor releases are assessed using the Category R, Abnormal Rad Release /

Rad Effluent, EALs.

The existence of an in-line charcoal filter (SGT) does not make a release path indirect since the filter is not effective at removing fission noble gases. Typical filters have an efficiency of 95-99% removal Attachment 7.2, Fission Product Barrier Loss/Potential Loss Matrix and Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 265 of 274 of iodine. Given the magnitude of the core inventory of iodine, significant releases could still occur. In addition, since the fission product release would be driven by boiling in the reactor vessel, the high humidity in the release stream can be expected to render the filters ineffective in a short period.

The threshold is met if the breach is not isolable from the Control Room or an attempt for isolation from the Control Room has been made and was unsuccessful. An attempt for isolation from the Control Room should be made prior to the emergency classification. If operator actions from the Control Room are successful, this threshold is not applicable. Credit is not given for operator actions taken in-plant (outside the Control Room) to isolate the breach.

EOP flowcharts (PPM 5.2.1, Primary Containment Control) may specify primary containment venting and intentional bypassing of the containment isolation valve logic even if offsite radioactivity release rate limits are exceeded (Ref. 2). Under these conditions with a valid containment isolation signal, the Containment barrier should be considered lost.

CGS Basis Reference(s):

1. Technical Specifications LCS Table 1.6.1.3-1

2. PPM 5.2.1 Primary Containment Control Attachment 7.2, Fission Product Barrier Loss/Potential Loss Matrix and Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 266 of 274 Barrier: Containment Category: C. Isolation Degradation Threat: Loss Threshold:

4. Intentional PC venting per EOPs to prevent failure NEI 99-01 Basis:

These thresholds address incomplete containment isolation that allows direct release to the environment.

EOPs may direct containment isolation valve logic(s) to be intentionally bypassed, regardless of radioactivity release rates. Under these conditions with a valid containment isolation signal, the containment should also be considered lost if containment venting is actually performed.

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

CGS Basis:

EOP flowcharts (PPM 5.2.1, Primary Containment Control, Step P-14) may specify primary containment venting and intentional bypassing of the containment isolation valve logic, even if offsite radioactivity release rate limits are exceeded (Ref. 1). The threshold is met when the operator begins venting the primary containment in accordance with EOP Support Procedures (PPM 5.5.14 or PPM 5.5.15), not when actions are taken to bypass interlocks prior to opening the vent valves. Purge and vent actions specified in PPM 5.2.1 Step P-1 to control primary containment pressure below the drywell high pressure scram setpoint or Step H-2 to lower hydrogen concentration does not meet this threshold because such action is only permitted if offsite radioactivity release rates will remain below the ODCM RFO limits.

CGS Basis Reference(s):

1. PPM 5.2.1 Primary Containment Control

2. PPM 5.5.14 Emergency Wetwell Venting

3. PPM 5.5.15 Emergency Drywell Venting Attachment 7.2, Fission Product Barrier Loss/Potential Loss Matrix and Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 267 of 274 Barrier: Containment Category: C. Isolation Degradation Threat: Loss Threshold:

2. UNISOLABLE primary system leakage outside PC as indicated by exceeding EITHER:

RB area maximum safe operating temperature (PPM 5.3.1 Table 23)

OR RB area maximum safe operating radiation (PPM 5.3.1 Table 24)

NEI 99-01 Basis:

This threshold addresses incomplete containment isolation that allows direct release to the environment.

The presence of area radiation or temperature Max Safe Operating setpoints indicating unisolable primary system leakage outside the primary containment are addressed after a containment isolation.

The indicators should be confirmed to be caused by RCS leakage.

There is no Potential Loss threshold associated with this item.

CGS Basis:

The presence of elevated general area temperatures or radiation levels in the Reactor Building (RB) may be indicative of unisolable primary system leakage outside the primary containment. The maximum safe operating values define this Containment barrier threshold because they are indicative of problems in the secondary containment that are spreading and pose a threat to achieving a safe plant shutdown. This threshold addresses problematic discharges outside primary containment that may not originate from a high-energy line break. The locations into which the primary system discharge is of concern correspond to the areas addressed in PPM 5.3.1 Tables 23 and 24 (Ref. 1).

RB maximum safe operating temperatures are conservatively defined by the qualification temperature of safety related equipment in the area. The equipment qualification program has proven that safety related equipment will perform satisfactorily to at least this temperature. In an area with multiple components and different qualification temperatures, the maximum safe operating temperature assigned to that area is generally the lowest of the individual temperatures. (Ref. 2)

The maximum safe operating radiation value is defined to be 10,000 mR/hr in areas other than the refueling floor. This is the maximum indication on all but the high level instruments. This value is high enough to be indicative of substantial and immediate problems yet low enough to allow time for shutdown or isolation of a leak without exceeding the total integrated dose allowable for even the Attachment 7.2, Fission Product Barrier Loss/Potential Loss Matrix and Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 268 of 274 most sensitive safety related equipment. No area radiation levels are defined for the refueling floor because no primary systems are routed there. (Ref. 2)

In general, multiple indications should be used to determine if a primary system is discharging outside Primary Containment. For example, a high area radiation condition does not necessarily indicate that a primary system is discharging into the secondary containment since this may be caused by radiation shine from nearby steam lines or the movement of radioactive materials. Conversely, a high area radiation condition in conjunction with other indications (e.g. room flooding, high area temperatures, reports of steam in the secondary containment, an unexpected rise in feedwater flowrate, or unexpected main turbine control valve closure) may indicate that a primary system is discharging into the secondary containment.

Attachment 7.2, Fission Product Barrier Loss/Potential Loss Matrix and Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 269 of 274 PPM 5.3.1 Table 23 - RB Area Temps 23 RB Area Temps Maximum Safe Area Instrument Alarm Operating Value (OF) (OF)

RWCU Pump iA Room LD-TE-3A (B) 160 320 RWCU Pump I B Room LD-TE-3C (D) 160 320 Above RWCU Pump Rooms " LD-TE-24G (H) 160 320 RCIC Pump Room LD-TE-4A (B) 160 200 RWCU Pipe Routing Area " LD-TE-24C (D) 160 180 TIP Mezzanine , LD-TE-24J (K) 160 212 RWCU Pipe Routing Area "%<J LD-TE-24E (F) 160 340 RWCU Pipe Routing Area R'wN LD-TE-24A (B) 160 165 Steam Tunnel LD-TE-31 A (BC,D) 164 320 RHR A Pump Room LD-TE-18C (D) 140 210 RHR B [IX Room LD-TE-18L (M) 130 210 LD-TE- 18J (K) 140 212 RHR A HX Room LD-TE-18E (F) 130 212 LD-TE-18G (H) 150 212 RHR B Pump Room LD-TE-18A (8) 140 212 Attachment 7.2, Fission Product Barrier Loss/Potential Loss Matrix and Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 270 of 274 PPM 5.3.1 Table 24 - RB Area Radiation 24 RB Area Radiation Maximum Safe Area Insbrument Alarm Operating Value (mR/hr) (mPihr) (Rlhr)

East CRD Area ARM-RIS-4 104 N/A West CRD Area ARM-RIS-5 10 4 N/A H2 Recomber Area ARM-RIS-6 104 N/A TIP Drive Area ARM-RIS-7 104 N/A SGT Filter Area ARM-RIS44 RGT Pumper Aroo ARM-RINIA RHR Pump A Room ARM-RIS-9 a 104 N/A RHR Pump B Room ARM-RIS-l0 L ; 10. N/A RHR Pump C Room ARM-RIS-12 I10 49-4 NIA RClC Pump Room ARM-RIS-1 2 14 NIA 09 HPCS Pump Room ARM-RIS-13 E 4 N/A CRD Pump Room ARM-RIS-23 104 N/A 471 Elev - West Area ARM-RIS-24 10 4 N/A 471 EIev - NE ARM-RIS-32 N/A 10 501 Elev - NW ARM-RIS-33 N/A 10 CGS Basis Reference(s):

1. PPM 5.3.1 Secondary Containment Control

2. PPM 5.0.10 Flowchart Training Manual Attachment 7.2, Fission Product Barrier Loss/Potential Loss Matrix and Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 271 of 274 Barrier: Containment Category: D. Rad Degradation Threat: Loss Threshold:

None Attachment 7.2, Fission Product Barrier Loss/Potential Loss Matrix and Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 272 of 274 Barrier: Containment Category: D. Rad Degradation Threat: Potential Loss Threshold:

5. Containment Radiation Monitor CMS-RIS-27E or CMS-RIS-27F reading GT 14,000 R/hr NEI 99-01 Basis:

The 14,000 R/hr reading is a value that indicates significant fuel damage well in excess of that required for loss of RCS and Fuel Clad.

Regardless of whether containment is challenged, this amount of activity in containment, if released, could have such severe consequences that it is prudent to treat this as a potential loss of containment, such that a General Emergency declaration is warranted.

There is no Loss threshold associated with this item.

CGS Basis:

Four high range area radiation detectors (CMS-RE-27A, B, E and F) are installed in the drywell.

CMS-RE-27A and -27B are located in the bioshield wall at elevations 522' and 525', azimuth 600 and 2970, respectively. CMS-RE-27E and -27F are located inside containment at elevation 515', azimuth 2900 and 51.50, respectively. The companion containment radiation monitors (CMS-RIS-27A, B, E and F) are located on RAD Boards 22 and 23 in the Main Control Room. (Ref. 1)

The threshold value was calculated assuming the instantaneous release and dispersal of the reactor coolant noble gas and iodine inventory associated with 20% fuel clad damage into the drywell atmosphere. Evaluation of detector location, geometry and anticipated response suggests CMS-RIS-27E or F will provide the desired response to a given radiation source in the drywell and are, therefore, identified as the preferred monitors for evaluating this Containment barrier Potential Loss threshold. (Ref. 2)

CGS Basis Reference(s):

1. TM-2117 TSG - Core Thermal Engineer, Attachment 4.2

2. Calculation NE-02-94-57 Attachment 7.2, Fission Product Barrier Loss/Potential Loss Matrix and Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 273 of 274 Barrier: Containment Category: E. Judgment Degradation Threat: Loss Threshold:

6. &n condition in the opinion of the Emergency Director that indicates loss of the Containment barrier NEI 99-01 Basis:

This threshold addresses any other factors that are to be used by the Emergency Director in determining whether the Containment barrier is lost. In addition, the inability to monitor the barrier should also be incorporated in this threshold as a factor in Emergency Director judgment that the barrier may be considered lost.

The Containment barrier should not be declared lost based on exceeding Technical Specification action statement criteria, unless there is an event in progress requiring mitigation by the Containment barrier. When no event is in progress (Loss or Potential Loss of either Fuel Clad and/or RCS) the Containment barrier status is addressed by Technical Specifications.

CGS Basis:

The Emergency Director judgment threshold addresses any other factors relevant to determining if the Containment 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 two hours based on a projection of current safety system performance. The term "imminent' refers to recognition of the inability to reach safety 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 seguences lead to degradation of all fission product barriers and likely entry to the EOPs. The Emergency Director should be mindful of the Loss of AC power (Station Blackout) and ATWS EALs to assure timely emergency classification declarations.

CGS Basis Reference(s):

None Attachment 7.2, Fission Product Barrier Loss/Potential Loss Matrix and Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: Draft Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 274 of 274 Barrier: Containment Category: E. Judgment Degradation Threat: Potential Loss Threshold:

6. Any condition in the opinion of the Emergency Director that indicates potential loss of the Containment barrier NEI 99-01 Basis:

This threshold addresses any other factors that are to be used by the Emergency Director in determining whether the Containment barrier is potentially lost. In addition, the inability to monitor the barrier should also be incorporated in this threshold as a factor in Emergency Director judgment that the barrier may be considered potentially lost.

The Containment barrier should not be declared potentially lost based on exceeding Technical Specification action statement criteria, unless there is an event in progress requiring mitigation by the Containment barrier. When no event is in progress (Loss or Potential Loss of either Fuel Clad and/or RCS) the Containment barrier status is addressed by Technical Specifications.

CGS Basis:

The Emergency Director 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 degradation exists if the degradation will likely occur within two hours based on a projection of current safety system performance. The term "imminent' refers to recognition of the inability to reach safety 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 sequences lead to degradation of all fission product barriers and likely entry to the EOPs. The Emergency Director should be mindful of the Loss of AC power (Station Blackout) and ATWS EALs to assure timely emergency classification declarations.

CGS Basis Reference(s):

None Attachment 7.2, Fission Product Barrier Loss/Potential Loss Matrix and Bases