L-16-256, Response to Request for Additional Information Regarding a Request to Revise the Emergency Plan

Response to Request for Additional Information Regarding a Request to Revise the Emergency Plan

ML16250A855
Person / Time
Site:	Davis Besse
Issue date:	09/06/2016
From:	Boles B FirstEnergy Nuclear Operating Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
CAC MF7364, L-16-256
Download: ML16250A855 (275)

Text

5501 North State Route 2 Oak Harbor, Ohio 43449 FENOC Brian D. Boles 419-321-7676 Vice President - Nuclear Fax: 419-321-7582 September 6, 2016 L-16-256 10 CFR 50, Appendix E ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

SUBJECT:

Davis-Besse Nuclear Power Station, Unit No. 1 Docket No. 50-346, License No. NPF-3 Response to Request for Additional Information Regarding a Request to Revise the Emergency Plan (CAC No. MF7364)

By correspondence dated February 17, 2016 (Accession No. ML16049A513),

FirstEnergy Nuclear Operating Company (FENOC) submitted a request to revise the current Davis-Besse Nuclear Power Station (DBNPS) Emergency Plan emergency action level scheme to one based on Nuclear Energy Institute (NEI) 99-01, "Development of Emergency Action Levels for Non-Passive Reactors," Revision 6.

By correspondence dated July 22, 2016 (Accession No. ML16196A015), the Nuclear Regulatory Commission (NRG) requested additional information to complete its review.

Attachment 1 provides FENOC's response to this request. During development of the responses, FENOC identified additional changes needed to clarify the original February 17, 2016 submittal. Attachment 2 provides a summary of these changes. In addition, the DBNPS replaced the seismic monitoring system during the spring 2016 outage. Details regarding the seismic monitoring system upgrade have been incorporated into RAI response 2. As a result, the Emergency Action Level Technical Bases Document has been updated and is provided as an enclosure. No changes were identified to the previously provided significant hazards or environmental considerations.

On July 21, 2016, FENOC and NRG staff completed a teleconference during which an increase in the license amendment implementation period from 120 days to 180 days was discussed. As a result, FENOC is requesting a revised implementation period of 180 days following issuance of the amendment.

There are no regulatory commitments contained in this submittal. If there are any questions or additional information is required, please contact Mr. Thomas A Lentz, Manager- Fleet Licensing, at (330) 315-6810.

Davis-Besse Nuclear Power Station, Unit No. 1 l-16-256 Page2 I declare under penalty of perjury that the foregoing is true and correct. Executed on September~. 2016.

Sincerely, Brian D. Boles Attachments:

1. Response to July 22, 2016 Request for Additional Information

2. FENOC Identified Changes

Enclosure:

Emergency Action Level Technical Bases Document cc: NRC Region Ill Administrator NRC Resident Inspector NRC Project Manager Executive Director, Ohio Emergency Management Agency, State of Ohio (NRC Liaison)

Utility Radiological Safety Board

Attachment 1 L-16-256 Response to July 22, 2016 Request for Additional Information Page 1 of 11 By correspondence dated February 17, 2016, FirstEnergy Nuclear Operating Company (FENOC) submitted a license amendment request for Nuclear Regulatory Commission (NRC) review and approval. By correspondence dated July 22, 2016, NRC staff requested additional information to complete its review. The requested information is presented below in bold type, followed by the FENOC response.

RAI 1

Section 2.5, Technical Basis Information, of the EAL Technical Bases document describes the general format used for the EAL Bases. Under the Basis subheading, Section 2.5 states:

A Generic basis section that provides a description of the rationale for the EAL as provided in NEI 99-01 Rev. 6. This is followed by a Plant-Specific basis section that provides DBNPS-relevant information concerning the EAL.

This format is different than the format in NEI 99-01, Revision 6. The potential exists for decision-makers to be confused between these two sections if the information appears to be inconsistent. Justify having two separate bases sections or revise accordingly to eliminate potential confusion by the user.

Response

The Davis-Besse Nuclear Power Station (DBNPS) site-specific and NEI 99-01 generic bases sections have been combined into a single bases section for each EAL. The revised title for this combined section is now Basis and Section 2.5, Technical Bases Information, has been revised accordingly. Additionally, DBNPS Basis Reference(s) has been changed throughout and now reads Basis Reference(s). Redundant bases statements, where applicable, have been deleted.

L-16-256 Page 2 of 11 RAI 2 (EAL HU2.1)

Proposed EAL HU2.1 is for a seismic event greater than the operating bases earthquake (OBE).

a) The DBNPS basis for proposed EAL HU2.1 states that the OBE is ground motion acceleration of 0.08 g horizontally or 0.053 g vertically. Proposed EAL HU2.1 relies upon the OBE alarm on seismic alarm panel C5764A. The DBNPS basis for this EAL indicates that this alarm is for an earthquake of 0.08 g or greater. Thus, it appears that this proposed EAL would only address a seismic event exceeding the 0.08 g horizontal acceleration OBE. Explain how a seismic event that exceeds the vertical acceleration OBE, but not the horizontal acceleration OBE, would be declared under this EAL, or revise accordingly.

Response

The DBPS seismic monitoring system was replaced subsequent to the February 17, 2016 submittal of the proposed EAL scheme change. An updated description is provided as follows: Four triaxial strong motion accelerometers are installed within plant structures and networked together, such that any one accelerometer can also trigger the other devices and start event recording. Data is retrievable from each accelerometer (via an analog-to-digital recorder and data storage memory) through a computer in the Control Room. Time-history data and response spectra plots can be readily viewed on a video display without need for conversion to hardcopy. Quick and accurate determinations can be made based on the seismic event. Peak acceleration values are displayed showing whether the operating basis earthquake (OBE) limits have or have not been exceeded. Additionally, immediate Control Room alarm indication of an earthquake exceeding 0.08 g horizontal or 0.053g vertical is annunciated on the seismic alarm panel (C5764A), as sensed by the containment concrete foundation accelerograph.

EAL HU2.1 and associated bases have been revised to incorporate the new seismic monitoring system.

b) Section 4.6, Basis Document, of NEI 99-01, Revision 6, states:

A Basis section should not contain information that could modify the meaning or intent of the associated IC [initiating condition] or EAL. Such information should be incorporated within the IC or EAL statements, or as an EAL Note. Information in the Basis should only clarify and inform decision-making for an emergency classification.

The DBNPS basis for the proposed EAL HU2.1 state:

When the Seismic Monitoring Cabinet key locked switch is placed in OFF, ALL OBE and SSE [Safe Shutdown Earthquake] alarms and cabinet recording functions (TNC 8.3.3 functions 1 and 3 and the DBRM-EMER-5003 L-16-256 Page 3 of 11 Seismic Event Detection function) are nonfunctional. Therefore, the Compensatory Measures listed in DBRM-EMER-5003, Equipment Important to Emergency Response for a loss of the Seismic Event Detection function are required to be employed by the operators to determine if an earthquake occurs and to determine the magnitude of the event in a timely manner .

This statement appears to modify the proposed EAL. If the licensee intends to have a compensatory EAL for when seismic monitoring equipment is out of service, then this needs to be incorporated into the actual EAL (using OR logic) or as a note for the EAL, and the basis section clarified accordingly. Otherwise, this statement should be removed from the basis for the proposed EAL. If a compensatory EAL is added, explain how the timing of this EAL declaration would be affected.

Response

EAL HU2.1 has been revised to delete the cited statement.

RAI 3 (EALs HU4.1 and HU4.2)

Proposed EALs HU4.1 and HU4.2 are for fires that potentially degrade the level of safety of the plant.

a) As part of the proposed EAL HU4.1, the licensee lists fire detection indications including Receipt of multiple (more than 1) fire alarms (2nd bullet). This differs from the associated NEI 99-01, Revision 6, guidance which lists Receipt of multiple (more than 1) fire alarms or indications. The licensee stated that deletion of or indications was based upon it being redundant.

This EAL is intended to ensure that decision-makers consider multiple sources of information when an actual fire exists in order to make the appropriate EAL classification. The NEI 99-01, Revision 6, basis for this EAL states: In addition to alarms, other indications of a FIRE could be a drop in fire main pressure, automatic activation of a suppression system, etc. Revise the EAL to include or indications in bullet 2, or provide further justification for removal.

Response

EAL HU4.1 has been revised to include or indications in the second bullet.

b) The NEI 99-01 guidance for HU4 states that for both emergency declaration and fire duration the clock starts at the time the fire alarm, indication, or report is received, and not the time that a subsequent verification was performed.

The DBNPS basis information for proposed EALs HU4.1 and HU4.2 has many statements that are not consistent with the NEI 99-01 guidance and may cause confusion as to when this particular EAL is to be declared or when the clock starts for classification purposes.

L-16-256 Page 4 of 11 Clarify the basis section for EALs HU4.1 and HU4.2 related to EAL classification timeliness and clock start. If differences from NEI 99-01, Revision 6, are desired, provide justification for each difference and explain how these differences impact classification timeliness. In particular, explain the terms credible and valid so that a decision-maker can determine if it is acceptable to delay classification in order to determine validity or to wait until a credible source notifies the control room of a fire.

Response

EALs HU4.1 and HU4.2 have been revised to delete the DBNPS discussion related to classification timeliness.

c) The developer notes in NEI 99-01, Revision 6, for EAL HU4 state: The site-specific list of plant rooms or areas should specify those rooms or areas that contain SAFETY SYSTEM equipment. The areas listed in Table H-1, Safe Shutdown Fire Areas, of the DBNPS EAL basis document may be overly broad and result in an emergency declaration under HU4.1 or HU4.2 for a fire in a room that does not contain safety systems. Confirm that Table H-1 issufficiently detailed to ensure that unnecessary emergency declarations are not made or revise the list accordingly.

Response

Table H-1, Safe Shutdown Fire Areas, is based on the fire hazard analysis report and includes those structures containing functions and systems required for safe operation, shutdown, and cooldown of the plant (SAFETY SYSTEMS). The fire hazard analysis report has been previously provided to the NRC on November 21, 2014.

The SAFETY SYSTEMS identified in the report were reviewed to develop a balance between defining major plant structures containing safe shutdown equipment as fire areas versus a detailed list of areas for every safety system component location. The Table H-1 list of fire areas achieves that balance in support of timely and accurate emergency classification for the end-user. As such, no further reasonable refinement can be achieved.

d) Proposed EAL HU4.1 requires the notification of an unusual event if a fire is not extinguished within 15 minutes under certain conditions. Proposed EAL HU4.2 requires the notification of an unusual event if a fire is not verified within 30 minutes under certain conditions. The DBNPS basis section for proposed HU4.2 discusses the extinguishing of a fire, but the extinguishing of a fire is not part of the criteria for EAL HU4.2. If a fire is proven to exist, then EAL HU4.1 would apply. Clarify the basis for EAL HU4.2 with respect to extinguishing the fire.

Response

Per RAI 3.b response, the DBNPS discussion related to extinguishing of a fire has been deleted from EAL HU4.2.

L-16-256 Page 5 of 11 RAI 4 (EALs HA5.1, HS6.1, and RA3.2)

Proposed EALs HA5.1 and RA3.2 are for hazardous gas releases and radiation levels, respectively, that impede access to equipment necessary for normal plant operations, cooldown, or shutdown. Proposed HS6.1 is for the inability to control a key safety function following a control room evacuation.

NEI 99-01, Revision 6, EALs HA5, HS6, and AA3 provide guidance applicable to DBNPS proposed EALS HA5.1, HS6.1, and RA3.2, respectively. The NEI guidance indicates that these EALs are applicable in all modes. However, the licensees proposed EALs limit the applicability but the bases for these EALs do not explain why. Clarify the bases for these EALs to indicate they were evaluated for all modes and explain why there were limited to the specified modes.

Response

The mode applicability of EALs HA5.1 and RA3.2 were revised to align with the mode applicability restrictions imposed in Table H-2 and R-2, as justified in EAL Technical Bases Document Attachment 3, Safe Shutdown Rooms/Areas Tables R-2 & H-2 Bases.

The following note has been added to the HA5.1 and RA3.2 EAL bases to ensure any future change in Table H-2 or R-2 mode applicability is reflected in overall EAL applicability:

EAL HA5.1 was revised to add the following:

NOTE: IC HA5 mode applicability has been limited to the applicable modes identified in Table H-2 Safe Shutdown Rooms/Areas. If due to plant operating procedure or plant configuration changes, the applicable plant modes specified in Table H-2 are changed, a corresponding change to Attachment 3 Safe Shutdown Rooms/Areas Tables R-2 & H-2 Bases and to IC HA5 mode applicability is required.

EAL RA3.2 was revised to add the following:

NOTE: EAL RA3.2 mode applicability has been limited to the applicable modes identified in Table R-2 Safe Shutdown Rooms/Areas. If due to plant operating procedure or plant configuration changes, the applicable plant modes specified in Table R-2 are changed, a corresponding change to Attachment 3, Safe Shutdown Rooms/Areas Tables R-2 & H-2 Bases, and to EAL RA3.2 mode applicability is required.

The mode applicability of EAL HS6.1 has been revised consistent with NRC EPFAQ 2015-014 by expanding overall mode applicability to modes 1 through 6.

Reactivity control mode applicability has been restricted to modes 1, 2, and 3 only.

L-16-256 Page 6 of 11 RAI 5 (EALs RG1.1, RS1.1, RA1.1, and RU1.1)

Proposed EALs RG1.1, RS1.1, RA1.1, and RU1.1 are for the release of radioactive gases. The threshold values for the noble gas monitors for these EALs have all been reduced from the previously approved EAL scheme for DBNPS (ADAMS Accession No. ML083450120). Explain why these values have been reduced when the bases for these EALs has not changed.

Response

The proposed EAL RU1.1 values are based on the same NEI 99.01, Revision 5 calculation methodology. However, in June 2012 there was a change to the Offsite Dose Assessment Calculation Manual (ODCM), which changed the atmospheric dispersion factor used in the calculation.

The proposed EALs RG1.1, RS1.1, and RA1.1 calculation methodologies are different from the currently approved EAL scheme calculation methodologies in the following areas:

Source terms:

The currently approved EAL scheme source term is based on actual core inventory (referenced from the Updated Safety Analysis Report), and calculations were performed to derive fuel and gap activities used in the evaluation. Additional hand calculations were performed for process reduction factors.

The NEI 99-01, Revision 6, submittal source term is contained within the Meteorological Information and Dose Assessment System (MIDAS) software program. The source terms were determined during software development, based on the guidance contained in NUREG-1228, Source Term Estimation During Incident Response to Severe Nuclear Power Plant Accidents. This software was not in use when the currently approved EAL scheme was submitted.

Atmospheric dispersion factors:

The currently approved EAL scheme atmospheric dispersion factors were hand-calculated using guidance and methodology contained in Meteorology and Atomic Energy, 1968.

The NEI 99-01, Revision 6, submittal atmospheric dispersion factors are contained within the MIDAS dose assessment software program. The dispersion factors were determined during the software development. This software was not in use when the currently approved EAL scheme was submitted.

EAL setpoints:

The currently approved EAL scheme used a best fit curve to determine the EAL setpoints. Since the Unusual Event value did not use the accident source terms, the calculated value was higher than for the Alert values. The best fit curve was developed for end user ease of use, which was allowable under NEI 99-01, Revision 5 guidance.

The NEI 99-01, Revision 6, submittal EAL setpoints values were determined using the MIDAS program. The General Emergency value was empirically determined, and these values were entered into the MIDAS program, using the standardized accident parameters, until a corresponding offsite dose projection that met the L-16-256 Page 7 of 11 Protective Action Guideline value was met. This value was then ratioed for the Site Area Emergency and Alert values. Confirming calculations were performed to verify the results.

The EAL Calculations were previously provided to the NRC on February 17, 2016.

RAI 6 (EAL RA2.1)

Proposed EAL RA2.1 is for the uncovering of irradiated fuel in the refueling pathway and is referenced as an Unusual Event. However, it appears this should be categorized as an Alert. Revise EAL RA2.1 to correct this issue.

Response

EAL RA2.1 classification designation has been corrected to reflect the Alert classification.

RAI 7 (EALs CG1.1 and CS1.1)

Proposed EALs CG1.1 and CS1.1 are for the loss of reactor coolant system (RCS) inventory and are applicable in modes 5 and 6. These EALs include readings from the Refueling Bridge Portable Area Radiation Monitor. The DBNPS bases for these EALs state: The Refuel Bridge Portable Radiation Monitor, when installed during refueling operations, is designed to provide monitoring of radiation due to a fuel handling event or loss of shielding during refueling operations.

As written, proposed EALs CG1.1 and CS1.1 cannot be fully implemented since the Refueling Bridge Portable Area Radiation Monitor is not always available during the applicable modes. Either revise these EALs to indication that the monitor may be used when installed during refueling operations or revise the bases to indicate that the monitor will be available during the applicable modes.

Response

EALs CG1.1 and CS1.1 have been revised and added (when installed) to Refueling Bridge Portable Area Radiation Monitor reading > 30 R/hr.

L-16-256 Page 8 of 11 RAI 8 (EALs CU2.1, SA1.1, and SU1.1)

Proposed EALs CU2.1, SA1.1, and SU1.1 are for the loss of alternating current (ac) power sources to essential buses.

a) Proposed EALs CU2.1, SA1.1, and SU1.1 reference tables which list transformer X11 (back-fed via Main Transformer) as an offsite power source.

The DBNPS bases for these EALs state: Credit for the X11 back-feed can only be taken if already aligned, as it takes greater than 15 minutes to align.

As noted previously, the bases should not modify the IC or the EAL. Revise the listing of transformer X11 (back-fed via Main Transformer) to clarify that it can be credited as an offsite source if already aligned.

Response

Tables C-2 and S-1 Offsite/Onsite AC Power Sources, have been revised specifically the bullet addressing X11 back-fed via Main Transformer under Offsite to specify if already aligned.

b) Proposed EALs SA1.1 and SU1.1 reference tables listing transformer X11 and X11 (back-fed via Main Transformer) as offsite power sources. The DBNPS bases for these EALs state:

The essential buses during plant operation are normally powered from the 13.8KV [kilovolt] offsite power system through their respective 13.8KV/4160V [volt] bus tie transformers, via the Unit Auxiliary Transformer (X11). In non-power operating modes, the essential buses may be back-fed via the X11 and Main Transformer provided the main generator lead disconnect links are removed....

The two listings of transformer X11 may cause confusion. Revise the tables for SA1.1 and SU1.1 to clarify the use of transformer X11 as an offsite power source of justify the current listing.

Response

Transformer X11 (during normal power operations) and Transformer X11 (back-fed via Main Transformer) are two distinct offsite AC power sources that would not be confused by plant operators or the Emergency Response Organization. As such, no additional changes were made to these EALs or Table S-1.

L-16-256 Page 9 of 11 RAI 9 (EALs CA2.1, SG1.1, SG1.2, and SS1.1)

Proposed EALs CA2.1, SG1.1, and SS1.1 are for the loss of all ac power sources to essential power buses. Proposed EAL SG1.2 is for the loss of all essential ac and direct current power sources. The NEI 99-01, Revision 6, guidance uses the following wording as part of these EALs: Loss of ALL offsite and ALL onsite AC Power to (site-specific emergency buses). Proposed EALs SG1.1, SG1.2, and SS1.1, contain the following wording: Loss of ALL offsite and ALL onsite AC power capability, Table S-1, to essential 4160V buses C1 and D1, where Table S-1 lists offsite and onsite ac power sources. Proposed EAL CA2.1 has similar wording except it refers to Table C-2.

The focus of these EALs is on a complete loss of power to essential buses.

Tables S-1 and C-2 appear to be unnecessary and may cause confusion. In addition, alternative power sources, such as those used for a mitigation strategy, may be able to power the essential buses. Revise these EALs to remove Tables S-1 and C-2 or justify their use with these EALs.

Response

The following basis section changes were made:

EAL CA2.1 was revised to delete Table C-2 and a reference to Table C-2.

EAL SS1.1 was revised to delete Table S-1 and a reference to Table S-1.

EAL SG1.1 was revised to delete Table S-1 and a reference to Table S-1.

EAL SG1.2 was revised to delete Table S-1 and a reference to Table S-1.

RAI 10 (CU5.1, CU5.2, CU5.3, SU7.1, SU7.2, and SU7.3)

Proposed EALs CU5.1, CU5.2, CU5.3, SU7.1, SU7.2, and SU7.3 address the loss of communications equipment.

a) Emergency notifications to offsite response organizations and to the NRC are expected to be made from the control room. If satellite phones or cellular phones cannot be used in the control room due to radio frequency interference, then they should not be included in the list of available communications methods for EALs CU5.1, CU5.2, CU5.3, SU7.1, SU7.2, and SU7.3. Remove satellite phones and cellular phones from the list of available communications methods or justify retaining them on the list.

b) Satellite phones are included in the list of available communication methods for EALs CU5.1, CU5.2, CU5.3, SU7.1, SU7.2, and SU7.3. Explain how satellite phones can be used inside a building, where a responder is protected from radiological hazards, or remove them from the list.

c) Cellular phones are included in the list of available communication methods for EALs CU5.1, CU5.2, CU5.3, SU7.1, SU7.2, and SU7.3. The NEI guidance associated with these EALs indicates that credit for personal devices should not be credited. Remove cellular phones from the list or clarify the list to indicate that they are dedicated phones.

L-16-256 Page 10 of 11

Response

Satellite phones and cellular phones have been removed from Table C-4 and Table S-5 and associated bases revised as appropriate.

RAI 11 (EALs SA3.1 and SU3.1)

Proposed EALs SA3.1 and SU3.1 are for the loss of control room indications. The associated NEI 99-01, Revision 6, guidance for these EALs list RCS level as one of the important parameters for monitoring. The NEI 99-01 developer notes state that either the pressurizer or reactor vessel level may be specified in the EAL in place of RCS level. The proposed EALs SA3.1 and SU3.1 do not include RCS level or one of the alternatives in the list of safety system parameters. Justify not including RCS level in the list of safety parameters for these EALs or revise the list to include RCS level or an equivalent parameter.

Response

Table S-2 has been revised to add Pressurizer level.

RAI 12 (EAL SU4.1)

Proposed EAL SU4.1 is for reactor coolant activity greater than the technical specification allowable limits.

a) Explain why instrument RE 1998 is called the Letdown Monitor in proposed EAL SU4.1 and the Failed Fuel Monitor in the basis section. Clarify the EAL and basis accordingly.

b) The DBNPS basis section states: A monitor value of 2.0E+06 (0.1% clad damage) was chosen for its ability to be recognized even though exceeding Technical Specification Limits could potentially result in much higher readings. The NEI 99-01, Revision 6, developer notes applicable to this EAL state:

The monitor reading values should correspond to an RCS activity level approximately at Technical Specification allowable limits.

If there is no existing method/capability for determining this EAL, then it should not be included.

Provide justification for including EAL SU4.1 when it appears that the monitoring instrument reading would not correspond to an RCS activity level near the technical specification limits.

Response

Proposed EAL SU4.1 has been deleted. There is no direct correlation of DBNPS Technical Specification coolant activity and a reading on RE 1998 Failed Fuel Monitor.

As a result, proposed EAL SU4.2 has been re-numbered and is now SU4.1.

L-16-256 Page 11 of 11 RAI 13 (EAL SU5.1 and NEI 99-01, Revision 6, EAL SU4)

NEI 99-01, Revision 6, SU4 provides guidance for the notification of an unusual event for RCS leakage, which includes three separate EALs:

SU4.1 is for unidentified and pressure boundary RCS leakage; SU4.2 is for identified RCS leakage; and SU4.3 is for RCS leakage to a location outside of containment.

Proposed DBNPS EAL SU5.1 would require an unusual event to be declared for RCS leakage greater than 10 gallons per minute (gpm) for greater than or equal to 15 minutes. DBNPS does not distinguish between the different types of leakage as recommended in NEI 99-01. The DBNPS basis states:

DBNPS does not have the capability to classify leakage as identified leakage within 15 minutes. Therefore, for the purpose of this IC and EAL all RCS leakage is considered unidentified leakage and the 10 gpm leak rate applies.

a) The NEI guidance for EAL SU4.2 specifies a threshold of 25 gpm, which is greater than the 10 gpm threshold for the DBNPS EAL SU5.1. Justify why NEI EAL SU4.3 was not developed for DBNPS, taking into consideration the difference in the thresholds.

b) The currently approved EAL scheme for DBNPS includes EALs similar to NEI SU4.1 and SU4.2. In addition, the statement from the DBNPS basis does not appear to consider that there can be previously identified leakage. Explain in further detail why EAL SU5.1 could not include identified leakage as a threshold given that is currently approved for DBNPS.

Response

Proposed EAL SU5.1 has been revised to include thresholds for identified leakage

> 25 gpm as well as leakage from the RCS to a location outside containment > 25 gpm.

The EAL bases have been revised to include the NEI bases for identified leakage greater than 25 gpm and leakage from the RCS to a location outside containment. In addition, EAL SU5.1 has been revised to delete the cited statement.

Attachment 2 L-16-256 FENOC Identified Changes Page 1 of 3 The following FirstEnergy Nuclear Operating Company (FENOC) identified changes have been included in the updated technical bases document. For each change, the affected emergency action level (EAL) or area within the technical bases document is presented below in bold type, followed by a brief description and basis for the change.

EAL HU4.1 Description of Change:

Deleted generic bases related to Appendix R from HU4.1. The information pertaining to Appendix R is only applicable to EAL HU4.2.

EAL HG1.1 Description of Change:

Deleted IC HG1 and EAL HG1.1. As stated in EPFAQ 2015-013, there are several other redundant Initiating Conditions (IC) better suited to ensure timely and effective emergency declarations.

Since the current IC HG1 has two distinct parts, they will be addressed separately as follows:

1. Hostile Action in the Protected Area is bounded by ICs HS1 and HS7. Hostile Action resulting in a loss of physical control is bound by EAL HG7, as well as any event that may lead to radiological releases to the public in excess of Environmental Protection Agency (EPA) Protective Action Guides (PAGs).

a. If, for whatever reason, the Control Room must be evacuated, and control of safety functions (reactivity control, core cooling, and RCS heat removal) cannot be reestablished, then IC HS6 would apply, as well as IC HS7 if desired by the EAL decision-maker.

b. Also, as stated above, any event (including Hostile Action) that could reasonably be expected to have a release exceeding EPA PAGs would be bound by IC HG7.

c. From a Hostile Action perspective, ICs HS1, HS7 and HG7 are appropriate, and therefore, make this part of HG1 redundant and unnecessary.

d. From a loss of physical control perspective, ICs HS6, HS7 and HG7 are appropriate, and therefore, make this part of HG1 redundant and unnecessary.

L-16-256 Page 2 of 3

2. Any event which causes a loss of spent fuel pool level will be bounded by ICs RA2, RS2 and RG2, regardless of whether it was based upon a Hostile Action or not, thus making this part of HG1 redundant and unnecessary.

a. An event that leads to a radiological release will be bounded by ICs RU1, RA1, RS1 and RG1. Events that lead to radiological releases in excess of EPA PAGs will be bounded by EALs RG1 and HG7, thus making this part of HG1 redundant and unnecessary.

Based on these considerations, and given the confusion these redundant EALs had on EAL decision-making at the General Emergency level, DBNPS determined not to include HG1 in a site-specific EAL scheme. However, ICs RA2, RS2, RG2, RS1, RG1, HS1, HS6, HS7 and HG7 have been incorporated into the proposed site-specific EAL scheme to ensure the intended event is appropriately bound at the correct ECL.

The Table of Contents and Section 6, DBNPS TO NEI 99-01 Rev. 6 EAL CROSS-REFERENCE, have been revised and pages renumbered to reflect the deletion of IC HG1.

IC HU1 Description of Change:

Revised EALs HU1.1, HU1.2, and HU1.3 basis language referencing security events assessed as hostile actions and deleted HG1 as a classifiable IC. The bases section now reads Security events assessed as HOSTILE ACTIONS are classifiable under ICs HA1 and HS1.

IC HA1 Description of Change:

Revised EALs HA1.1 and HA1.2 basis language and deleted reference to HG1 as an emergency classification level escalation path. Revised language now reads Escalation of emergency classification level would be via IC HS1.

IC HS1 Description of Change:

Revised EAL HS1.1 basis language and deleted reference to HG1 as an emergency classification level escalation path. Revised language now reads Escalation of emergency classification level would be via IC FG1.

L-16-256 Page 3 of 3 IC SA3 Description of Change:

Revised EAL SA3.1 secondary threshold ANY Significant Transient is or may be in progress, Table S-3 and deleted or may be. The additional language was deleted from the EAL to better align with the NEI 99-01, Revision 6 language and to eliminate potential confusion to the operators.

Enclosure L-16-256 Emergency Action Level Technical Bases Document (258 Pages Follow)

DAVIS-BESSE REFERENCE MATERIAL EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT IEAL BASES DOCUMENT Rev. 0 Page 1 of 2581

TABLE OF CONTENTS SECTION PAGE 1.0 PURPOSE ................................................................................................................................... 6 2.0 DISCUSSION ............................................................................................................................... 6 2.1 Background ............................................................................................................................... 6 2.2 Fission Product Barriers ............................................................................................................ 7 2.3 Fission Product Barrier Classification Criteria ........................................................................... 7 2.4 EAL Organization ...................................................................................................................... 8 2.5 Technical Bases Information ................................................................................................... 10 2.6 Operating Mode Applicability .................................................................................................. 11 3.0 GUIDANCE ON MAKING EMERGENCY CLASSIFICATIONS .................................................. 12 3.1 General Considerations .......................................................................................................... 12 3.2 Classification Methodology ..................................................................................................... 13

4.0 REFERENCES

........................................................................................................................... 17 4.1 Developmental ........................................................................................................................ 17 4.2 Implementing .......................................................................................................................... 17 5.0 DEINITIONS, ACRONYMS & ABBREVIATIONS ....................................................................... 18 5.1 Definitions ..18 5.2 Abbreviations/Acronyms . 23 6.0 DBNPS TO NEI 99-01 Rev. 6 EAL CROSS-REFERENCE ........................................................ 26 7.0 ATTACHMENTS ........................................................................................................................ 30 1 Emergency Action Level Technical Bases ..................................................................... 30 Category H Hazards ............................................................................................. 31 HS1.1 ................................................................................................ 33 HA1.1 ................................................................................................ 35 HA1.2 ................................................................................................ 37 HU1.1 ................................................................................................ 39 HU1.2 ................................................................................................ 41 HU1.3 ................................................................................................ 43 HU2.1 ................................................................................................ 45 HU3.1 ................................................................................................ 47 HU3.2 ................................................................................................ 48 HU3.3 ................................................................................................ 49 HU3.4 ................................................................................................ 50 HU4.1 ................................................................................................ 51 EAL BASES DOCUMENT Rev. 0 Page 2 of 258

HU4.2 ................................................................................................ 53 HU4.3 ................................................................................................ 56 HU4.4 ................................................................................................ 57 HA5.1 ................................................................................................ 58 HS6.1 ................................................................................................ 61 HA6.1 ................................................................................................ 63 HG7.1 ................................................................................................ 65 HS7.1 ................................................................................................ 67 HA7.1 ................................................................................................ 68 HU7.1 ................................................................................................ 69 Category R Abnormal Rad Release / Rad Effluent ............................................... 70 RG1.1 ................................................................................................ 71 RG1.2 ................................................................................................ 73 RG1.3 ................................................................................................ 75 RS1.1 ................................................................................................ 77 RS1.2 ................................................................................................ 79 RS1.3 ................................................................................................ 81 RA1.1 ................................................................................................ 83 RA1.2 ................................................................................................ 85 RA1.3 ................................................................................................ 87 RA1.4 ................................................................................................ 89 RU1.1 ................................................................................................ 91 RU1.2 ................................................................................................ 94 RG2.1 ................................................................................................ 96 RS2.1 ................................................................................................ 97 RA2.1 ................................................................................................ 98 RA2.2 .............................................................................................. 100 RA2.3 .............................................................................................. 102 RU2.1 .............................................................................................. 103 RA3.1 .............................................................................................. 105 RA3.2 .............................................................................................. 106 Category E DFSF ............................................................................................... 108 EU1.1 .............................................................................................. 109 Category C Cold Shutdown / Refueling System Malfunction .............................. 111 CG1.1 .............................................................................................. 113 EAL BASES DOCUMENT Rev. 0 Page 3 of 258

CS1.1 .............................................................................................. 117 CA1.1 .............................................................................................. 119 CA1.2 .............................................................................................. 120 CU1.1 .............................................................................................. 122 CU1.2 .............................................................................................. 124 CA2.1 .............................................................................................. 126 CU2.1 .............................................................................................. 128 CA3.1 .............................................................................................. 131 CU3.1 .............................................................................................. 134 CU3.2 .............................................................................................. 136 CU4.1 .............................................................................................. 138 CU5.1 .............................................................................................. 140 CU5.2 .............................................................................................. 142 CU5.3 .............................................................................................. 144 CA6.1 .............................................................................................. 146 Category S System Malfunction ......................................................................... 149 SG1.1 .............................................................................................. 151 SG1.2 .............................................................................................. 154 SS1.1............................................................................................... 157 SA1.1............................................................................................... 159 SU1.1 .............................................................................................. 162 SS2.1............................................................................................... 164 SA3.1............................................................................................... 166 SU3.1 .............................................................................................. 169 SU4.1 .............................................................................................. 171 SU5.1 .............................................................................................. 172 SS6.1............................................................................................... 174 SA6.1............................................................................................... 176 SU6.1 .............................................................................................. 178 SU6.2 .............................................................................................. 181 SU7.1 .............................................................................................. 184 SU7.2 .............................................................................................. 186 SU7.3 .............................................................................................. 188 SU8.1 .............................................................................................. 190 SU8.2 .............................................................................................. 191 EAL BASES DOCUMENT Rev. 0 Page 4 of 258

SA9.1............................................................................................... 193 Category F Fission Product Barrier Degradation................................................ 195 FG1.1 .............................................................................................. 197 FS1.1 ............................................................................................... 198 FA1.1 ............................................................................................... 199 2 Fission Product Barrier Loss / Potential Loss Matrix and Bases ......................................................................................................... 200 3 Safe Shutdown Rooms/Areas Tables R-2 & H-2 Bases .............................................. 249 EAL BASES DOCUMENT Rev. 0 Page 5 of 258

1.0 PURPOSE This document provides an explanation and rationale for each Emergency Action Level (EAL) included in the EAL Upgrade Project for Davis-Besse Nuclear Power Station (DBNPS). It should be used to facilitate review of the DBNPS EALs and provide historical documentation for future reference. Decision-makers responsible for implementation of RA-EP-01500 Emergency Classification, may use this document as a technical reference in support of EAL interpretation. This information may assist the Emergency Director in making classifications, particularly those involving judgment or multiple events. The basis information may also be useful in training and for explaining event classifications to off-site officials.

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

Because the information in a basis document can affect emergency classification decision-making (e.g., the Emergency Coordinator refers to it during an event), the NRC staff expects that changes to the basis document will be evaluated in accordance with the provisions of 10 CFR 50.54(q).

2.0 DISCUSSION 2.1 Background EALs are the plant-specific indications, conditions or instrument readings that are utilized to classify emergency conditions defined in the DBNPS Emergency Plan (ref. 4.1.13).

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

NEI 99-01 (NUMARC/NESP-007) Revisions 4 and 5 were subsequently issued for industry implementation. Enhancements over earlier revisions included:

Consolidating the system malfunction initiating conditions and example emergency action levels, which address conditions that may occur during plant shutdown conditions.

Initiating conditions and example emergency action levels that fully address conditions that may be postulated to occur at permanently Defueled Stations and Independent Spent Fuel Storage Installations (ISFSIs).

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

Subsequently, Revision 6 of NEI 99-01 has been issued which incorporates resolutions to numerous implementation issues including the NRC EAL Frequently Asked Questions (FAQs).

Using NEI 99-01 Revision 6, "Methodology for the Development of Emergency Action Levels for Non-Passive Reactors, (ref. 4.1.1), DBNPS conducted an EAL implementation upgrade project that produced the EALs discussed herein.

EAL BASES DOCUMENT Rev. 0 Page 6 of 258

2.2 Fission Product Barriers FISSION PRODUCT BARRIER THRESHOLDS represent threats to the defense in depth design concept that precludes the release of radioactive fission products to the environment.

This concept relies on multiple physical barriers, any one of which, if maintained intact, precludes the release of significant amounts of radioactive fission products to the environment.

Many of the EALs derived from the NEI methodology are FISSION PRODUCT BARRIER THRESHOLD based. That is, the conditions that define the EALs are based upon thresholds that represent the loss or potential loss of one or more of the three fission product barriers.

Loss and Potential Loss signify the relative damage and threat of damage to the barrier. A Loss threshold means the barrier no longer assures containment of radioactive materials. A Potential Loss threshold implies an increased probability of barrier loss and decreased certainty of maintaining the barrier.

The primary fission product barriers are:

A. Fuel Clad (FC): The Fuel Clad Barrier consists of the cladding material that contains the fuel pellets.

B. Reactor Coolant System (RC): The RC Barrier includes the RCS primary side and its connections up to and including the pressurizer safety and relief valves, and other connections up to and including the primary isolation valves.

C. Containment (CT): The Containment Barrier includes the containment building, and connections up to and including the outermost containment isolation valves. This barrier also includes the main steam, feedwater, and blowdown line extensions outside the containment building up to and including the outermost secondary side isolation valve. Containment Barrier thresholds are used as criteria for escalation of the ECL from Alert to a Site Area Emergency or a General Emergency 2.3 Fission Product Barrier Classification Criteria The following criteria are the bases for event classification related to fission product barrier loss or potential loss:

Alert:

ANY loss or ANY potential loss of EITHER Fuel Clad or RC Barrier Site Area Emergency:

Loss or potential loss of ANY two barriers General Emergency:

Loss of ANY two barriers AND loss or potential loss of the third barrier EAL BASES DOCUMENT Rev. 0 Page 7 of 258

2.4 EAL Organization The DBNPS EAL scheme includes the following features:

Division of the EAL set into three broad groups:

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

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

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

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.

Within each group, assignment of EALs to recognition categories and subcategories:

Recognition category and subcategory titles were selected to represent conditions that are operationally significant to the EAL-user. The DBNPS EAL recognition categories align to and represent the NEI 99-01 Revision 6 Recognition Categories. Subcategories are used in the DBNPS scheme to further divide the EALs of a category into logical sets of possible emergency classification thresholds. The DBNPS EAL categories and subcategories are listed below.

EAL BASES DOCUMENT Rev. 0 Page 8 of 258

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

H - Hazards and Other Conditions 1 - Security Affecting Plant Safety 2 - Seismic Event 3 - Natural or Technological Hazard 4 - Fire 5 - Hazardous Gas 6 - Control Room Evacuation 7 - Emergency Director Judgment R - Abnormal Rad Levels / Rad Effluent 1 - Radiological Effluent 2 - Irradiated Fuel Event 3 - Area Radiation Levels E - Dry Fuel Storage Facility (DFSF) 1 - Confinement Boundary Hot Conditions:

S - System Malfunction 1 - Loss of Essential AC Power 2 - Loss of Essential DC Power 3 - Loss of Control Room Indications 4 - RCS Activity 5 - RCS Leakage 6 - RPS Failure 7 - Loss of Communications 8 - Containment Failure 9 - Hazardous Event Affecting Safety Systems F - Fission Product Barrier Degradation None Cold Conditions:

C - Cold Shutdown / Refueling System 1 - RCS Level Malfunction 2 - Loss of Essential AC Power 3 - RCS Temperature 4 - Loss of Essential DC Power 5 - Loss of Communications 6 - Hazardous Event Affecting Safety Systems The primary tool for determining the emergency classification level is the EAL Classification Matrix. The user of the EAL Classification Matrix may (but is not required to) consult the EAL Technical Bases Document in order to obtain additional information concerning the EALs under classification consideration. The user should consult Section 3.0 and Attachments 1 & 2 of this document for such information.

EAL BASES DOCUMENT Rev. 0 Page 9 of 258

2.5 Technical Bases Information EAL technical bases are provided in Attachment 1 for each EAL according to EAL group (ANY, Hot, Cold), EAL category (R, C, H, S, F and E) and EAL subcategory. A summary explanation of each category and subcategory is given at the beginning of the technical bases discussions of the EALs included in the category. For each EAL, the following information is provided:

Category Letter & Title Subcategory Number & Title Initiating Condition (IC)

Site-specific description of the generic IC given in NEI 99-01 Rev. 6 EAL Identifier (enclosed in rectangle)

Each EAL is assigned a unique identifier to support accurate communication of the emergency classification to onsite and offsite personnel. Four characters define each EAL identifier:

1. First character (letter): Corresponds to the EAL recognition category as described in Section 2.4 (H, R, E, C, S or F)

2. Second character (letter): The emergency classification (G, S, A or U)

G = General Emergency S = Site Area Emergency A = Alert U = Unusual Event

3. Third character (number): Subcategory number within the given category.

Subcategories are sequentially numbered beginning with the number one (1). If a category does not have a subcategory, this character is assigned the number one (1).

4. Fourth character (number): The numerical sequence of the EAL within the EAL subcategory. If the subcategory has only one EAL, it is given the number one (1).

Classification (enclosed in rectangle):

Unusual Event (U), Alert (A), Site Area Emergency (S) or General Emergency (G)

EAL (enclosed in rectangle)

Exact wording of the EAL as it appears in the EAL Classification Matrix EAL BASES DOCUMENT Rev. 0 Page 10 of 258

Mode Applicability One or more of the following plant operating conditions comprise the mode to which each EAL is applicable: 1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown, 5 - Cold Shutdown, 6 - Refueling, D - Defueled, or ALL. (See Section 2.6 for operating mode definitions)

Basis:

A basis section that provides a description of the rationale for the EAL as provided in NEI 99-01, Rev. 6, as well as DBNPS-relevant information concerning the EAL.

Basis Reference(s):

Site-specific source documentation from which the EAL is derived 2.6 Operating Mode Applicability (ref. 4.1.8)

Reactivity Avg. Reactor Coolant Mode  % Rated Power*

Condition (Keff) Temperature (ºF)

1) Power Operation 0.99 > 5% N/A

2) Startup 0.99 5% N/A

3) Hot Standby < 0.99 N/A 280

4) Hot Shutdown < 0.99 N/A 280 > Tavg > 200

5) Cold Shutdown < 0.99 N/A 200

6) Refueling One or more vessel head closure bolts less than fully tensioned.

ALL reactor fuel removed from reactor pressure vessel (full core offload D) Defueled during refueling or extended outage).

Refer to Section 3.3.2 for guidance on event caused mode changes.

EAL BASES DOCUMENT Rev. 0 Page 11 of 258

3.0 GUIDANCE ON MAKING EMERGENCY CLASSIFICATIONS 3.1 General Considerations When making an emergency classification, the Emergency Director must consider ALL information having a bearing on the proper assessment of an Initiating Condition (IC). This includes the Emergency Action Level (EAL) plus the associated Operating Mode Applicability, Notes, and the technical basis information. In the Recognition Category F matrices, EALs are based on loss or potential loss of FISSION PRODUCT BARRIER THRESHOLDS.

3.1.1 Classification Timeliness NRC regulations require the licensee to establish and maintain the capability to assess, classify, and declare an emergency condition within 15 minutes after the availability of indications to plant operators that an emergency action level has been exceeded and to promptly declare the emergency condition as soon as possible following identification of the appropriate emergency classification level. The NRC staff has provided guidance on implementing this requirement in NSIR/DPR-ISG-01, "Interim Staff Guidance, Emergency Planning for Nuclear Power Plants" (ref. 4.1.11).

3.1.2 Valid Indications ALL emergency classification assessments shall be based upon VALID indications, reports or conditions. A VALID indication, report, or condition, is one that has been verified through appropriate means such that there is no doubt regarding the indicators operability, the conditions existence, or the reports accuracy. For example, verification could be accomplished through an instrument channel check, response on related or redundant indicators, or direct observation by plant personnel.

An indication, report, or condition is considered 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 indicators operability, the conditions existence, or the reports accuracy is removed. The validation of indications should be completed in a manner that supports timely emergency declaration.

3.1.3 IMMINENT Conditions For ICs and EALs that have a stipulated time duration (e.g., 15 minutes, 30 minutes, etc.), the Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the condition has exceeded, or will likely exceed, the applicable time. If an ongoing radiological release is detected and the release start time is unknown, it should be assumed that the release duration specified in the IC/EAL has been exceeded, absent data to the contrary.

3.1.4 Planned vs. Unplanned Events A planned work activity that results in an expected event or condition which meets or exceeds an EAL does not warrant an emergency declaration provided that: 1) the activity proceeds as planned, and 2) the plant remains within the limits imposed by the operating license. Such activities include planned work to test, manipulate, repair, maintain or modify a system or EAL BASES DOCUMENT Rev. 0 Page 12 of 258

component. In these cases, the controls associated with the planning, preparation and execution of the work will ensure that compliance is maintained with ALL aspects of the operating license provided that the activity proceeds and concludes as expected. Events or conditions of this type may be subject to the reporting requirements of 10 § CFR 50.72 (ref. 4.1.4).

3.1.5 Classification Based on Analysis The assessment of some EALs is based on the results of analyses that are necessary to ascertain whether a specific EAL threshold has been exceeded (e.g., dose assessments, chemistry sampling, RCS leak rate calculation, etc.). For these EALs, the EAL wording or the associated basis discussion will identify the necessary analysis. In these cases, the 15-minute declaration period starts with the availability of the analysis results that show the threshold to be exceeded (i.e., this is the time that the EAL information is first available). The NRC expects licensees to establish the capability to initiate and complete EAL-related analyses within a reasonable period of time (e.g., maintain the necessary expertise on-shift) (ref. 4.1.11).

3.1.6 Emergency Director Judgment While the EALs have been developed to address a full spectrum of possible events and conditions which may warrant emergency classification, a provision for classification based on operator/management experience and judgment is still necessary. The NEI 99-01 EAL scheme provides the Emergency Director with the ability to classify events and conditions based upon judgment using EALs that are consistent with the Emergency Classification Level (ECL) definitions (refer to Category H). The Emergency Director will need to determine if the effects or consequences of the event or condition reasonably meet or exceed a particular ECL definition. A similar provision is incorporated in the Fission Product Barrier Tables; judgment may be used to determine the status of a fission product barrier.

3.1.7 Emergency Action Levels with Embedded Time Requirements Some EALs have embedded time requirements. Declaration must be made as soon as the Emergency Director recognizes that the conditions will not be successfully resolved within 15 minutes. Therefore, for EALs with time-embedded requirements the time for emergency declaration starts with the initial alarm or indication of the event, not after the embedded time has elapsed.

For EALs with longer embedded time requirements, the 15-minute clock for declaration begins with recognition that the assigned time limit will be exceeded. For example, SG1.1 -

Restoration of at least one essential bus in < 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is not likely. If 20 minutes after loss of the C-1 bus and the D-1 bus it becomes apparent that restoration of either bus will not likely occur within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> the Emergency Director must immediately declare the General Emergency.

3.2 Classification Methodology To make an emergency classification, the user will compare an event or condition (i.e., the relevant plant indications and reports) to an EAL(s) and determine if the EAL has been met or exceeded. The evaluation of an EAL must be consistent with the related Operating Mode Applicability and Notes. If an EAL has been met or exceeded, the associated IC is likewise met, the emergency classification process clock starts, and the ECL must be declared in EAL BASES DOCUMENT Rev. 0 Page 13 of 258

accordance with plant procedures no later than fifteen minutes after the process clock started.

When assessing an EAL that specifies a time duration for the off-normal condition, the clock for the EAL time duration runs concurrently with the emergency classification process clock.

For a full discussion of this timing requirement, refer to NSIR/DPR-ISG-01 (ref. 4.1.11).

3.2.1 Classification of Multiple Events and Conditions When multiple emergency events or conditions are present, the user will identify ALL met or exceeded EALs. The highest applicable ECL identified during this review is declared. For example:

If an Alert EAL and a Site Area Emergency EAL are met, a Site Area Emergency should be declared.

There is no additive effect from multiple EALs meeting the same ECL. For example:

If two Alert EALs are met, an Alert should be declared.

Related guidance concerning classification of rapidly escalating events or conditions is provided in Regulatory Issue Summary (RIS) 2007-02, Clarification of NRC Guidance for Emergency Notifications During Quickly Changing Events (ref. 4.1.2).

3.2.2 Consideration of Mode Changes During Classification The mode in effect at the time an event or condition occurred, and prior to any plant or operator response, is the mode that determines whether an IC is applicable. If an event or condition occurs, and results in a mode change before the emergency is declared, the emergency classification level is still based on the mode that existed at the time that the event or condition was initiated (and not when it was declared). Once a different mode is reached, any new event or condition, not related to the original event or condition, requiring emergency classification should be evaluated against the ICs and EALs applicable to the operating mode at the time of the new event or condition.

For events that occur in Cold Shutdown or Refueling, escalation is via EALs that are applicable in the Cold Shutdown or Refueling modes, even if Hot Shutdown (or a higher mode) is entered during the subsequent plant response. In particular, the fission product barrier EALs are applicable only to events that initiate in the Hot Shutdown mode or higher.

3.2.3 Classification of IMMINENT Conditions Although EALs provide specific thresholds, the Emergency Director must remain alert to events or conditions that could lead to meeting or exceeding an EAL within a relatively short period of time (i.e., a change in the ECL is IMMINENT). If, in the judgment of the Emergency Director, meeting an EAL is IMMINENT, the emergency classification should be made as if the EAL has been met. While applicable to ALL emergency classification levels, this approach is particularly important at the higher emergency classification levels since it provides additional time for implementation of protective measures.

EAL BASES DOCUMENT Rev. 0 Page 14 of 258

3.2.4 Emergency Classification Level Upgrading and Downgrading An ECL may be downgraded when the event or condition that meets the highest IC and EAL no longer exists, and other site-specific downgrading requirements are met. If downgrading the ECL is deemed appropriate, the new ECL would then be based on a lower applicable IC(s) and EAL(s). The ECL may also be terminated.

As noted above, guidance concerning classification of rapidly escalating events or conditions is provided in RIS 2007-02 (ref. 4.1.2).

3.2.5 Classification of Short-Lived Events Event-based ICs and EALs define a variety of specific occurrences that have potential or actual safety significance. By their nature, some of these events may be short-lived and, thus, over before the emergency classification assessment can be completed. If an event occurs that meets or exceeds an EAL, the associated ECL must be declared regardless of its continued presence at the time of declaration. Examples of such events include an earthquake or a failure of the reactor protection system to automatically trip the reactor followed by a successful manual trip.

3.2.6 Classification of Transient Conditions Many of the ICs and/or EALs employ time-based criteria. These criteria will require that the IC/EAL conditions be present for a defined period of time before an emergency declaration is warranted. In cases where no time-based criterion is specified, it is recognized that some transient conditions may cause an EAL to be met for a brief period of time (e.g., a few seconds to a few minutes). The following guidance should be applied to the classification of these conditions.

EAL momentarily met during expected plant response - In instances where an EAL is briefly met during an expected (normal) plant response, an emergency declaration is not warranted provided that associated systems and components are operating as expected, and operator actions are performed in accordance with procedures.

EAL momentarily met but the condition is corrected prior to an emergency declaration - If an operator takes prompt manual action to address a condition, and the action is successful in correcting the condition prior to the emergency declaration, then the applicable EAL is not considered met and the associated emergency declaration is not required.

It is important to stress that the 15-minute emergency classification assessment period (process clock) is not a grace period during which a classification may be delayed to allow the performance of a corrective action that would obviate the need to classify the event.

Emergency classification assessments must be deliberate and timely, with no undue delays.

The provision discussed above addresses only those rapidly evolving situations when an operator is able to take a successful corrective action prior to the Emergency Director completing the review and steps necessary to make the emergency declaration. This provision is included to ensure that any public protective actions resulting from the emergency classification are truly warranted by the plant conditions.

EAL BASES DOCUMENT Rev. 0 Page 15 of 258

3.2.7 After-the-Fact Discovery of an Emergency Event or Condition In some cases, an EAL may be met but the emergency classification was not made at the time of the event or condition. This situation can occur when personnel discover that an event or condition existed which met an EAL, but no emergency was declared, and the event or condition no longer exists at the time of discovery. This may be due to the event or condition not being recognized at the time or an error that was made in the emergency classification process.

In these cases, no emergency declaration is warranted; however, the guidance contained in NUREG-1022 (ref. 4.1.3) is applicable. Specifically, the event should be reported to the NRC in accordance with 10 CFR § 50.72 (ref. 4.1.4) within one hour of the discovery of the undeclared event or condition. The licensee should also notify appropriate State and local agencies in accordance with the agreed upon arrangements.

3.2.8 Retraction of an Emergency Declaration Guidance on the retraction of an emergency declaration reported to the NRC is discussed in NUREG-1022 (ref. 4.1.3).

EAL BASES DOCUMENT Rev. 0 Page 16 of 258

4.0 REFERENCES

4.1 Developmental 4.1.1 NEI 99-01 Revision 6, Methodology for the Development of Emergency Action Levels for Non-Passive Reactors, ADAMS Accession Number ML13091A209 4.1.2 RIS 2007-02 Clarification of NRC Guidance for Emergency Notifications During Quickly Changing Events, February 2, 2007 4.1.3 NUREG-1022 Event Reporting Guidelines: 10CFR50.72 and 50.73 4.1.4 10 § CFR 50.72 Immediate Notification Requirements for Operating Nuclear Power Reactors 4.1.5 10 § CFR 50.73 License Event Report System 4.1.6 DBNPS ODCM 4.1.7 DBNPS UFSAR Figure 1.2-12 Site Plan 4.1.8 Technical Specifications Table 1.1-1 Modes 4.1.9 DB-OP-06904 Shutdown Operations 4.1.10 NG-QS-00121 DBNPS Writers Guide 4.1.11 NSIR/DPR-ISG-01 Interim Staff Guidance, Emergency Planning for Nuclear Power Plants 4.1.12 Site CSAR NUH-003 4.1.13 DBNPS Emergency Plan 4.1.14 DB-OP-02000 RPS, SFAS, SFRCS Trip, or SG Tube Rupture 4.2 Implementing 4.2.1 RA-EP-01500 Emergency Classification 4.2.2 NEI 99-01 Rev. 6 to DBNPS EAL Comparison Matrix 4.2.3 DBNPS EAL Matrix EAL BASES DOCUMENT Rev. 0 Page 17 of 258

5.0 DEFINITIONS, ACRONYMS & ABBREVIATIONS 5.1 Definitions (ref. 4.1.1 except as noted)

Selected terms used in Initiating Condition and Emergency Action Level statements are set in ALL capital letters (e.g., ALL CAPS). These words are defined terms that have specific meanings as used in this document. The definitions of these terms are provided below.

Alert Events are in progress or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant or a security event that involves probable life 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.

Confinement Boundary The barrier(s) between spent fuel and the environment once the spent fuel is processed for dry storage. As related to the DBNPS Dry Fuel Storage Facility, CONFINEMENT BOUNDARY is defined as the Dry Shielded Canister (DSC) (ref. 4.1.12).

Containment Closure The procedurally defined actions taken to secure containment and its associated structures, systems, and components as a functional barrier to fission product release under existing plant conditions (ref. 4.1.9).

Emergency Action Level (EAL)

A pre-determined, site-specific, observable threshold for an Initiating Condition that, when met or exceeded, places the plant in a given emergency classification level.

Emergency Classification Level (ECL)

One of a set of names or titles established by the US Nuclear Regulatory Commission (NRC) for grouping off-normal events or conditions according to (1) potential or actual effects or consequences, and (2) resulting onsite and offsite response actions. The emergency classification levels, in ascending order of severity, are:

Unusual Event (UE)

Alert (A)

Site Area Emergency (SAE)

General Emergency (GE)

EPA PAGs Environment Protection Agency Protective Action Guidelines. The EPA PAGs are expressed in terms of dose commitment: 1 Rem TEDE or 5 Rem CDE Thyroid. Actual or projected offsite exposures in excess of the EPA PAGs requires DBNPS to recommend protective actions for the general public to offsite planning agencies.

EAL BASES DOCUMENT Rev. 0 Page 18 of 258

Explosion A rapid, violent and catastrophic failure of a piece of equipment due to combustion, chemical reaction or overpressurization. A release of steam (from high energy lines or components) or an electrical component failure (caused by short circuits, grounding, arcing, etc.) should not automatically be considered an EXPLOSION. Such events may require a post-event inspection to determine if the attributes of an EXPLOSION are present.

Faulted The term applied to a steam generator that has a steam leak on the secondary side of sufficient size to cause an uncontrolled drop in steam generator pressure or the steam generator to become completely depressurized.

Fire Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute FIRES. Observation of flame is preferred but is NOT required if large quantities of smoke and heat are observed.

Fission Product Barrier Threshold A pre-determined, site-specific, observable threshold indicating the loss or potential loss of a fission product barrier.

Flooding A condition where water is entering a room or area faster than installed equipment is capable of removal, resulting in a rise of water level within the room or area.

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

Hostage A person(s) held as leverage against the station to ensure that demands will be met by the station.

Hostile Action An act toward DBNPS 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 DBNPS. Non-terrorism-based EALs should be EAL BASES DOCUMENT Rev. 0 Page 19 of 258

used to address such activities (i.e., this may include violent acts between individuals in the OWNER CONTROLLED AREA).

Hostile Force One or more individuals who are engaged in a determined assault, overtly or by stealth and deception, equipped with suitable weapons capable of killing, maiming, or causing destruction.

Imminent The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions.

Impede(d)

Personnel access to a room or area is hindered to an extent that extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., requiring use of protective equipment, such as SCBAs, that is not routinely employed).

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.

Initiating Condition (IC)

An event or condition that aligns with the definition of one of the four emergency classification levels by virtue of the potential or actual effects or consequences.

Maintain Take appropriate action to hold the value of an identified parameter within specified limits.

Owner Controlled Area The property associated with the station and owned by the company. Access is normally limited to persons entering for official business (ref. 4.1.13).

Projectile An object directed toward a Nuclear Power Plant that could cause concern for its continued operability, reliability, or personnel safety.

Protected Area An area that normally encompasses all controlled areas within the security protected area fence.

EAL BASES DOCUMENT Rev. 0 Page 20 of 258

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).

Refueling Pathway The reactor refueling canal, spent fuel pool and fuel transfer canal comprise the REFUELING PATHWAY.

Restore Take the appropriate action required to return the value of an identified parameter to the applicable limits Ruptured The condition of a steam generator in which primary-to-secondary leakage is of sufficient magnitude to require a safety injection (ref. 4.1.14).

Safety System A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2):

Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:

(1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.

Security Condition ANY Security Event as listed in the approved security contingency plan that constitutes a threat/compromise to site security, threat/risk to site personnel, or a potential degradation to the level of safety of the plant. A SECURITY CONDITION does not involve a HOSTILE ACTION.

Site Area Emergency Events are in progress or have occurred which involve actual or likely major failures of plant functions needed for protection of the public or HOSTILE 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.

EAL BASES DOCUMENT Rev. 0 Page 21 of 258

Site Boundary Area as depicted in UFSAR Figure 1.2-12 Site Plan (ref. 4.1.7). The SITE BOUNDARY is defined at a minimum exclusion distance of 0.75 miles. This is the nearest distance from potential release points at which protective actions would be required for members of the public.

Unisolable An open or breached system line that cannot be isolated, remotely or locally.

Unplanned A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.

Unusual Event Events are in progress or have occurred which indicate a potential degradation of the level of safety of the plant or indicate a security threat to facility protection has been initiated. No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of SAFETY SYSTEMS occurs.

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 indicators operability, the conditions existence, or the reports accuracy is removed. Implicit in this definition is the need for timely assessment.

Visible Damage Damage to a component or structure that is readily observable without measurements, testing, or analysis. The visual impact of the damage is sufficient to cause concern regarding the operability or reliability of the affected component or structure.

EAL BASES DOCUMENT Rev. 0 Page 22 of 258

5.2 Abbreviations/Acronyms

°F ....................................................................................................... Degrees Fahrenheit

° ........................................................................................................................... Degrees AC ....................................................................................................... Alternating Current AOP ................................................................................. Abnormal Operating Procedure ATWS ...................................................................... Anticipated Transient Without Scram B&W ..................................................................................................... Babcock & Wilcox BWST .................................................................................. Borated Water Storage Tank DBNPS ..................................................................... Davis-Besse Nuclear Power Station CCW ........................................................................................ Component Cooling Water CDE ....................................................................................... Committed Dose Equivalent CFR ..................................................................................... Code of Federal Regulations DBA ............................................................................................... Design Basis Accident DC ...............................................................................................................Direct Current DFSF .......................................................................................... Dry Fuel Storage Facility DHR ................................................................................................. Decay Heat Removal DSC ................................................................................................ Dry Shielded Canister 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 EPIP ................................................................ Emergency Plan Implementing Procedure ERG ................................................................................ Emergency Response Guideline ESF......................................................................................... Engineered Safety Feature FAA.................................................................................. Federal Aviation Administration FBI ................................................................................... Federal Bureau of Investigation FEMA............................................................... Federal Emergency Management Agency FSAR .................................................................................... Final Safety Analysis Report GE ..................................................................................................... General Emergency IC ..........................................................................................................Initiating Condition IPEEE ................. Individual Plant Examination of External Events (Generic Letter 88-20)

ISFSI............................................................ Independent Spent Fuel Storage Installation EAL BASES DOCUMENT Rev. 0 Page 23 of 258

Keff ......................................................................... Effective Neutron Multiplication Factor LCO .................................................................................. Limiting Condition of Operation LER................................................................................................ Licensee Event Report LOCA ......................................................................................... Loss of Coolant Accident MSIV ....................................................................................... Main Steam Isolation Valve MSL ........................................................................................................ Main Steam Line MSSV ........................................................................................ Main Steam Safety Valve mR, mRem, mrem, mREM .............................................. milli-Roentgen Equivalent Man MU-HPI .................................................................... Makeup and High Pressure Injection MW .................................................................................................................... Megawatt NEI .............................................................................................. Nuclear Energy Institute NESP ................................................................... National Environmental Studies Project NRC ................................................................................ Nuclear Regulatory Commission NSSS ................................................................................ Nuclear Steam Supply System NORAD................................................... North American Aerospace Defense Command OBE ...................................................................................... Operating Basis Earthquake OCA ...............................................................................................Owner Controlled Area ODCM............................................................................ Off-site Dose Calculation Manual ORO ................................................................................. Offsite Response Organization PA .............................................................................................................. Protected Area PAG ........................................................................................ Protective Action Guideline PORV .................................................................................. Power Operated Relief Valve PRA/PSA ..................... Probabilistic Risk Assessment / Probabilistic Safety Assessment PWR ....................................................................................... Pressurized Water Reactor PSIG ................................................................................ Pounds per Square Inch Gauge R ........................................................................................................................ Roentgen RCDT..................................................................................... Reactor Coolant Drain Tank RCS ............................................................................................ Reactor Coolant System Rem, rem, REM ....................................................................... Roentgen Equivalent Man RPS ........................................................................................ Reactor Protection System RV .............................................................................................................Reactor Vessel SAR ............................................................................................... Safety Analysis Report SBO ......................................................................................................... Station Blackout SBODG........................................................................ Station Blackout Diesel Generator EAL BASES DOCUMENT Rev. 0 Page 24 of 258

SCBA ....................................................................... Self-Contained Breathing Apparatus SG ......................................................................................................... Steam Generator SI .............................................................................................................. Safety Injection SPDS ........................................................................... Safety Parameter Display System SRO ............................................................................................ Senior Reactor Operator TEDE ............................................................................... Total Effective Dose Equivalent TOAF .................................................................................................... Top of Active Fuel TSC .......................................................................................... Technical Support Center TRM ................................................................................ Technical Requirements Manual EAL BASES DOCUMENT Rev. 0 Page 25 of 258

6.0 DBNPS-TO-NEI 99-01 Rev. 6 EAL CROSS-REFERENCE This cross-reference is provided to facilitate association and location of a DBNPS EAL within the NEI 99-01 IC/EAL identification scheme. Further information regarding the development of the DBNPS EALs based on the NEI guidance can be found in the EAL Comparison Matrix.

DBNPS NEI 99-01 Rev. 6 Example EAL IC EAL HU1.1 HU1 1 HU1.2 HU1 2 HU1.3 HU1 3 HU2.1 HU2 1 HU3.1 HU3 1 HU3.2 HU3 2 HU3.3 HU3 3 HU3.4 HU3 4 HU4.1 HU4 1 HU4.2 HU4 2 HU4.3 HU4 3 HU4.4 HU4 4 HU7.1 HU7 1 HA1.1 HA1 1 HA1.2 HA1 2 HA5.1 HA5 1 HA6.1 HA6 1 HA7.1 HA7 1 HS1.1 HS1 1 HS6.1 HS6 1 HS7.1 HS7 1 EAL BASES DOCUMENT Rev. 0 Page 26 of 258

DBNPS NEI 99-01 Rev. 6 Example EAL IC EAL HG7.1 HG7 1 RU1.1 AU1 1, 2 RU1.2 AU1 3 RU2.1 AU2 1 RA1.1 AA1 1 RA1.2 AA1 2 RA1.3 AA1 3 RA1.4 AA1 4 RA2.1 AA2 1 RA2.2 AA2 2 RA2.3 AA2 3 RA3.1 AA3 1 RA3.2 AA3 2 RS1.1 AS1 1 RS1.2 AS1 2 RS1.3 AS1 3 RS2.1 AS2 1 RG1.1 AG1 1 RG1.2 AG1 2 RG1.3 AG1 3 RG2.1 AG2 1 EU1.1 E-HU1 1 CU1.1 CU1 1 CU1.2 CU1 2 CU2.1 CU2 1 EAL BASES DOCUMENT Rev. 0 Page 27 of 258

DBNPS NEI 99-01 Rev. 6 Example EAL IC EAL CU3.1 CU3 1 CU3.2 CU3 2 CU4.1 CU4 1 CU5.1 CU5 1 CU5.2 CU5 2 CU5.3 CU5 3 CA1.1 CA1 1 CA1.2 CA1 2 CA2.1 CA2 1 CA3.1 CA3 1, 2 CA6.1 CA6 1 CS1.1 CS1 3 CG1.1 CG1 2 SU1.1 SU1 1 SU3.1 SU2 1 SU4.1 SU3 2 SU5.1 SU4 1, 2, 3 SU6.1 SU5 1 SU6.2 SU5 2 SU7.1 SU6 1 SU7.2 SU6 2 SU7.3 SU6 3 SU8.1 SU7 1 SU8.2 SU7 2 SA1.1 SA1 1 EAL BASES DOCUMENT Rev. 0 Page 28 of 258

DBNPS NEI 99-01 Rev. 6 Example EAL IC EAL SA3.1 SA2 1 SA6.1 SA5 1 SA9.1 SA9 1 SS1.1 SS1 1 SS2.1 SS8 1 SS6.1 SS5 1 SG1.1 SG1 1 SG1.2 SG8 1 FA1.1 FA1 1 FS1.1 FS1 1 FG1.1 FG1 1 EAL BASES DOCUMENT Rev. 0 Page 29 of 258

7.0 ATTACHMENTS Attachment 1, Emergency Action Level Technical Bases Attachment 2, Fission Product Barrier Matrix and Basis Attachment 3, Safe Shutdown Rooms/Areas Tables R-2 & H-2 Bases EAL BASES DOCUMENT Rev. 0 Page 30 of 258

ATTACHMENT 1 EAL Bases Category H - Hazards and Other Conditions Affecting Plant Safety EAL Group: ANY (EALs in this category are applicable to ANY plant condition, hot or cold.)

Hazards are non-plant, system-related events that can directly or indirectly affect plant operation, reactor plant safety or personnel safety.

1. Security Unauthorized entry attempts into the PROTECTED AREA, bomb threats, sabotage attempts, and actual security compromises threatening loss of physical control of the plant.

2. Seismic Event Natural events such as earthquakes have potential to cause plant structure or equipment damage of sufficient magnitude to threaten personnel or plant safety.

3. Natural or Technology Hazard Other natural and non-naturally occurring events that can cause damage to plant facilities include tornados, FLOODING, hazardous material releases and events restricting site access warranting classification.

4. Fire FIREs can pose significant hazards to personnel and reactor safety. Appropriate for classification are FIREs within the site PROTECTED AREA or FIREs that may affect operability of equipment needed for safe shutdown.

5. Hazardous Gas Toxic, corrosive, asphyxiant or flammable gas leaks can affect normal plant operations or preclude access to plant areas required to safely shutdown the plant.

6. Control Room Evacuation Events that are indicative of loss of Control Room habitability. If the Control Room must be evacuated, additional support for monitoring and controlling plant functions is necessary through the emergency response facilities.

7. Emergency Director 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 that 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 EAL BASES DOCUMENT Rev. 0 Page 31 of 258

ATTACHMENT 1 EAL Bases category provide the Emergency Director the latitude to classify emergency conditions consistent with the established classification criteria based upon Emergency Director judgment.

EAL BASES DOCUMENT Rev. 0 Page 32 of 258

ATTACHMENT 1 EAL Bases Category: H - Hazards Subcategory: 1 - Security Initiating Condition: HOSTILE ACTION within the PROTECTED AREA EAL:

HS1.1 Site Area Emergency A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by the Security Shift Supervisor Mode Applicability:

ALL Basis:

This IC addresses the occurrence of a HOSTILE ACTION within the PROTECTED AREA.

This event will require rapid response and assistance due to the possibility for damage to plant equipment.

Timely and accurate communications between Security Shift Supervision and the Control Room is essential for proper classification of a security-related event (ref. 2, 3).

Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan and Independent Spent Fuel Storage Installation Security Program.

As time and conditions allow, these events require a heightened state of readiness by the plant staff and implementation of onsite protective measures (e.g., evacuation, dispersal or sheltering). The Site Area Emergency declaration will mobilize Offsite Response Organization (ORO) resources and have them available to develop and implement public protective actions in the unlikely event that the attack is successful in impairing multiple safety functions.

This IC does not apply to incidents that are accidental events, acts of civil disobedience, or otherwise are not a HOSTILE ACTION perpetrated by a HOSTILE FORCE. Examples include the crash of a small aircraft, shots from hunters, physical disputes between employees, etc.

Reporting of these types of events is adequately addressed by other EALs, or the requirements of 10 CFR § 73.71 or 10 CFR § 50.72.

Emergency plans and implementing procedures are public documents; therefore, EALs should not incorporate Security-sensitive information. This includes information that may be advantageous to a potential adversary, such as the particulars concerning a specific threat or threat location. Security-sensitive information should be contained in non-public documents such as the DBNPS Physical Security Plan (ref. 1).

EAL BASES DOCUMENT Rev. 0 Page 33 of 258

ATTACHMENT 1 EAL Bases Escalation of the emergency classification level would be via IC FG1.

The Security Shift Supervision is defined as the Security Shift Supervisor or designee.

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 is highly controlled due to the strict secrecy controls placed on the DBNPS Physical Security Plan (Safeguards) information (ref. 1).

Basis Reference(s):

1. DBNPS Physical Security Plan (safeguards)

2. DB-OP-02544 Security Events or Threats (restricted)

3. RA-EP-02890 Emergency Response Organization Response to Security Events or Threats

4. NEI 99-01 HS1 EAL BASES DOCUMENT Rev. 0 Page 34 of 258

ATTACHMENT 1 EAL Bases Category: H - Hazards Subcategory: 1 - Security Initiating Condition: HOSTILE ACTION within the OWNER CONTROLLED AREA or airborne attack threat within 30 minutes EAL:

HA1.1 Alert A HOSTILE ACTION is occurring or has occurred within the OWNER CONTROLLED AREA as reported by the Security Shift Supervisor Mode Applicability:

ALL Basis:

This IC addresses the occurrence of a HOSTILE ACTION within the OWNER CONTROLLED AREA. This event will require rapid response and assistance due to the possibility of the attack progressing to the PROTECTED AREA.

Timely and accurate communications between Security Shift Supervision and the Control Room is essential for proper classification of a security-related event (ref. 2, 3).

Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan and Independent Spent Fuel Storage Installation Security Program.

As time and conditions allow, these events require a heightened state of readiness by the plant staff and implementation of onsite protective measures (e.g., evacuation, dispersal or sheltering). The Alert declaration will also heighten the awareness of Offsite Response Organizations (OROs), allowing them to be better prepared should it be necessary to consider further actions.

This IC does not apply to incidents that are accidental events, acts of civil disobedience, or otherwise are not a HOSTILE ACTION perpetrated by a HOSTILE FORCE. Examples include the crash of a small aircraft, shots from hunters, physical disputes between employees, etc.

Reporting of these types of events is adequately addressed by other EALs, or the requirements of 10 CFR § 73.71 or 10 CFR § 50.72.

This EAL is applicable for ANY HOSTILE ACTION occurring, or that has occurred, in the OWNER CONTROLLED AREA. This includes any action directed against an ISFSI that is located outside the plant PROTECTED AREA.

EAL BASES DOCUMENT Rev. 0 Page 35 of 258

ATTACHMENT 1 EAL Bases Emergency plans and implementing procedures are public documents; therefore, EALs should not incorporate Security-sensitive information. This includes information that may be advantageous to a potential adversary, such as the particulars concerning a specific threat or threat location. Security-sensitive information should be contained in non-public documents such as the DBNPS Physical Security Plan (ref. 1).

The Security Shift Supervision is defined as the Security Shift Supervisor or designee.

Escalation of the emergency classification level would be via IC HS1.

Basis Reference(s):

1. DBNPS Physical Security Plan (safeguards)

2. DB-OP-02544 Security Events or Threats (restricted)

3. RA-EP-02890 Emergency Response Organization Response to Security Events or Threats

4. NEI 99-01 HA1 EAL BASES DOCUMENT Rev. 0 Page 36 of 258

ATTACHMENT 1 EAL Bases Category: H - Hazards Subcategory: 1 - Security Initiating Condition: HOSTILE ACTION within the OWNER CONTROLLED AREA or airborne attack threat within 30 minutes EAL:

HA1.2 Alert A validated notification from NRC of an aircraft attack threat within 30 min. of the site Mode Applicability:

ALL Basis:

This IC addresses the notification of an aircraft attack threat. This event will require rapid response and assistance due to the need to prepare the plant and staff for a potential aircraft impact.

Timely and accurate communications between Security Shift Supervision and the Control Room is essential for proper classification of a security-related event (ref. 2, 3).

Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan and Independent Spent Fuel Storage Installation Security Program.

As time and conditions allow, these events require a heightened state of readiness by the plant staff and implementation of onsite protective measures (e.g., evacuation, dispersal or sheltering). The Alert declaration will also heighten the awareness of Offsite Response Organizations (OROs), allowing them to be better prepared should it be necessary to consider further actions.

This IC does not apply to incidents that are accidental events, acts of civil disobedience, or otherwise are not a HOSTILE ACTION perpetrated by a HOSTILE FORCE. Examples include the crash of a small aircraft, shots from hunters, physical disputes between employees, etc.

Reporting of these types of events is adequately addressed by other EALs, or the requirements of 10 CFR § 73.71 or 10 CFR § 50.72.

This EAL addresses the threat from the impact of an aircraft on the plant, and the anticipated arrival time is within 30 minutes. The intent of this EAL is to ensure that threat-related notifications are made in a timely manner so that plant personnel and OROs are in a heightened state of readiness. This EAL is met when the threat-related information has been validated in accordance with site-specific security procedures.

EAL BASES DOCUMENT Rev. 0 Page 37 of 258

ATTACHMENT 1 EAL Bases The NRC Headquarters Operations Officer (HOO) will communicate to the licensee if the threat involves an aircraft. The status and size of the plane may be provided by NORAD through the NRC.

In some cases, it may not be readily apparent if an aircraft impact within the OWNER CONTROLLED AREA was intentional (i.e., a HOSTILE ACTION). It is expected, although not certain, that notification by an appropriate Federal agency to the site would clarify this point. In this case, the appropriate federal agency is intended to be NORAD, FBI, FAA or NRC. The emergency declaration, including one based on other ICs/EALs, should not be unduly delayed while awaiting notification by a Federal agency.

Emergency plans and implementing procedures are public documents; therefore, EALs should not incorporate Security-sensitive information. This includes information that may be advantageous to a potential adversary, such as the particulars concerning a specific threat or threat location. Security-sensitive information should be contained in non-public documents such as the DBNPS Physical Security Plan (ref. 1).

The Security Shift Supervision is defined as the Security Shift Supervisor or designee.

Escalation of the emergency classification level would be via IC HS1.

Basis Reference(s):

1. DBNPS Physical Security Plan (safeguards)

2. DB-OP-02544 Security Events or Threats (restricted)

3. RA-EP-02890 Emergency Response Organization Response to Security Events or Threats

4. NEI 99-01 HA1 EAL BASES DOCUMENT Rev. 0 Page 38 of 258

ATTACHMENT 1 EAL Bases Category: H - Hazards Subcategory: 1 - Security Initiating Condition: Confirmed SECURITY CONDITION or threat EAL:

HU1.1 Unusual Event A SECURITY CONDITION that does not involve a HOSTILE ACTION as reported by the Security Shift Supervisor Mode Applicability:

ALL Basis:

This IC addresses events that pose a threat to plant personnel or SAFETY SYSTEM equipment, and thus represent a potential degradation in the level of plant safety. Security events which do not meet one of these EALs are adequately addressed by the requirements of 10 CFR § 73.71 or 10 CFR § 50.72. Security events assessed as HOSTILE ACTIONS are classifiable under ICs HA1, and HS1.

Timely and accurate communications between Security Shift Supervision and the Control Room is essential for proper classification of a security-related event (ref. 2, 3, 4).

Classification of these events will initiate appropriate threat-related notifications to plant personnel and Offsite Response Organizations.

Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan and Independent Spent Fuel Storage Installation Security Program.

This EAL references the Security Shift Supervisor because these are the individuals trained to confirm that a security event is occurring or has occurred. Training on security event confirmation and classification is controlled due to the nature of Safeguards and 10 CFR

§ 2.39 information.

Emergency plans and implementing procedures are public documents; therefore, EALs should not incorporate Security-sensitive information. This includes information that may be advantageous to a potential adversary, such as the particulars concerning a specific threat or threat location. Security-sensitive information should be contained in non-public documents such as the DBNPS Physical Security Plan (ref. 1).

Escalation of the emergency classification level would be via IC HA1.

EAL BASES DOCUMENT Rev. 0 Page 39 of 258

ATTACHMENT 1 EAL Bases The Security Shift Supervision is defined as the Security Shift Supervisor or designee.

This EAL is based on the DBNPS Physical Security Plan (ref. 1).

Basis Reference(s):

1. DBNPS Physical Security Plan (safeguards)

2. DB-OP-02544 Security Events or Threats (restricted)

3. RA-EP-02890 Emergency Response Organization Response to Security Events or Threats

4. NEI 99-01 HU1 EAL BASES DOCUMENT Rev. 0 Page 40 of 258

ATTACHMENT 1 EAL Bases Category: H - Hazards Subcategory: 1 - Security Initiating Condition: Confirmed SECURITY CONDITION or threat EAL:

HU1.2 Unusual Event Notification of a credible security threat directed at the site Mode Applicability:

ALL Basis:

This IC addresses events that pose a threat to plant personnel or SAFETY SYSTEM equipment, and thus represent a potential degradation in the level of plant safety. Security events which do not meet one of these EALs are adequately addressed by the requirements of 10 CFR § 73.71 or 10 CFR § 50.72. Security events assessed as HOSTILE ACTIONS are classifiable under ICs HA1, and HS1.

Timely and accurate communications between Security Shift Supervision and the Control Room is essential for proper classification of a security-related event (ref. 2, 3, 4).

Classification of these events will initiate appropriate threat-related notifications to plant personnel and Offsite Response Organizations.

Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan and Independent Spent Fuel Storage Installation Security Program.

This EAL addresses the receipt of a credible security threat. The credibility of the threat is assessed in accordance with the DBNPS Physical Security Plan (ref. 1).

Emergency plans and implementing procedures are public documents; therefore, EALs should not incorporate Security-sensitive information. This includes information that may be advantageous to a potential adversary, such as the particulars concerning a specific threat or threat location. Security-sensitive information should be contained in non-public documents such as the DBNPS Physical Security Plan (ref. 1).

Escalation of the emergency classification level would be via IC HA1.

The Security Shift Supervision is defined as the Security Shift Supervisor or designee.

This EAL is based on the DBNPS Physical Security Plan (ref. 1).

EAL BASES DOCUMENT Rev. 0 Page 41 of 258

ATTACHMENT 1 EAL Bases Basis Reference(s):

1. DBNPS Physical Security Plan (safeguards)

2. DB-OP-02544 Security Events or Threats (restricted)

3. RA-EP-02890 Emergency Response Organization Response to Security Events or Threats

4. NEI 99-01 HU1 EAL BASES DOCUMENT Rev. 0 Page 42 of 258

ATTACHMENT 1 EAL Bases Category: H - Hazards Subcategory: 1 - Security Initiating Condition: Confirmed SECURITY CONDITION or threat EAL:

HU1.3 Unusual Event A validated notification from the NRC providing information of an aircraft threat Mode Applicability:

ALL Basis:

This IC addresses events that pose a threat to plant personnel or SAFETY SYSTEM equipment, and thus represent a potential degradation in the level of plant safety. Security events which do not meet one of these EALs are adequately addressed by the requirements of 10 CFR § 73.71 or 10 CFR § 50.72. Security events assessed as HOSTILE ACTIONS are classifiable under ICs HA1 and HS1.

Timely and accurate communications between Security Shift Supervision and the Control Room is essential for proper classification of a security-related event (ref. 2, 3, 4).

Classification of these events will initiate appropriate threat-related notifications to plant personnel and Offsite Response Organizations.

Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan and Independent Spent Fuel Storage Installation Security Program.

This EAL addresses the threat from the impact of an aircraft on the plant. The NRC Headquarters Operations Officer (HOO) will communicate to the licensee if the threat involves an aircraft. The status and size of the plane may also be provided by NORAD through the NRC. Validation of the threat is performed in accordance with the DBNPS Physical Security Plan (ref. 1).

Emergency plans and implementing procedures are public documents; therefore, EALs should not incorporate Security-sensitive information. This includes information that may be advantageous to a potential adversary, such as the particulars concerning a specific threat or threat location. Security-sensitive information should be contained in non-public documents such as the DBNPS Physical Security Plan (ref. 1).

Escalation of the emergency classification level would be via IC HA1.

EAL BASES DOCUMENT Rev. 0 Page 43 of 258

ATTACHMENT 1 EAL Bases The Security Shift Supervision is defined as the Security Shift Supervisor or designee.

This EAL is based on the DBNPS Physical Security Plan (ref. 1).

Basis Reference(s):

1. DBNPS Physical Security Plan (safeguards)

2. DB-OP-02544 Security Events or Threats (restricted)

3. RA-EP-02890 Emergency Response Organization Response to Security Events or Threats

4. NEI 99-01 HU1 EAL BASES DOCUMENT Rev. 0 Page 44 of 258

ATTACHMENT 1 EAL Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 2 - Seismic Event Initiating Condition: Seismic event greater than OBE levels EAL:

HU2.1 Unusual Event Seismic event > OBE as indicated by OBE alarm on seismic alarm panel C5764A Mode Applicability:

ALL Basis:

This IC addresses a seismic event that results in accelerations at the plant site greater than those specified for an Operating Basis Earthquake (OBE). An earthquake greater than an OBE but less than a Safe Shutdown Earthquake (SSE) should have no significant impact on safety-related systems, structures and components; however, some time may be required for the plant staff to ascertain the actual post-event condition of the plant (e.g., performs walk-downs and post-event inspections). Given the time necessary to perform walk-downs and inspections, and fully understand any impacts, this event represents a potential degradation of the level of safety of the plant.

Event verification with external sources should not be necessary during or following an OBE.

Earthquakes of this magnitude should be readily felt by on-site personnel and recognized as a seismic event (e.g., lateral accelerations in excess of 0.08g). The Shift Manager or Emergency Director may seek external verification if deemed appropriate (e.g., a call to the USGS, check internet news sources, etc.); however, the verification action must not preclude a timely emergency declaration.

Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via IC CA6 or SA9.

Ground motion acceleration of 0.08g horizontal or 0.053g vertical is the Maximum Probable Earthquake as is considered generically as the Operating Basis Earthquake for DBNPS (ref. 1, 2).

Four triaxial strong motion accelerometers are installed within plant structures and networked together, such that any one accelerometer can also trigger the other devices and start event recording. Data is retrievable from each accelerometer (via an analog-to-digital recorder and data storage memory) through a personal computer in the Control Room. Time-history data and response spectra plots can be readily reviewed on a video display without need for conversion to hardcopy. Quick and accurate determinations can be made based on the seismic event. Peak acceleration values are displayed showing whether the Operating Basis Earthquake (OBE) limits have or have not been exceeded. Additionally, immediate Control EAL BASES DOCUMENT Rev. 0 Page 45 of 258

ATTACHMENT 1 EAL Bases Room alarm indication of an earthquake of 0.08 g horizontal or 0.053g vertical is annunciated on the seismic control panel (C5764A), as sensed by the Containment Concrete Foundation accelerometer (ref. 2).

RA-EP-02820 Earthquake provides the guidance for determining any required response actions if the OBE earthquake threshold is exceeded (ref. 3).

To avoid inappropriate emergency classification resulting from spurious actuation of the seismic instrumentation or felt motion not attributable to seismic activity, an offsite agency (USGS, National Earthquake Information Center) can confirm that an earthquake has occurred in the area of the plant. However, such confirmation should not preclude a timely emergency declaration based on receipt of the OBE alarm. Contact the NEIC by calling (303) 273-8500.

Select option #1, then #2 and inform the analyst you wish to confirm recent seismic activity near DBNPS. Provide the analyst with the following DBNPS coordinates: 41º 35' 49" north latitude, 83º 05' 16" west longitude and exact time of seismic activity (ref. 4). Alternatively, access near real-time seismic activity via the NEIC website:

http://earthquake.usgs.gov/eqcenter/

Basis Reference(s):

1. Updated FSAR Section 3.2.1 Seismic Classification

2. Updated FSAR Section 3.7.2 Seismic Design

3. RA-EP-02820 Earthquake

4. UFSAR Section 2.1.1 Site Location

5. NEI 99-01 HU2 EAL BASES DOCUMENT Rev. 0 Page 46 of 258

ATTACHMENT 1 EAL Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 3 - Natural or Technology Hazard Initiating Condition: Hazardous event EAL:

HU3.1 Unusual Event A tornado strike within the PROTECTED AREA Mode Applicability:

ALL Basis:

This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant.

EAL HU3.1 addresses a tornado striking (touching down) within the PROTECTED AREA.

Escalation of the emergency classification level would be based on ICs in Recognition Categories R, F, S or C.

Response actions associated with a tornado onsite is provided in RA-EP-02810 Tornado or High Winds (ref. 1).

If damage is confirmed visually or by other in-plant indications, the event may be escalated to an Alert under EAL CA6.1 or SA9.1.

A tornado striking (touching down) within the PROTECTED AREA warrants declaration of an Unusual Event regardless of the measured wind speed at the meteorological tower.

Basis Reference(s):

1. RA-EP-02810 Tornado or High Winds

2. NEI 99-01 HU3 EAL BASES DOCUMENT Rev. 0 Page 47 of 258

ATTACHMENT 1 EAL Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 3 - Natural or Technology Hazard Initiating Condition: Hazardous event EAL:

HU3.2 Unusual Event Internal room or area FLOODING of a magnitude sufficient to require manual or automatic electrical isolation of a SAFETY SYSTEM component needed for the current operating mode Mode Applicability:

ALL Basis:

This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant.

This EAL addresses FLOODING of a building room or area that results in operators isolating power to a SAFETY SYSTEM component due to water level or other wetting concerns.

Classification is also required if the water level or related wetting causes an automatic isolation of a SAFETY SYSTEM component from its power source (e.g., a breaker or relay trip). To warrant classification, operability of the affected component must be required by Technical Specifications for the current operating mode.

Escalation of the emergency classification level would be based on ICs in Recognition Categories R, F, S or C.

Areas containing safe shutdown equipment susceptible to internal FLOODING are Service Water Pump Room and adjacent areas, Component Cooling Water Pump Room and Emergency Core Cooling System Room(s) (ref.1). Refer to EAL CA6.1 or SA9.1 for internal FLOODING affecting one or more SAFETY SYSTEM trains.

Basis Reference(s):

1. RA-EP-02880 Internal Flooding

2. NEI 99-01 HU3 EAL BASES DOCUMENT Rev. 0 Page 48 of 258

ATTACHMENT 1 EAL Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 3 - Natural or Technology Hazard Initiating Condition: Hazardous event EAL:

HU3.3 Unusual Event Movement of personnel within the PROTECTED AREA is IMPEDED due to an offsite event involving hazardous materials (e.g., an offsite chemical spill or toxic gas release)

Mode Applicability:

ALL Basis:

This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant.

This EAL addresses a hazardous materials event originating at an offsite location and of sufficient magnitude to IMPEDE the movement of personnel within the PROTECTED AREA.

Escalation of the emergency classification level would be based on ICs in Recognition Categories R, F, S or C.

As used here, the term "offsite" means the areas external to the DBNPS PROTECTED AREA.

Basis Reference(s):

1. NEI 99-01 HU3 EAL BASES DOCUMENT Rev. 0 Page 49 of 258

ATTACHMENT 1 EAL Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 3 - Natural or Technology Hazard Initiating Condition: Hazardous event EAL:

HU3.4 Unusual Event A hazardous event that results in on-site conditions sufficient to prohibit the plant staff from accessing the site via personal vehicles (Note 7)

Note 7: This EAL does not apply to routine traffic impediments such as fog, snow, ice, or vehicle breakdowns or accidents.

Mode Applicability:

ALL Basis:

This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant.

This EAL addresses a hazardous event that causes an on-site impediment to vehicle movement and significant enough to prohibit the plant staff from accessing the site using personal vehicles. Examples of such an event include site FLOODING caused by a hurricane, heavy rains, up-river water releases, dam failure, etc., or an on-site train derailment blocking the access road.

This EAL is not intended to apply to routine impediments such as fog, snow, ice, or vehicle breakdowns or accidents, but rather to more significant conditions such as the Hurricane Andrew strike on Turkey Point in 1992, the flooding around the Cooper Station during the Midwest floods of 1993, or the flooding around Ft. Calhoun Station in 2011.

Escalation of the emergency classification level would be based on ICs in Recognition Categories R, F, S or C.

As used here, the term "onsite" means the areas within the DBNPS OWNER CONTROLLED AREA.

Basis Reference(s):

1. NEI 99-01 HU3 EAL BASES DOCUMENT Rev. 0 Page 50 of 258

ATTACHMENT 1 EAL Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 4 - FIRE Initiating Condition: FIRE potentially degrading the level of safety of the plant EAL:

HU4.1 Unusual Event A FIRE is not extinguished within 15 min. of ANY of the following FIRE detection indications (Note 1):

Report from the field (i.e., visual observation)

Receipt of multiple (more than 1) fire alarms or indications Field verification of a single fire alarm AND The FIRE is located within ANY Table H-1 area Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Table H-1 Safe Shutdown Fire Areas Containment Control Room Auxiliary Building Intake Structure Borated Water Storage Tank Mode Applicability:

ALL Basis:

This IC addresses the magnitude and extent of FIRES that may be indicative of a potential degradation of the level of safety of the plant.

For EAL HU4.1 the intent of the 15-minute duration is to size the FIRE and to discriminate against small FIRES that are readily extinguished (e.g., smoldering waste paper basket). In addition to alarms, other indications of a FIRE could be a drop in fire main pressure, automatic activation of a suppression system, etc.

EAL BASES DOCUMENT Rev. 0 Page 51 of 258

ATTACHMENT 1 EAL Bases Upon receipt, operators will take prompt actions to confirm the validity of an initial fire alarm, indication, or report. For EAL assessment purposes, the emergency declaration clock starts at the time that the initial alarm, indication, or report was received, and not the time that a subsequent verification action was performed. Similarly, the fire duration clock also starts at the time of receipt of the initial alarm, indication or report.

Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via IC CA6 or SA9.

Table H-1 Fire Areas are based on DBNPS Unit 1 Fire Hazard Analysis Report. Table H-1 Fire Areas include those structures containing functions and systems required for safe shutdown of the plant (SAFETY SYSTEMS) (ref. 1, 2).

For the purposes of declaring an emergency event, the term "extinguished" means no visible flames.

Basis Reference(s):

1. DBNPS Fire Hazard Analysis Report

2. DB-OP-02529 Fire Procedure

3. DBNPS-1465.00-00-0006 Design Basis Specification for the Plant Fire Protection

4. AP/0/A/5500/045 Plant Fire

5. Condition Report 09-66994, Emergency Action Level Clarification

4. Condition Report 09-69475, White Finding Identified For Inadequate Emergency Classification of Event

5. NEI 99-01 HU4 EAL BASES DOCUMENT Rev. 0 Page 52 of 258

ATTACHMENT 1 EAL Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 4 - FIRE Initiating Condition: FIRE potentially degrading the level of safety of the plant EAL:

HU4.2 Unusual Event Receipt of a single fire alarm (i.e., no other indications of a FIRE)

AND The fire alarm is indicating a FIRE within ANY Table H-1 area AND The existence of a FIRE is not verified within 30 min. of alarm receipt (Note 1)

Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Table H-1 Safe Shutdown Fire Areas Containment Control Room Auxiliary Building Intake Structure Borated Water Storage Tank Mode Applicability:

ALL Basis:

This IC addresses the magnitude and extent of FIRES that may be indicative of a potential degradation of the level of safety of the plant.

This EAL addresses receipt of a single fire alarm, and the existence of a FIRE is not verified (i.e., proved or disproved) within 30-minutes of the alarm. Upon receipt, operators will take prompt actions to confirm the validity of a single fire alarm. For EAL assessment purposes, the 30-minute clock starts at the time that the initial alarm was received, and not the time that a subsequent verification action was performed.

EAL BASES DOCUMENT Rev. 0 Page 53 of 258

ATTACHMENT 1 EAL Bases A single fire alarm, absent other indication(s) of a FIRE, may be indicative of equipment failure or a spurious activation, and not an actual FIRE. For this reason, additional time is allowed to verify the validity of the alarm. The 30-minute period is a reasonable amount of time to determine if an actual FIRE exists; however, after that time, and absent information to the contrary, it is assumed that an actual FIRE is in progress.

If an actual FIRE is verified by a report from the field, then HU4.1 is immediately applicable, and the emergency must be declared if the FIRE is not extinguished within 15-minutes of the report. If the alarm is verified to be due to an equipment failure or a spurious activation, and this verification occurs within 30-minutes of the receipt of the alarm, then this EAL is not applicable and no emergency declaration is warranted.

Basis-Related Requirements from Appendix R Appendix R to 10 CFR 50, states in part:

Criterion 3 of Appendix A to this part specifies that "Structures, systems, and components important to safety shall be designed and located to minimize, consistent with other safety requirements, the probability and effect of fires and explosions."

When considering the effects of fire, those systems associated with achieving and maintaining safe shutdown conditions assume major importance to safety because damage to them can lead to core damage resulting from loss of coolant through boil-off.

Because fire may affect safe shutdown systems and because the loss of function of systems used to mitigate the consequences of design basis accidents under post-fire conditions does not per se impact public safety, the need to limit fire damage to systems required to achieve and maintain safe shutdown conditions is greater than the need to limit fire damage to those systems required to mitigate the consequences of design basis accidents.

In addition, Appendix R to 10 CFR 50, requires, among other considerations, the use of 1-hour fire barriers for the enclosure of cable and equipment and associated non-safety circuits of one redundant train (G.2.c). As used in this EAL, the 30-minutes to verify a single alarm is well within this worst-case 1-hour time period.

Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via IC CA6 or SA9.

Table H-1 Fire Areas are based on DBNPS Unit 1 Fire Hazard Analysis Report. Table H-1 Fire Areas include those structures containing functions and systems required for safe shutdown of the plant (SAFETY SYSTEMS) (ref. 1, 2).

EAL BASES DOCUMENT Rev. 0 Page 54 of 258

ATTACHMENT 1 EAL Bases Basis Reference(s):

1. DBNPS Fire Hazard Analysis Report

2. DB-OP-02529 Fire Procedure

3. DBNPS-1465.00-00-0006 Design Basis Specification for the Plant Fire Protection

4. AP/0/A/5500/045 Plant Fire

5. Condition Report 09-66994, Emergency Action Level Clarification

6. Condition Report 09-69475, White Finding Identified For Inadequate Emergency Classification of Event

7. NEI 99-01 HU4 EAL BASES DOCUMENT Rev. 0 Page 55 of 258

ATTACHMENT 1 EAL Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 4 - FIRE Initiating Condition: FIRE potentially degrading the level of safety of the plant EAL:

HU4.3 Unusual Event A FIRE within the plant PROTECTED AREA not extinguished within 60 min. of the initial report, alarm or indication (Note 1)

Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Mode Applicability:

ALL Basis:

This IC addresses the magnitude and extent of FIRES that may be indicative of a potential degradation of the level of safety of the plant.

In addition to a FIRE addressed by EAL HU4.1 or HU4.2, a FIRE within the plant PROTECTED AREA not extinguished within 60-minutes may also potentially degrade the level of plant safety.

Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via IC CA6 or SA9.

Basis Reference(s):

1. NEI 99-01 HU4 EAL BASES DOCUMENT Rev. 0 Page 56 of 258

ATTACHMENT 1 EAL Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 4 - FIRE Initiating Condition: FIRE potentially degrading the level of safety of the plant EAL:

HU4.4 Unusual Event A FIRE within the plant PROTECTED AREA that requires firefighting support by an offsite fire response agency to extinguish Mode Applicability:

ALL Basis:

This IC addresses the magnitude and extent of FIRES that may be indicative of a potential degradation of the level of safety of the plant.

If a FIRE within the plant PROTECTED AREA is of sufficient size to require a response by an offsite firefighting agency (e.g., a local town Fire Department), then the level of plant safety is potentially degraded. The dispatch of an offsite firefighting agency to the site requires an emergency declaration only if it is needed to actively support firefighting efforts because the FIRE is beyond the capability of the Fire Brigade to extinguish. Declaration is not necessary if the agency resources are placed on stand-by, or supporting post-extinguishment recovery or investigation actions.

Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via IC CA6 or SA9.

Basis Reference(s):

1. NEI 99-01 HU4 EAL BASES DOCUMENT Rev. 0 Page 57 of 258

ATTACHMENT 1 EAL Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 5 - Hazardous Gases Initiating Condition: Gaseous release IMPEDING access to equipment necessary for normal plant operations, cooldown or shutdown EAL:

HA5.1 Alert Release of a toxic, corrosive, asphyxiant or flammable gas into ANY Table H-2 Safe Shutdown Rooms or Areas AND Entry into the room or area is prohibited or IMPEDED (Note 5)

Note 5: If the equipment in the listed room or area was already inoperable or out-of-service before the event occurred, then no emergency classification is warranted.

Table H-2 Safe Shutdown Rooms/Areas Room/Area Mode Applicability Aux Bldg. 565 ele. Room 236 #2 Mechanical 1, 2, 3 Penetration Room Aux Bldg. 585 ele. Room 304 corridor outside 1, 2, 3

1. 3 Mechanical Penetration Room Aux Bldg. 603 ele. Room 427 - #2 Electrical 1, 2, 3 Penetration Room Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby Basis:

This IC addresses an event involving a release of a hazardous gas that precludes or IMPEDES access to equipment necessary to maintain normal plant operation, or required for a normal plant cooldown and shutdown. This condition represents an actual or potential substantial degradation of the level of safety of the plant.

An Alert declaration is warranted if entry into the affected room/area is, or may be, procedurally required during the plant operating mode in effect at the time of the gaseous release. The emergency classification is not contingent upon whether entry is actually necessary at the time of the release.

Evaluation of the IC and EAL do not require atmospheric sampling; it only requires the Emergency Director judgment that the gas concentration in the affected room/area is sufficient EAL BASES DOCUMENT Rev. 0 Page 58 of 258

ATTACHMENT 1 EAL Bases to preclude or significantly IMPEDE procedurally required access. This judgment may be based on a variety of factors including an existing job hazard analysis, report of ill effects on personnel, advice from a subject matter expert or operating experience with the same or similar hazards. Access should be considered as IMPEDED if extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., requiring use of protective equipment, such as SCBAs, that is not routinely employed).

An emergency declaration is not warranted if ANY of the following conditions apply:

The plant is in an operating mode different than the mode specified for the affected room/area (i.e., entry is not required during the operating mode in effect at the time of the gaseous release). For example, the plant is in Mode 1 when the gaseous release occurs, and the procedures used for normal operation, cooldown and shutdown do not require entry into the affected room until Mode 4.

The gas release is a planned activity that includes compensatory measures, which address the temporary inaccessibility of a room or area (e.g., fire suppression system testing).

The action for which room/area entry is required is of an administrative or record keeping nature (e.g., normal rounds or routine inspections).

The access control measures are of a conservative or precautionary nature, and would not actually prevent or IMPEDE a required action.

If the equipment in the listed room or area was already inoperable, or out-of-service, before the event occurred, then no emergency should be declared since the event will have no adverse impact beyond that already allowed by Technical Specifications at the time of the event.

An asphyxiant is a gas capable of reducing the level of oxygen in the body to dangerous levels.

Most commonly, asphyxiants work by merely displacing air in an enclosed environment. This reduces the concentration of oxygen below the normal level of around 19%, which can lead to breathing difficulties, unconsciousness or even death. A steam leak of significant magnitude that prevents access to the area should be considered an asphyxiant.

This EAL does not apply to firefighting activities that automatically or manually activate a fire suppression system in an area.

Escalation of the emergency classification level would be via Recognition Category R, C or F ICs.

The list of plant rooms or areas with entry-related mode applicability identified specify those rooms or areas that contain equipment which require a manual/local action as specified in operating procedures used for normal plant operation, cooldown and shutdown. Rooms or areas in which actions of a contingent or emergency nature would be performed (e.g., an action to address an off-normal or emergency condition such as emergency repairs, corrective measures or emergency operations) are not included. In addition, the list specifies the plant mode(s) during which entry would be required for each room or area (ref. 1).

EAL BASES DOCUMENT Rev. 0 Page 59 of 258

ATTACHMENT 1 EAL Bases NOTE: IC HA5 mode applicability has been limited to the applicable modes identified in Table H-2 Safe Shutdown Rooms/Areas. If due to plant operating procedure or plant configuration changes, the applicable plant modes specified in Table H-2 are changed, a corresponding change to Attachment 3 Safe Shutdown Rooms/Areas Tables R-2 & H-2 Bases and to IC HA5 mode applicability is required.

Basis Reference(s):

1. Attachment 3 Safe Shutdown Rooms/Areas Tables R-2 & H-2 Bases

2. NEI 99-01 HA5 EAL BASES DOCUMENT Rev. 0 Page 60 of 258

ATTACHMENT 1 EAL Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 6 - Control Room Evacuation Initiating Condition: Inability to control a key safety function from outside the Control Room EAL:

HS6.1 Site Area Emergency An event has resulted in plant control being transferred from the Control Room to the Auxiliary Shutdown Panel AND Control of ANY of the following key safety functions is not reestablished within 15 min.

(Note 1):

Reactivity (modes 1, 2 and 3 only)

Core Cooling (RCS inventory)

RCS heat removal (ability to maintain heat sink)

Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4- Hot Shutdown, 5 - Cold Shutdown, 6 - Refueling Basis:

This IC addresses an evacuation of the Control Room that results in transfer of plant control to alternate locations, and the control of a key safety function cannot be reestablished in a timely manner. The failure to gain control of a key safety function following a transfer of plant control to alternate locations is a precursor to a challenge to one or more fission product barriers within a relatively short period of time.

The determination of whether or not control is established at the remote safe shutdown location(s) is based on Emergency Director judgment. The Emergency Director is expected to make a reasonable, informed judgment within 15-minutes whether or not the operating staff has control of key safety functions from the remote safe shutdown location(s).

Escalation of the emergency classification level would be via IC FG1 or CG1.

EAL BASES DOCUMENT Rev. 0 Page 61 of 258

ATTACHMENT 1 EAL Bases Shift Manager determines the Control Room requires evacuation. Control Room may become uninhabitable by FIRE, dense smoke, noxious fumes, bomb threat in or adjacent to the Control Room, or other life threatening conditions (ref. 1, 2).

The 15-minute time for transfer is based on analysis or assessments as to how quickly control must be reestablished without core uncovering and/or core damage. The 15-minute time period starts when EITHER 1) control of the plant is no longer maintained in the Control Room or 2) the last operator has left the Control Room, whichever comes first.

Basis Reference(s):

1. DB-OP-02508 Control Room Evacuation

2. DB-OP 02519 Serious Control Room Fire

3. NEI 99-01 HS6 EAL BASES DOCUMENT Rev. 0 Page 62 of 258

ATTACHMENT 1 EAL Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 6 - Control Room Evacuation Initiating Condition: Control Room evacuation resulting in transfer of plant control to alternate locations EAL:

HA6.1 Alert An event has resulted in plant control being transferred from the Control Room to the Auxiliary Shutdown Panel Mode Applicability:

ALL Basis:

This IC addresses an evacuation of the Control Room that results in transfer of plant control to alternate locations outside the Control Room. The loss of the ability to control the plant from the Control Room is considered to be a potential substantial degradation in the level of plant safety.

Following a Control Room evacuation, control of the plant will be transferred to alternate shutdown locations. The necessity to control a plant shutdown from outside the Control Room, in addition to responding to the event that required the evacuation of the Control Room, will present challenges to plant operators and other on-shift personnel. Activation of the ERO and emergency response facilities will assist in responding to these challenges.

This EAL is only applicable when the decision has been made to evacuate the Control Room, and not when conditions that may require evacuation are being evaluated by referring to either DB-OP-02519 or DB-OP-02508 (ref. 1, 2).

Escalation of the emergency classification level would be via IC HS6.

Shift Manager (SM) determines Control Room 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, 2).

Inability to establish plant control from outside the Control Room escalates this event to a Site Area Emergency per EAL HS6.1.

EAL BASES DOCUMENT Rev. 0 Page 63 of 258

ATTACHMENT 1 EAL Bases Basis Reference(s):

1. DB-OP-02508 Control Room Evacuation

2. DB-OP 02519 Serious Control Room Fire

3. NEI 99-01 HA6 EAL BASES DOCUMENT Rev. 0 Page 64 of 258

ATTACHMENT 1 EAL Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 7 - Emergency Director Judgment Initiating Condition: Other conditions exist which in the judgment of the Emergency Director warrant declaration of a General Emergency EAL:

HG7.1 General Emergency Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or have occurred which involve actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity or HOSTILE ACTION that results in an actual loss of physical control of the facility. Releases can be reasonably expected to exceed EPA Protective Action Guideline exposure levels offsite for more than the immediate site area.

Mode Applicability:

ALL Basis:

This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the emergency classification level description for a General Emergency.

The Emergency Director is the designated onsite individual having the responsibility and authority for implementing the DBNPS Emergency Response Plan. The Operations 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 Plant Management is expected to manage the emergency response as soon as available to do so in anticipation of the possible wide-ranging responsibilities associated with managing a major emergency (ref. 1).

Releases can reasonably be expected to exceed EPA PAG plume exposure levels outside the SITE BOUNDARY.

EAL BASES DOCUMENT Rev. 0 Page 65 of 258

ATTACHMENT 1 EAL Bases Basis Reference(s):

1. DBNPS Emergency Plan section 5.2 DBNPS Emergency Management

2. NEI 99-01 HG7 EAL BASES DOCUMENT Rev. 0 Page 66 of 258

ATTACHMENT 1 EAL Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 7 - Emergency Director Judgment Initiating Condition: Other conditions exist which in the judgment of the Emergency Director warrant declaration of a Site Area Emergency EAL:

HS7.1 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 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 Basis:

This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the emergency classification level description for a Site Area Emergency.

The Emergency Director is the designated onsite individual having the responsibility and authority for implementing the DBNPS Emergency Response Plan. The Operations 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 Plant Management is expected to manage the emergency response as soon as available to do so in anticipation of the possible wide-ranging responsibilities associated with managing a major emergency (ref. 1).

Basis Reference(s):

1. DBNPS Emergency Plan section 5.2 DBNPS Emergency Management

2. NEI 99-01 HS7 EAL BASES DOCUMENT Rev. 0 Page 67 of 258

ATTACHMENT 1 EAL Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 7 - Emergency Director Judgment Initiating Condition: Other conditions exist which in the judgment of the Emergency Director warrant declaration of an Alert EAL:

HA7.1 Alert Other conditions exist which, in the judgment of the Emergency Director, indicate that events are in progress or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant or a security event that involves probable life threatening risk to site personnel or damage to site equipment because of HOSTILE ACTION. ANY releases are expected to be limited to small fractions of the EPA Protective Action Guideline exposure levels.

Mode Applicability:

ALL Basis:

This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the emergency classification level description for an Alert.

The Emergency Director is the designated onsite individual having the responsibility and authority for implementing the DBNPS Emergency Response Plan. The Operations 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 Plant Management is expected to manage the emergency response as soon as available to do so in anticipation of the possible wide-ranging responsibilities associated with managing a major emergency (ref.1).

Basis Reference(s):

1. DBNPS Emergency Plan section 5.2 DBNPS Emergency Management

2. NEI 99-01 HA7 EAL BASES DOCUMENT Rev. 0 Page 68 of 258

ATTACHMENT 1 EAL Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 7 - Emergency Director Judgment Initiating Condition: Other conditions exist which in the judgment of the Emergency Director warrant declaration of a UE EAL:

HU7.1 Unusual Event Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or have occurred which indicate a potential degradation of the level of safety of the plant or indicate a security threat to facility protection has been initiated.

No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of SAFETY SYSTEMS occurs.

Mode Applicability:

ALL Basis:

This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the emergency classification level description for an Unusual Event.

The Emergency Director is the designated onsite individual having the responsibility and authority for implementing the DBNPS Emergency Response Plan. The Operations 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 Plant Management is expected to manage the emergency response as soon as available to do so in anticipation of the possible wide-ranging responsibilities associated with managing a major emergency (ref. 1).

Basis Reference(s):

1. DBNPS Emergency Plan section 5.2 DBNPS Emergency Management

2. NEI 99-01 HU7 EAL BASES DOCUMENT Rev. 0 Page 69 of 258

ATTACHMENT 1 EAL Bases Category R - Abnormal Rad Release / Rad Effluent EAL Group: ANY (EALs in this category are applicable to ANY plant condition, hot or cold.)

Many EALs are based on actual or potential degradation of fission product barriers because of the elevated potential for offsite radioactivity release. Degradation of fission product barriers though is not always apparent via non-radiological symptoms. Therefore, direct indication of elevated radiological effluents or area radiation levels are appropriate symptoms for 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. Radiological Effluent Direct indication of effluent radiation monitoring systems provides a rapid assessment mechanism to determine releases in excess of classifiable limits. Projected offsite doses, actual offsite field measurements or measured release rates via sampling indicate doses or dose rates above classifiable limits.

2. Irradiated Fuel Event Conditions indicative of a loss of adequate shielding or damage to irradiated fuel may preclude access to vital plant areas or result in radiological releases that warrant emergency classification.

3. Area Radiation Levels Sustained general area radiation levels, which may preclude access to areas requiring continuous occupancy, also warrant emergency classification.

EAL BASES DOCUMENT Rev. 0 Page 70 of 258

ATTACHMENT 1 EAL Bases Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid CDE EAL:

RG1.1 General Emergency Station Vent Channel 1 Noble Gas (RE 4598 AB/BB) > 8.4E +00 µCi/cc for 15 minutes or longer (Notes 1, 2, 3, 4)

Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.

Note 3: If the effluent flow past an effluent monitor is known to have stopped, indicating that the release path is isolated, the effluent monitor reading is no longer VALID for classification purposes.

Note 4: The pre-calculated effluent monitor values presented in EALs RA1.1, RS1.1 and RG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.

Mode Applicability:

ALL Basis:

This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude will require implementation of protective actions for the public.

Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.

The TEDE dose is set at the EPA PAG of 1,000 mrem while the 5,000 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.

Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer VALID for classification purposes.

EAL BASES DOCUMENT Rev. 0 Page 71 of 258

ATTACHMENT 1 EAL Bases The DBNPS gaseous effluent limits for RG1.1 are based on values that equate to an offsite dose greater than EITHER:

1000 mrem TEDE 5000 mrem CDE Thyroid Note: EAL thresholds reflect the State of Ohio guidance to utilities within their jurisdiction to evaluate the consequences of radiological releases in terms of a Child Thyroid PAG rather than an EPA-400 CDE Thyroid PAG.

The station vent is the primary effluent monitor release point. Other pathways such as Safety and relief valves, as well as AFW and MSSV, do not have direct effluent release monitors.

Releases from these points would be verified and monitored by RG1.3 survey readings.

The GE Station Vent gaseous effluent monitor value corresponds to calculated doses of 100% of the EPA Protective Action Guidelines (TEDE or CDE Thyroid) per EP-EALCALC-DB-0703 Radiological Gaseous Effluent EAL Values (EALs RG1, RS1, RA1 and RU1) (ref. 1).

Basis Reference(s):

1. EP-EALCALC-DB-0703 Rev. 2 Radiological Gaseous Effluent EAL Values (EALs RG1, RS1, RA1 and RU1)

2. NEI 99-01 AG1 EAL BASES DOCUMENT Rev. 0 Page 72 of 258

ATTACHMENT 1 EAL Bases Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid CDE EAL:

RG1.2 General Emergency Dose assessment using actual meteorology indicates doses > 1,000 mrem TEDE or 5,000 mrem thyroid CDE at or beyond the SITE BOUNDARY (Note 4)

Note 4: The pre-calculated effluent monitor values presented in EALs RA1.1, RS1.1 and RG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.

Mode Applicability:

ALL Basis:

This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude will require implementation of protective actions for the public.

Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.

The TEDE dose is set at the EPA PAG of 1,000 mrem while the 5,000 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.

Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer VALID for classification purposes.

Dose assessments are performed by computer-based methods (ref. 1, 2).

Note: EAL thresholds reflect the State of Ohio guidance to utilities within their jurisdiction to evaluate the consequences of radiological releases in terms of a Child Thyroid PAG rather than an EPA-400 CDE Thyroid PAG.

EAL BASES DOCUMENT Rev. 0 Page 73 of 258

ATTACHMENT 1 EAL Bases Basis Reference(s):

1. NOP-LP-5402 Davis-Besse MIDAS Dose Assessment Software

2. RA-EP-02240 Offsite Dose Assessment

3. NEI 99-01 AG1 EAL BASES DOCUMENT Rev. 0 Page 74 of 258

ATTACHMENT 1 EAL Bases Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid CDE EAL:

RG1.3 General Emergency Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY:

Closed window dose rates > 1,000 mR/hr expected to continue for 60 min.

Analyses of field survey samples indicate thyroid CDE > 5,000 mrem for 60 min. of inhalation.

(Notes 1, 2)

Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.

Mode Applicability:

ALL Basis:

This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude will require implementation of protective actions for the public.

Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.

The TEDE dose is set at the EPA PAG of 1,000 mrem while the 5,000 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.

EAL BASES DOCUMENT Rev. 0 Page 75 of 258

ATTACHMENT 1 EAL Bases NOP-LP-5015 FENOC Field Monitoring Teams Radiation Monitoring Teams Field Surveys provides guidance for emergency or post-accident radiological environmental monitoring (ref. 1).

Note: EAL thresholds reflect the State of Ohio guidance to utilities within their jurisdiction to evaluate the consequences of radiological releases in terms of a Child Thyroid PAG rather than an EPA-400 CDE Thyroid PAG.

Basis Reference(s):

1. NOP-LP-5015 FENOC Field Monitoring Teams Radiation Monitoring Teams Field Surveys

2. NEI 99-01 AG1 EAL BASES DOCUMENT Rev. 0 Page 76 of 258

ATTACHMENT 1 EAL Bases Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem thyroid CDE EAL:

RS1.1 Site Area Emergency Station Vent Channel 1 Noble Gas (RE 4598 AB/BB) > 8.4E -01 µCi/cc for 15 minutes or longer (Notes 1, 2, 3, 4)

Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.

Note 3: If the effluent flow past an effluent monitor is known to have stopped, indicating that the release path is isolated, the effluent monitor reading is no longer VALID for classification purposes.

Note 4: The pre-calculated effluent monitor values presented in EALs RA1.1, RS1.1 and RG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.

Mode Applicability:

ALL Basis:

This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to 10% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public.

Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.

The TEDE dose is set at 10% of the EPA PAG of 1,000 mrem while the 500 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.

Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer VALID for classification purposes.

EAL BASES DOCUMENT Rev. 0 Page 77 of 258

ATTACHMENT 1 EAL Bases Escalation of the emergency classification level would be via IC RG1.

The DBNPS gaseous effluent limits for RS1.1 are based on values that equate to an offsite dose greater than EITHER:

100 mrem TEDE 500 mrem CDE Thyroid Note: EAL thresholds reflect the State of Ohio guidance to utilities within their jurisdiction to evaluate the consequences of radiological releases in terms of a Child Thyroid PAG rather than an EPA-400 CDE Thyroid PAG.

The station vent is the primary effluent monitor release point. Other pathways such as Safety and relief valves, as well as AFW and MSSV, do not have direct effluent release monitors.

Releases from these points would be verified and monitored by RS1.3 survey readings.

The SAE Station Vent gaseous effluent monitor value corresponds to calculated doses of 10% of the EPA Protective Action Guidelines (TEDE or CDE Thyroid) per EP-EALCALC-DB-0703 Radiological Gaseous Effluent EAL Values (EALs RG1, RS1, RA1 and RU1) (ref. 1).

Basis Reference(s):

1. EP-EALCALC-DB-0703 Rev.2 Radiological Gaseous Effluent EAL Values (EALs RG1, RS1, RA1 and RU1)

2. NEI 99-01 AS1 EAL BASES DOCUMENT Rev. 0 Page 78 of 258

ATTACHMENT 1 EAL Bases Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem thyroid CDE EAL:

RS1.2 Site Area Emergency Dose assessment using actual meteorology indicates doses > 100 mrem TEDE or 500 mrem thyroid CDE at or beyond the SITE BOUNDARY (Note 4)

Note 4: The pre-calculated effluent monitor values presented in EALs RA1.1, RS1.1 and RG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.

Mode Applicability:

ALL Basis:

This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to 10% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public.

Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.

The TEDE dose is set at 10% of the EPA PAG of 1,000 mrem while the 500 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.

Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer VALID for classification purposes.

Escalation of the emergency classification level would be via IC RG1.

Dose assessments are performed by computer-based or manual methods (ref. 1, 2).

EAL BASES DOCUMENT Rev. 0 Page 79 of 258

ATTACHMENT 1 EAL Bases Note: EAL thresholds reflect the State of Ohio guidance to utilities within their jurisdiction to evaluate the consequences of radiological releases in terms of a Child Thyroid PAG rather than an EPA-400 CDE Thyroid PAG.

Basis Reference(s):

1. NOP-LP-5402 Davis-Besse MIDAS Dose Assessment Software

2. RA-EP-02240 Offsite Dose Assessment

3. NEI 99-01 AS1 EAL BASES DOCUMENT Rev. 0 Page 80 of 258

ATTACHMENT 1 EAL Bases Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem thyroid CDE EAL:

RS1.3 Site Area Emergency Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY:

Closed window dose rates > 100 mR/hr expected to continue for 60 min.

Analyses of field survey samples indicate thyroid CDE > 500 mrem for 60 min. of inhalation.

(Notes 1, 2)

Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.

Mode Applicability:

ALL Basis:

This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to 10% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public.

Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.

The TEDE dose is set at 10% of the EPA PAG of 1,000 mrem while the 500 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.

Escalation of the emergency classification level would be via IC RG1.

EAL BASES DOCUMENT Rev. 0 Page 81 of 258

ATTACHMENT 1 EAL Bases NOP-LP-5015 FENOC Field Monitoring Teams Radiation Monitoring Teams Field Surveys provides guidance for emergency or post-accident radiological environmental monitoring (ref. 1).

Note: EAL thresholds reflect the State of Ohio guidance to utilities within their jurisdiction to evaluate the consequences of radiological releases in terms of a Child Thyroid PAG rather than an EPA-400 CDE Thyroid PAG.

Basis Reference(s):

1. NOP-LP-5015 FENOC Field Monitoring Teams Radiation Monitoring Teams Field Surveys

2. NEI 99-01 AS1 EAL BASES DOCUMENT Rev. 0 Page 82 of 258

ATTACHMENT 1 EAL Bases Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid CDE EAL:

RA1.1 Alert Station Vent Channel 1 Noble Gas (RE 4598 AB/BB) > 8.4E -02 µCi/cc for 15 minutes or longer (Notes 1, 2, 3, 4)

Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.

Note 3: If the effluent flow past an effluent monitor is known to have stopped, indicating that the release path is isolated, the effluent monitor reading is no longer VALID for classification purposes.

Note 4: The pre-calculated effluent monitor values presented in EALs RA1.1, RS1.1 and RG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.

Mode Applicability:

ALL Basis:

This IC addresses a release of gaseous or liquid radioactivity that results in projected or actual offsite doses greater than or equal to 1% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude represent an actual or potential substantial degradation of the level of safety of the plant as indicated by a radiological release that significantly exceeds regulatory limits (e.g., a significant uncontrolled release).

Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.

The TEDE dose is set at 1% of the EPA PAG of 1,000 mrem while the 50 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.

EAL BASES DOCUMENT Rev. 0 Page 83 of 258

ATTACHMENT 1 EAL Bases Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer VALID for classification purposes.

Escalation of the emergency classification level would be via IC RS1.

The DBNPS gaseous effluent limits for RA1.1 are based on values that equate to an offsite dose greater than EITHER:

10 mrem TEDE 50 mrem CDE Thyroid Note: EAL thresholds reflect the State of Ohio guidance to utilities within their jurisdiction to evaluate the consequences of radiological releases in terms of a Child Thyroid PAG rather than an EPA-400 CDE Thyroid PAG.

The Alert Station Vent gaseous effluent monitor value corresponds to calculated doses of 1%

of the EPA Protective Action Guidelines (TEDE or CDE Thyroid) per EP-EALCALC-DB-0703 Radiological Gaseous Effluent EAL Values (EALs RG1, RS1, RA1 and RU1) (ref. 1).

The station vent is the primary effluent monitor release point. Other pathways such as Safety and relief valves, as well as AFW and MSSV, do not have direct effluent release monitors.

Releases from these points would be verified and monitored by RA1.3 survey readings.

Basis Reference(s):

1. EP-EALCALC-DB-0703 Rev.2 Radiological Gaseous Effluent EAL Values (EALs RG1, RS1, RA1 and RU1)

2. NEI 99-01 AA1 EAL BASES DOCUMENT Rev. 0 Page 84 of 258

ATTACHMENT 1 EAL Bases Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid CDE EAL:

RA1.2 Alert Dose assessment using actual meteorology indicates doses > 10 mrem TEDE or 50 mrem thyroid CDE at or beyond the SITE BOUNDARY (Note 4)

Note 4: The pre-calculated effluent monitor values presented in EALs RA1.1, RS1.1 and RG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.

Mode Applicability:

ALL Basis:

This IC addresses a release of gaseous or liquid radioactivity that results in projected or actual offsite doses greater than or equal to 1% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude represent an actual or potential substantial degradation of the level of safety of the plant as indicated by a radiological release that significantly exceeds regulatory limits (e.g., a significant uncontrolled release).

Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.

The TEDE dose is set at 1% of the EPA PAG of 1,000 mrem while the 50 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.

Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer VALID for classification purposes.

Escalation of the emergency classification level would be via IC RS1.

Dose assessments are performed by computer-based and manual methods (ref. 1, 2).

EAL BASES DOCUMENT Rev. 0 Page 85 of 258

ATTACHMENT 1 EAL Bases Note: EAL thresholds reflect the State of Ohio guidance to utilities within their jurisdiction to evaluate the consequences of radiological releases in terms of a Child Thyroid PAG rather than an EPA-400 CDE Thyroid PAG.

Basis Reference(s):

1. NOP-LP-5402 Davis-Besse MIDAS Dose Assessment Software

2. RA-EP-02240 Offsite Dose Assessment

3. NEI 99-01 AA1 EAL BASES DOCUMENT Rev. 0 Page 86 of 258

ATTACHMENT 1 EAL Bases Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid CDE EAL:

RA1.3 Alert Analysis of a liquid effluent sample indicates a concentration or release rate that would result in doses > 10 mrem TEDE or 50 mrem thyroid CDE at or beyond the SITE BOUNDARY for 60 min. of exposure (Notes 1, 2)

Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.

Mode Applicability:

ALL Basis:

This IC addresses a release of gaseous or liquid radioactivity that results in projected or actual offsite doses greater than or equal to 1% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude represent an actual or potential substantial degradation of the level of safety of the plant as indicated by a radiological release that significantly exceeds regulatory limits (e.g., a significant uncontrolled release).

Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.

The TEDE dose is set at 1% of the EPA PAG of 1,000 mrem while the 50 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.

Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer VALID for classification purposes.

EAL BASES DOCUMENT Rev. 0 Page 87 of 258

ATTACHMENT 1 EAL Bases Escalation of the emergency classification level would be via IC RS1.

Dose assessments based on liquid releases are performed per Offsite Dose Calculation Manual (ref. 1).

Note: EAL thresholds reflect the State of Ohio guidance to utilities within their jurisdiction to evaluate the consequences of radiological releases in terms of a Child Thyroid PAG rather than an EPA-400 CDE Thyroid PAG.

Basis Reference(s):

1. DBNPS ODCM

2. NEI 99-01 AA1 EAL BASES DOCUMENT Rev. 0 Page 88 of 258

ATTACHMENT 1 EAL Bases Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid CDE EAL:

RA1.4 Alert Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY:

Closed window dose rates > 10 mR/hr expected to continue for 60 min.

Analyses of field survey samples indicate thyroid CDE > 50 mrem for 60 min. of inhalation.

(Notes 1, 2)

Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.

Mode Applicability:

ALL Basis:

This IC addresses a release of gaseous or liquid radioactivity that results in projected or actual offsite doses greater than or equal to 1% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude represent an actual or potential substantial degradation of the level of safety of the plant as indicated by a radiological release that significantly exceeds regulatory limits (e.g., a significant uncontrolled release).

Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.

The TEDE dose is set at 1% of the EPA PAG of 1,000 mrem while the 50 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.

EAL BASES DOCUMENT Rev. 0 Page 89 of 258

ATTACHMENT 1 EAL Bases Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer VALID for classification purposes.

Escalation of the emergency classification level would be via IC RS1.

NOP-LP-5015 FENOC Field Monitoring Teams Radiation Monitoring Teams Field Surveys provides guidance for emergency or post-accident radiological environmental monitoring (ref. 1).

Note: EAL thresholds reflect the State of Ohio guidance to utilities within their jurisdiction to evaluate the consequences of radiological releases in terms of a Child Thyroid PAG rather than an EPA-400 CDE Thyroid PAG.

Basis Reference(s):

1. NOP-LP-5015 FENOC Field Monitoring Teams Radiation Monitoring Teams Field Surveys

2. NEI 99-01 AA1 EAL BASES DOCUMENT Rev. 0 Page 90 of 258

ATTACHMENT 1 EAL Bases Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous or liquid radioactivity greater than 2 times the ODCM limits for 60 minutes or longer EAL:

RU1.1 Unusual Event Station Vent Channel 1 Noble Gas (RE 4598AA/BA) > 5.72E-03 µCi/cc for 60 minutes or longer OR ANY of the following effluent monitors > 2 times the high alarm setpoint established by a current radioactivity discharge permit for 60 minutes or longer:

Waste Gas System Outlet (RE 1822A or B).

Clean Waste System Outlet (RE 1770A or B).

Miscellaneous Waste System Outlet (RE 1878A or B).

Discharge permit specified monitor.

(Notes 1, 2, 3)

Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.

Note 3: If the effluent flow past an effluent monitor is known to have stopped, indicating that the release path is isolated, the effluent monitor reading is no longer VALID for classification purposes.

Mode Applicability:

ALL Basis:

This IC addresses a potential decrease in the level of safety of the plant as indicated by a low-level radiological release that exceeds regulatory commitments for an extended period of time (e.g., an uncontrolled release). It includes ANY gaseous or liquid radiological release, EAL BASES DOCUMENT Rev. 0 Page 91 of 258

ATTACHMENT 1 EAL Bases monitored or un-monitored, including those for which a radioactivity discharge permit is normally prepared.

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

Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.

Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer VALID for classification purposes.

Releases should not be prorated or averaged. For example, a release exceeding 4 times release limits for 30 minutes does not meet the EAL.

This EAL addresses normally occurring continuous radioactivity releases from monitored gaseous or liquid effluent pathways or radioactivity releases that cause effluent radiation monitor readings to exceed 2 times the limit established by a radioactivity discharge permit.

This EAL will typically be associated with planned batch releases from non-continuous release pathways (e.g., radwaste, waste gas).

Escalation of the emergency classification level would be via IC RA1.

The UE gaseous and liquid release values represent two times the appropriate ODCM release rate limits associated with the specified monitors (ref. 2).

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

Station Vent - RE 4598AA or BA Waste Gas System Outlet - RE 1822 A or B (batch release)

The Station Vent gaseous effluent (RE 4598AA/BA) value of 2 times the ODCM setpoint was determined using formulas, isotopic dose conversion factors and meteorology data as specified in the ODCM based on a normal operating isotopic mixture (no clad damage condition). Assumptions and calculation inputs are provided in EP-EALCALC-DB-0703 Rev. 2 (ref. 2).

EAL BASES DOCUMENT Rev. 0 Page 92 of 258

ATTACHMENT 1 EAL Bases The Waste Gas Outlet (RE 1822 A or B for batch release) are the release monitors normally used for planned discharges. If a discharge is performed using a different flow path or effluent monitor (e.g., a portable or temporary effluent monitor), then the declaration criteria will be based on the monitor specified in the discharge permit.

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

Clean Waste System Outlet - RE 1770 A or B (batch release)

Misc. Waste System Outlet - RE 1878 A or B (batch release)

The liquid release monitors listed are those normally used for planned discharges. If a discharge is performed using a different flow path or effluent monitor (e.g., a portable or temporary effluent monitor), then the declaration criteria will be based on the monitor specified in the discharge permit.

Basis Reference(s):

1. DBNPS ODCM

2. EP-EALCALC-DB-0703 Rev. 2 Radiological Gaseous Effluent EAL Values (EALs RG1, RS1, RA1 and RU1)

3. NEI 99-01 AU1 EAL BASES DOCUMENT Rev. 0 Page 93 of 258

ATTACHMENT 1 EAL Bases Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous or liquid radioactivity greater than 2 times the ODCM limits for 60 minutes or longer.

EAL:

RU1.2 Unusual Event Sample analysis for a gaseous or liquid release indicates a concentration or release rate 2 x ODCM limits for 60 min. (Notes 1, 2)

Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.

Mode Applicability:

ALL Basis:

This IC addresses a potential decrease in the level of safety of the plant as indicated by a low-level radiological release that exceeds regulatory commitments for an extended period of time (e.g., an uncontrolled release). It includes ANY gaseous or liquid radiological release, monitored or un-monitored, including those for which a radioactivity discharge permit is normally prepared.

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

Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.

Releases should not be prorated or averaged. For example, a release exceeding 4 times release limits for 30 minutes does not meet the EAL.

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

EAL BASES DOCUMENT Rev. 0 Page 94 of 258

ATTACHMENT 1 EAL Bases Escalation of the emergency classification level would be via IC RA1.

Grab samples are used to determine release concentrations or release rates, confirm meter readings, or indicate the need for sampling when the effluent monitors are not in service or other alarms occur. The maximum instantaneous release rate limits are calculated in accordance with the ODCM. These are indicated on approved discharge permit release packages (ref. 1).

Basis Reference(s):

1. DBNPS ODCM

2. NEI 99-01 AU1 EAL BASES DOCUMENT Rev. 0 Page 95 of 258

ATTACHMENT 1 EAL Bases Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 2 - Irradiated Fuel Event Initiating Condition: Spent fuel pool level cannot be restored to at least the top of the fuel racks for 60 minutes or longer EAL:

RG2.1 General Emergency Spent fuel pool level cannot be restored to at least 1 ft. for > 60 min. (Note 1)

Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Mode Applicability:

ALL Basis:

This EAL addresses a significant loss of spent fuel pool inventory control and makeup capability leading to a prolonged uncovery of spent fuel. This condition will lead to fuel damage and a radiological release to the environment.

It is recognized that this IC would likely not be met until well after another General Emergency IC was met; however, it is included to provide classification diversity.

Post-Fukushima order EA-12-051 (ref.1) required the installation of reliable SFP level indication capable of identifying normal level (Level 1), SFP level 10 ft. above the top of the fuel racks (Level 2) and SFP level at the top of the fuel racks (Level 3).

Indicated level of 1 ft. on LI4801A (primary) or LI4801B (Backup) corresponds approximately to 1 ft. above the top of the SFP racks (ref. 2)

Basis Reference(s):

1. NRC EA-12-51 Issuance of Order to Modify Licenses with Regard to Reliable Spent Fuel Pool Instrumentation

2. Engineering Change Package 13-0596

3. NEI 99-01 AG2 EAL BASES DOCUMENT Rev. 0 Page 96 of 258

ATTACHMENT 1 EAL Bases Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 2 - Irradiated Fuel Event Initiating Condition: Spent fuel pool level at the top of the fuel racks EAL:

RS2.1 Site Area Emergency Lowering of spent fuel pool level to 1 ft.

Mode Applicability:

ALL Basis:

This EAL addresses a significant loss of spent fuel pool inventory control and makeup capability leading to IMMINENT fuel damage. This condition entails major failures of plant functions needed for protection of the public and thus warrant a Site Area Emergency declaration.

It is recognized that this IC would likely not be met until well after another Site Area Emergency IC was met; however, it is included to provide classification diversity.

Escalation of the emergency classification level would be via IC RG1 or RG2.

Post-Fukushima order EA-12-051 (ref.1) required the installation of reliable SFP level indication capable of identifying normal level (Level 1), SFP level 10 ft. above the top of the fuel racks (Level 2) and SFP level at the top of the fuel racks (Level 3).

Indicated level of 1 ft. on LI4801A (primary) or LI4801B (Backup) corresponds approximately to 1 ft. above the top of the SFP racks (ref. 2).

Basis Reference(s):

1. NRC EA-12-51 Issuance of Order to Modify Licenses with Regard to Reliable Spent Fuel Pool Instrumentation

2. Engineering Change Package 13-0596

3. NEI 99-01 AS2 EAL BASES DOCUMENT Rev. 0 Page 97 of 258

ATTACHMENT 1 EAL Bases Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 2 - Irradiated Fuel Event Initiating Condition: Significant lowering of water level above, or damage to, irradiated fuel EAL:

RA2.1 Alert Uncovery of irradiated fuel in the REFUELING PATHWAY Mode Applicability:

ALL Basis:

This IC addresses events that have caused IMMINENT or actual damage to an irradiated fuel assembly, or a significant lowering of water level within the spent fuel pool. These events present radiological safety challenges to plant personnel and are precursors to a release of radioactivity to the environment. As such, they represent an actual or potential substantial degradation of the level of safety of the plant.

This IC applies to irradiated fuel that is licensed for dry storage up to the point that the loaded storage cask is sealed. Once sealed, damage to a loaded cask causing loss of the CONFINEMENT BOUNDARY is classified in accordance with EAL EU1.1.

Escalation of the emergency would be based on either Recognition Category R or C ICs.

This EAL escalates from RU2.1 in that the loss of level, in the affected portion of the REFUELING PATHWAY, is of sufficient magnitude to have resulted in uncovery of irradiated fuel. Indications of irradiated fuel uncovery may include direct or indirect visual observation (e.g., reports from personnel or camera images), as well as significant changes in water and radiation levels, or other plant parameters. Computational aids may also be used (e.g., a boil-off curve). Classification of an event using this EAL should be based on the totality of available indications, reports and observations.

While an area radiation monitor could detect an increase in a dose rate due to a lowering of water level in some portion of the REFUELING PATHWAY, the reading may not be a reliable indication of whether or not the fuel is actually uncovered. To the degree possible, readings should be considered in combination with other available indications of inventory loss.

A drop in water level above irradiated fuel within the reactor vessel may be classified in accordance Recognition Category C during the Cold Shutdown and Refueling modes.

Escalation of the emergency classification level would be via IC RS1.

EAL BASES DOCUMENT Rev. 0 Page 98 of 258

ATTACHMENT 1 EAL Bases Basis Reference(s):

1. DB-OP-02003 3-1-B, SFP LVL

2. DB-OP-02003 3-1-A, REFUELING CANAL LVL

3. DB-OP-02547 Spent Fuel Pool Cooling Malfunctions

4. NEI 99-01 AA2 EAL BASES DOCUMENT Rev. 0 Page 99 of 258

ATTACHMENT 1 EAL Bases Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 2 - Irradiated Fuel Event Initiating Condition: Significant lowering of water level above, or damage to, irradiated fuel EAL:

RA2.2 Alert Damage to irradiated fuel resulting in a release of radioactivity as indicated by a high radiation alarm on ANY of the following radiation monitor indications:

RE 8426 SFP Area RE 8427 SFP Area RE 8417 Fuel Handling Area RE 8418 Fuel Handling Area RE 8425 Equipment Hatch Area RE 8446/8447 Fuel Handling Exhaust RE 4598AA/BA Station Vent Mode Applicability:

ALL Basis:

This IC addresses events that have caused IMMINENT or actual damage to an irradiated fuel assembly, or a significant lowering of water level within the spent fuel pool. These events present radiological safety challenges to plant personnel and are precursors to a release of radioactivity to the environment. As such, they represent an actual or potential substantial degradation of the level of safety of the plant.

Escalation of the emergency would be based on either Recognition Category R or C ICs.

This EAL addresses a release of radioactive material caused by mechanical damage to irradiated fuel. Damaging events may include the dropping, bumping or binding of an assembly, or dropping a heavy load onto an assembly. A rise in readings on radiation monitors should be considered in conjunction with in-plant reports or observations of a potential fuel damaging event (e.g., a fuel handling accident).

Escalation of the emergency classification level would be via IC RS1.

The specified radiation monitors are those expected to see increase area radiation levels as a result of damage to irradiated fuel (ref. 1).

EAL BASES DOCUMENT Rev. 0 Page 100 of 258

ATTACHMENT 1 EAL Bases The high alarm setpoints for the radiation monitors are set to be indicative of significant increases in area and/or airborne radiation (ref. 2).

Basis Reference(s):

1. DB-OP-02530 Fuel Handling Accident

2. Radiation Setpoints Manual

3. NEI 99-01 AA2 EAL BASES DOCUMENT Rev. 0 Page 101 of 258

ATTACHMENT 1 EAL Bases Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 2 - Irradiated Fuel Event Initiating Condition: Significant lowering of water level above, or damage to, irradiated fuel EAL:

RA2.3 Alert Lowering of spent fuel pool level to 10 ft.

Mode Applicability:

ALL Basis:

This IC addresses events that have caused IMMINENT or actual damage to an irradiated fuel assembly, or a significant lowering of water level within the spent fuel pool. These events present radiological safety challenges to plant personnel and are precursors to a release of radioactivity to the environment. As such, they represent an actual or potential substantial degradation of the level of safety of the plant.

Escalation of the emergency would be based on either Recognition Category R or C ICs.

Spent fuel pool water level at this value is within the lower end of the level range necessary to prevent significant dose consequences from direct gamma radiation to personnel performing operations in the vicinity of the spent fuel pool. This condition reflects a significant loss of spent fuel pool water inventory and thus it is also a precursor to a loss of the ability to adequately cool the irradiated fuel assembles stored in the pool.

Escalation of the emergency classification level would be via IC RS1or RS2.

Post-Fukushima order EA-12-051 (ref.1) required the installation of reliable SFP level indication capable of identifying normal level (Level 1), SFP level 10 ft. above the top of the fuel racks (Level 2) and SFP level at the top of the fuel racks (Level 3).

Indicated level of 10 ft. on LI4801A (primary) or LI4801B (Backup) corresponds to plant elevation 588 ft. 6 in. (ref. 2).

Basis Reference(s):

1. NRC EA-12-51 Issuance of Order to Modify Licenses with Regard to Reliable Spent Fuel Pool Instrumentation

2. Engineering Change Package 13-0596

3. NEI 99-01 AA2 EAL BASES DOCUMENT Rev. 0 Page 102 of 258

ATTACHMENT 1 EAL Bases Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 2 - Irradiated Fuel Event Initiating Condition: UNPLANNED loss of water level above irradiated fuel EAL:

RU2.1 Unusual Event UNPLANNED water level drop in the REFUELING PATHWAY as indicated by ANY of the following:

SFP level alarm or indication (Ll 1600)

SFP level indication (Ll 4801 A/B)

Refueling Canal low water level alarm or indication (Ll 1627)

Visual observation AND UNPLANNED rise in corresponding area radiation levels as indicated by ANY of the following radiation monitors:

RE 8426 SFP Area RE 8427 SFP Area RE 8417 Fuel Handling Area RE 8418 Fuel Handling Area RE 8425 Equipment Hatch Area Mode Applicability:

ALL Basis:

This IC addresses a decrease in water level above irradiated fuel sufficient to cause elevated radiation levels. This condition could be a precursor to a more serious event and is also indicative of a minor loss in the ability to control radiation levels within the plant. It is therefore a potential degradation in the level of safety of the plant.

A water level decrease will be primarily determined by indications from available level instrumentation. Other sources of level indications may include reports from plant personnel (e.g., from a refueling crew) or video camera observations (if available). A significant drop in the water level may also cause an increase in the radiation levels of adjacent areas that can be detected by monitors in those locations.

EAL BASES DOCUMENT Rev. 0 Page 103 of 258

ATTACHMENT 1 EAL Bases The effects of planned evolutions should be considered. For example, a refueling bridge area radiation monitor reading may increase due to planned evolutions such as lifting of the reactor vessel head or movement of a fuel assembly. Note that this EAL is applicable only in cases where the elevated reading is due to an UNPLANNED loss of water level.

A drop in water level above irradiated fuel within the reactor vessel may be classified in accordance Recognition Category C during the Cold Shutdown and Refueling modes.

Escalation of the emergency classification level would be via IC RA2.

The SFP low water level alarm setpoint is actuated by LS 1600 at a setpoint of 23.2 feet on LI 1600 (ref. 1).

The refueling canal low water level alarm setpoint is actuated by LS 1627 at a setpoint of 23.4 feet on LI 1627 (this corresponds to 1.5 inches below normal) (ref. 2).

The Ll 4801 A/B is indication only, with no alarm or computer point function.

Water level restoration instructions are performed in accordance with AOPs (ref. 1, 2, 3).

The specified radiation monitors are those expected to see increase area radiation levels as a result of a loss of REFUELING PATHWAY inventory (ref. 1). Increasing radiation indications on these monitors in the absence of indications of decreasing REFUELING PATHWAY level are not classifiable under this EAL.

When the spent fuel pool and reactor cavity are connected, there could exist the possibility of uncovering irradiated fuel. Therefore, this EAL is applicable for conditions in which irradiated fuel is being transferred to and from the reactor vessel and spent fuel pool.

Basis Reference(s):

1. DB-OP-02003 3-1-B, SFP LVL

2. DB-OP-02003 3-1-A, REFUELING CANAL LVL

3. DB-OP-02547 Spent Fuel Pool Cooling Malfunctions

4. NEI 99-01 AU2 EAL BASES DOCUMENT Rev. 0 Page 104 of 258

ATTACHMENT 1 EAL Bases Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 3 - Area Radiation Levels Initiating Condition: Radiation levels that IMPEDE access to equipment necessary for normal plant operations, cooldown or shutdown EAL:

RA3.1 Alert Dose rates > 15 mR/hr in EITHER of the following areas:

Control Room (RE 8430 or 8431)

Central Alarm Station (RE 8435 or 8436)

Mode Applicability:

ALL Basis:

This IC addresses elevated radiation levels in certain plant rooms/areas sufficient to preclude or IMPEDE personnel from performing actions necessary to maintain normal plant operation, or to perform a normal plant cooldown and shutdown. As such, it represents an actual or potential substantial degradation of the level of safety of the plant. The Emergency Director should consider the cause of the increased radiation levels and determine if another IC may be applicable.

Escalation of the emergency classification level would be via Recognition Category R, C or F ICs.

Areas that meet this threshold include the Control Room and the Central Alarm Station (CAS).

RE 8430 and 8431 monitors the Control room for area radiation. The CAS area does not have installed radiation monitoring but RE 8435 and 8436 are in near proximity and can be used to approximate CAS area radiation levels (ref. 1, 2). The CAS is included because of its importance to permitting access to areas required to assure safe plant operations.

Basis Reference(s):

1. DB-OP-06412 Process and Area Radiation Monitor

2. SD-017B System Description for Area Radiation Monitors

3. NEI 99-01 AA3 EAL BASES DOCUMENT Rev. 0 Page 105 of 258

ATTACHMENT 1 EAL Bases Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 3 - Area Radiation Levels Initiating Condition: Radiation levels that IMPEDE access to equipment necessary for normal plant operations, cooldown or shutdown EAL:

RA3.2 Alert An UNPLANNED event results in radiation levels that prohibit or IMPEDE access to ANY Table R-2 Safe Shutdown Rooms or Areas (Note 5)

Note 5: If the equipment in the listed room or area was already inoperable or out-of-service before the event occurred, then no emergency classification is warranted.

Table R-2 Safe Shutdown Rooms/Areas Room/Area Mode Applicability Aux Bldg. 565 ele. Room 236 #2 Mechanical 1, 2, 3 Penetration Room Aux Bldg. 585 ele. Room 304 corridor outside 1, 2, 3

1. 3 Mechanical Penetration Room Aux Bldg. 603 ele. Room 427 - #2 Electrical 1, 2, 3 Penetration Room Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby Basis:

This IC addresses elevated radiation levels in certain plant rooms/areas sufficient to preclude or IMPEDE personnel from performing actions necessary to maintain normal plant operation, or to perform a normal plant cooldown and shutdown. As such, it represents an actual or potential substantial degradation of the level of safety of the plant. The Emergency Director should consider the cause of the increased radiation levels and determine if another IC may be applicable.

For RA3.2, an Alert declaration is warranted if entry into the affected room/area is, or may be, procedurally required during the plant operating mode in effect at the time of the elevated radiation levels. The emergency classification is not contingent upon whether entry is actually necessary at the time of the increased radiation levels. Access should be considered as IMPEDED if extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., installing temporary shielding, requiring use of non-routine protective equipment, requesting an extension in dose limits beyond normal administrative limits).

EAL BASES DOCUMENT Rev. 0 Page 106 of 258

ATTACHMENT 1 EAL Bases An emergency declaration is not warranted if ANY of the following conditions apply:

The plant is in an operating mode different than the mode specified for the affected room/area (i.e., entry is not required during the operating mode in effect at the time of the elevated radiation levels). For example, the plant is in Mode 1 when the radiation increase occurs, and the procedures used for normal operation, cooldown and shutdown do not require entry into the affected room until Mode 4.

The increased radiation levels are a result of a planned activity that includes compensatory measures, which address the temporary inaccessibility of a room or area (e.g., radiography, spent filter or resin transfer, etc.).

The action for which room/area entry is required is of an administrative or record keeping nature (e.g., normal rounds or routine inspections).

The access control measures are of a conservative or precautionary nature, and would not actually prevent or IMPEDE a required action.

If the equipment in the listed room or area was already inoperable, or out-of-service, before the event occurred, then no emergency should be declared since the event will have no adverse impact beyond that already allowed by Technical Specifications at the time of the event.

Escalation of the emergency classification level would be via Recognition Category R, C or F ICs.

The list of plant rooms or areas with entry-related mode applicability identified specify those rooms or areas that contain equipment which require a manual/local action as specified in operating procedures used for normal plant operation, cooldown and shutdown. Rooms or areas in which actions of a contingent or emergency nature would be performed (e.g., an action to address an off-normal or emergency condition such as emergency repairs, corrective measures or emergency operations) are not included. In addition, the list specifies the plant mode(s) during which entry would be required for each room or area (ref. 1).

NOTE: EAL RA3.2 mode applicability has been limited to the applicable modes identified in Table R-2 Safe Shutdown Rooms/Areas. If due to plant operating procedure or plant configuration changes, the applicable plant modes specified in Table R-2 are changed, a corresponding change to Attachment 3 Safe Shutdown Rooms/Areas Tables R-2 & H-2 Bases and to EAL RA3.2 mode applicability is required.

Basis Reference(s):

1. Attachment 3 Safe Shutdown Rooms/Areas Tables R-2 & H-2 Bases

2. RA-EP-02861 Radiological Incidents

3. NEI 99-01 AA3 EAL BASES DOCUMENT Rev. 0 Page 107 of 258

ATTACHMENT 1 EAL Bases Category E - Dry Fuel Storage Facility (DFSF)

EAL Group: ANY (EALs in this category are applicable to ANY plant condition, hot or cold)

A dry fuel storage facility (DFSF) 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.

An Unusual Event is declared based on the occurrence of an event of sufficient magnitude that a loaded cask CONFINEMENT BOUNDARY is damaged or violated.

Minor surface damage that does not affect storage cask/canister boundary is excluded from the scope of these EALs.

EAL BASES DOCUMENT Rev. 0 Page 108 of 258

ATTACHMENT 1 EAL Bases Category: E - DFSF Sub-category: None Initiating Condition: Damage to a loaded cask CONFINEMENT BOUNDARY EAL:

EU1.1 Notification of Unusual Event Damage to a loaded canister CONFINEMENT BOUNDARY as indicated by an on-contact radiation reading on the surface of a loaded spent fuel HSM cask > EITHER of the following:

100 mrem/hr (neutron + gamma) on the HSM cask wall or roof 100 mrem/hr (neutron + gamma) on the center of the HSM cask door Mode Applicability:

ALL Basis:

This IC addresses an event that results in damage to the CONFINEMENT BOUNDARY of a storage cask containing spent fuel. It applies to irradiated fuel that is licensed for dry storage beginning at the point that the loaded storage cask is sealed. The issues of concern are the creation of a potential or actual release path to the environment, degradation of one or more fuel assemblies due to environmental factors, and configuration changes, which could cause challenges in removing the cask or fuel from storage.

The existence of damage is determined by radiological survey. The technical specification multiple of 2 times, which is also used in Recognition Category R IC RU1, is used here to distinguish between non-emergency and emergency conditions. The emphasis for this classification is the degradation in the level of safety of the spent fuel cask and not the magnitude of the associated dose or dose rate. It is recognized that in the case of extreme damage to a loaded cask, the fact that the on-contact dose rate limit is exceeded may be determined based on measurement of a dose rate at some distance from the cask.

Security-related events for DFSFs are covered under ICs HU1 and HA1.

The DBNPS DFSF utilizes one design for dry spent fuel storage: the Nuclear Horizontal Modular Storage (NUHOMS) system.

The system consists of Horizontal Storage Modules (HSM) that contains dry shielded canisters (DSCs). The DSC serves as the CONFINEMENT BOUNDARY. The DSC is welded and designed to provide confinement of all radionuclides under normal, off normal, and accident conditions.

EAL BASES DOCUMENT Rev. 0 Page 109 of 258

ATTACHMENT 1 EAL Bases CONFINEMENT BOUNDARY is defined as the barrier(s) between areas containing radioactive substances and the environment. Therefore, damage to a CONFINEMENT BOUNDARY must be a confirmed physical breach between the spent fuel and the environment for the DSC.

The values shown represent 2 times the limits specified in the DFSF Site CSAR for radiation external to a loaded DSC (ref. 1).

Basis Reference(s):

1. Site CSAR NUH-003 Table 7.3-2 Shielding Analysis Results for 5 year Cooled Fuel NUHOMS-24P System

3. NEI 99-01 E-HU1 EAL BASES DOCUMENT Rev. 0 Page 110 of 258

ATTACHMENT 1 EAL Bases Category C - Cold Shutdown / Refueling System Malfunction EAL Group: Cold Conditions (RCS temperature 200ºF); 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 (5 - Cold Shutdown, 6 - Refueling, D - Defueled).

The events of this category pertain to the following subcategories:

1. RCS Level RCS water level is directly related to the status of adequate core cooling and, therefore, fuel clad integrity.

2. Loss of Essential AC Power Loss of essential plant electrical power can compromise plant SAFETY SYSTEM operability including decay heat removal and emergency core cooling systems, which may be necessary to ensure fission product barrier integrity. This category includes loss of onsite and offsite power sources for 4160 VAC essential buses.

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

4. Loss of Essential DC Power Loss of emergency plant electrical power can compromise plant SAFETY SYSTEM operability including decay heat removal and emergency core cooling systems, which may be necessary to ensure fission product barrier integrity. This category includes loss of power to or degraded voltage on the 125 VDC buses.

EAL BASES DOCUMENT Rev. 0 Page 111 of 258

ATTACHMENT 1 EAL Bases

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

6. Hazardous Event Affecting SAFETY SYSTEMS Certain hazardous natural and technological events may result in visible damage to or degraded performance of SAFETY SYSTEMS warranting classification.

EAL BASES DOCUMENT Rev. 0 Page 112 of 258

ATTACHMENT 1 EAL Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - RCS Level Initiating Condition: Loss of RCS inventory affecting fuel clad integrity with containment challenged EAL:

CG1.1 General Emergency RCS level cannot be monitored for 30 min. (Note 1)

AND Core uncovery is indicated by ANY of the following:

UNPLANNED increase in Containment Sumps, Auxiliary Building Sumps, BWST or RCDT levels of sufficient magnitude to indicate core uncovery Containment Radiation Monitor (RE 4596A or B) reading > 16 R/hr Refueling Bridge Portable Area Radiation Monitor reading > 30 R/hr (when installed)

Erratic Source Range Monitor indication AND ANY Containment Challenge indication, Table C-1 Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Note 6: If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, declaration of a General Emergency is not required.

Table C-1 Containment Challenge Indications CONTAINMENT CLOSURE not established (Note 6)

Containment Hydrogen concentration > 4%

UNPLANNED rise in Containment pressure Mode Applicability:

5 - Cold Shutdown, 6 - Refueling EAL BASES DOCUMENT Rev. 0 Page 113 of 258

ATTACHMENT 1 EAL Bases Basis:

This IC addresses the inability to restore and maintain RCS level above the top of active fuel with containment challenged. This condition represents actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity. Releases can be reasonably expected to exceed EPA PAG exposure levels offsite for more than the immediate site area.

Following an extended loss of core decay heat removal and inventory makeup, decay heat will cause reactor coolant boiling and a further reduction in RCS level. If RCS level cannot be restored, fuel damage is probable.

With CONTAINMENT CLOSURE not established, there is a high potential for a direct and unmonitored release of radioactivity to the environment. If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, then declaration of a General Emergency is not required.

The existence of an explosive mixture means, at a minimum, that the containment atmospheric hydrogen concentration is sufficient to support a hydrogen burn (i.e., at the lower deflagration limit). A hydrogen burn will raise containment pressure and could result in collateral equipment damage leading to a loss of containment integrity. It therefore represents a challenge to Containment integrity.

In the early stages of a core uncovery event, it is unlikely that hydrogen buildup due to a core uncovery could result in an explosive gas mixture in containment. If all installed hydrogen gas monitors are out-of-service during an event leading to fuel cladding damage, it may not be possible to obtain a containment hydrogen gas concentration reading, as ambient conditions within the containment will preclude personnel access. During periods when installed containment hydrogen gas monitors are out-of-service, operators may use the other listed indications to assess whether or not containment is challenged.

The 30-minute criterion is tied to a readily recognizable event start time (i.e., the total loss of ability to monitor level), and allows sufficient time to monitor, assess and correlate reactor and plant conditions to determine if core uncovery has actually occurred (i.e., to account for various accident progression and instrumentation uncertainties). It also allows sufficient time for performance of actions to terminate leakage, recover inventory control/makeup equipment and/or restore level monitoring.

The inability to monitor RCS level may be caused by instrumentation and/or power failures, or water level dropping below the range of available instrumentation. If water level cannot be monitored, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the RCS.

This EAL addresses concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal; SECY 91-283, Evaluation of Shutdown and Low Power Risk Issues; NUREG-1449, Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the United States; and NUMARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management.

EAL BASES DOCUMENT Rev. 0 Page 114 of 258

ATTACHMENT 1 EAL Bases The lowest measurable RCS level is the elevation of the RCS hot leg mid-loop (0.4 ft. on LI 10596). Therefore, RCS inventory loss relative to the RCS level elevation corresponding to the top of active fuel must be detected by indirect leakage indications (ref. 1, 2, 3). Sump level increases must be evaluated against other potential sources of leakage. If the make-up rate to the RCS unexplainably rises above the pre-established rate, a loss of RCS inventory may be occurring even if the source of the leakage cannot be immediately identified. Visual observation of leakage from systems connected to the RCS in areas outside the containment that cannot be isolated, or a sump sample verified by Chemistry, could also be indicative of a loss of RCS inventory (ref. 1).

In the Refueling Mode, as water level in the reactor vessel lowers, the dose rate above the core will increase. The dose rate due to this core shine should result in indications on installed and portable area radiation monitors. RE 4596A or B, Containment Radiation Monitors are located in the containment, a reading greater than 16 R/hr is indicative of core uncovery. The Refuel Bridge Portable Radiation Monitor, when installed during refueling operations, is designed to provide monitoring of radiation due to a fuel handling event or loss of shielding during refueling operations. If this radiation monitor reaches and exceeds 30 R/hr, a loss of inventory indicative of core uncovery has occurred (ref. 2, 3).

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

Three conditions are associated with a challenge to containment integrity:

CONTAINMENT CLOSURE is not established (Ref. 4).

In the early stages of a core uncovery event, it is unlikely that hydrogen buildup due to a core uncovery could result in an explosive mixture of dissolved gases in the containment. However, containment monitoring and/or sampling should be performed to verify this assumption and a General Emergency declared if it is determined that an explosive mixture exists. An explosive mixture can be formed when hydrogen gas concentration in the containment atmosphere is greater than 4% by volume in the presence of oxygen. Hydrogen monitors, although available at all times, are not in service during normal operations. They are started per DB-OP-02000 Table 3. Action is required if the measured concentration reaches 0.6% (notify TSC) and 3% (notify TSC) (ref 5).

ANY UNPLANNED increase in containment pressure in the Cold Shutdown or Refueling mode indicates a potential loss of CONTAINMENT CLOSURE capability. UNPLANNED containment pressure increases indicates CONTAINMENT CLOSURE cannot be assured and the containment cannot be relied upon as a barrier to fission product release.

EAL BASES DOCUMENT Rev. 0 Page 115 of 258

ATTACHMENT 1 EAL Bases Basis Reference(s):

1. DB-OP-02527 Loss of Decay Heat Removal

2. EP-EALCALC-DB-0704 Radiation Monitor Readings for Core Uncovery During Refueling

3. DB-HP-01152 Performance of High Exposure Work

4. DB-OP-06904, Shutdown Operations

5. DB-OP-02000 Table 3, Containment Monitoring and Control

6. NEI 99-01 CG1 EAL BASES DOCUMENT Rev. 0 Page 116 of 258

ATTACHMENT 1 EAL Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - RCS Level Initiating Condition: Loss of RCS inventory affecting core decay heat removal capability EAL:

CS1.1 Site Area Emergency RCS level cannot be monitored for 30 min. (Note 1)

AND Core uncovery is indicated by ANY of the following:

UNPLANNED increase in Containment Sumps, Auxiliary Building Sumps, BWST or RCDT levels of sufficient magnitude to indicate core uncovery Containment Radiation Monitor (RE 4596A or B) reading > 16 R/hr Refueling Bridge Portable Area Radiation Monitor reading > 30 R/hr (when installed)

Erratic Source Range Monitor indication Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Mode Applicability:

5 - Cold Shutdown, 6 - Refueling Basis:

This IC addresses a significant and prolonged loss of RCS inventory control and makeup capability leading to IMMINENT fuel damage. The lost inventory may be due to a RCS component failure, a loss of configuration control or prolonged boiling of reactor coolant.

These conditions entail major failures of plant functions needed for protection of the public and thus warrant a Site Area Emergency declaration.

Following an extended loss of core decay heat removal and inventory makeup, decay heat will cause reactor coolant boiling and a further reduction in RCS level. If RCS level cannot be restored, fuel damage is probable.

The 30-minute criterion is tied to a readily recognizable event start time (i.e., the total loss of ability to monitor level), and allows sufficient time to monitor, assess and correlate reactor and plant conditions to determine if core uncovery has actually occurred (i.e., to account for various accident progression and instrumentation uncertainties). It also allows sufficient time for performance of actions to terminate leakage, recover inventory control/makeup equipment and/or restore level monitoring.

EAL BASES DOCUMENT Rev. 0 Page 117 of 258

ATTACHMENT 1 EAL Bases The inability to monitor RCS level may be caused by instrumentation and/or power failures, or water level dropping below the range of available instrumentation. If water level cannot be monitored, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the RCS.

This EAL addresses concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal; SECY 91-283, Evaluation of Shutdown and Low Power Risk Issues; NUREG-1449, Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the United States; and NUMARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management.

Escalation of the emergency classification level would be via IC CG1 or RG1.

The lowest measurable RCS level is the elevation of the RCS hot leg mid-loop (0.4 ft. on LI 10596). Therefore, RCS inventory loss relative to the RCS level elevation corresponding to the top of active fuel must be detected by indirect leakage indications (ref. 1, 2, 3). Sump level increases must be evaluated against other potential sources of leakage. If the make-up rate to the RCS unexplainably rises above the pre-established rate, a loss of RCS inventory may be occurring even if the source of the leakage cannot be immediately identified. Visual observation of leakage from systems connected to the RCS in areas outside the containment that cannot be isolated, or a sump sample verified by Chemistry, could also be indicative of a loss of RCS inventory (ref. 1).

In the Refueling Mode, as water level in the reactor vessel lowers, the dose rate above the core will increase. The dose rate due to this core shine should result in indications on installed and portable area radiation monitors. RE 4596A or B, Containment Radiation Monitors are located in the containment, a reading greater than 16 R/hr is indicative of core uncovery. The Refuel Bridge Portable Radiation Monitor, when installed during refueling operations, is designed to provide monitoring of radiation due to a fuel handling event or loss of shielding during refueling operations. If this radiation monitor reaches and exceeds 30 R/hr, a loss of inventory indicative of core uncovery has occurred (ref. 2, 3).

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

Basis Reference(s):

1. DB-OP-02527 Loss of Decay Heat Removal

2. EP-EALCALC-DB-0704 Radiation Monitor Readings for Core Uncovery During Refueling

3. DB-HP-01152 Performance of High Exposure Work

4. NEI 99-01 CS1 EAL BASES DOCUMENT Rev. 0 Page 118 of 258

ATTACHMENT 1 EAL Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - RCS Level Initiating Condition: Loss of RCS inventory EAL:

CA1.1 Alert Loss of RCS inventory as indicated by RCS level 0.4 ft. (LI 10596)

Mode Applicability:

5 - Cold Shutdown, 6 - Refueling Basis:

This IC addresses conditions that are precursors to a loss of the ability to adequately cool irradiated fuel (i.e., a precursor to a challenge to the fuel clad barrier). This condition represents a potential substantial reduction in the level of plant safety.

For this EAL, a lowering of RCS water level at or below 0.4 ft. indicates that operator actions have not been successful in restoring and maintaining RCS water level. The heat-up rate of the coolant will increase as the available water inventory is reduced. A continuing decrease in water level will lead to core uncovery.

Although related, this EAL is concerned with the loss of RCS inventory and not the potential concurrent effects on systems needed for decay heat removal (e.g., loss of a Decay Heat Removal suction point). An increase in RCS temperature caused by a loss of decay heat removal capability is evaluated under IC CA3.

If RCS water level continues to lower, then escalation to Site Area Emergency would be via IC CS1.

RCS level cannot be measured below the 571 feet elevation (0.4 feet on LI 10596) which is centerline of the hot leg inlet. Should RCS level drop below this point it is assumed water level cannot be monitored other than visually. Continued loss of RCS inventory may result in DHR pump cavitation (ref. 1).

Basis Reference(s):

1. DB-OP-02527 Loss of Decay Heat Removal

2. NEI 99-01 CA1 EAL BASES DOCUMENT Rev. 0 Page 119 of 258

ATTACHMENT 1 EAL Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - RCS Level Initiating Condition: Loss of RCS inventory EAL:

CA1.2 Alert RCS level cannot be monitored for 15 min. (Note 1)

AND EITHER UNPLANNED increase in Containment Sumps, Auxiliary Building Sumps, BWST or RCDT due to a loss of RCS inventory Visual observation of UNISOLABLE RCS leakage Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Mode Applicability:

5 - Cold Shutdown, 6 - Refueling Basis:

This IC addresses conditions that are precursors to a loss of the ability to adequately cool irradiated fuel (i.e., a precursor to a challenge to the fuel clad barrier). This condition represents a potential substantial reduction in the level of plant safety.

For this EAL, the inability to monitor RCS level may be caused by instrumentation and/or power failures, or water level dropping below the range of available instrumentation. If water level cannot be monitored, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the RCS.

The 15-minute duration for the loss of level indication was chosen because it is half of the EAL duration specified in IC CS1.

If the RCS inventory level continues to lower, then escalation to Site Area Emergency would be via IC CS1.

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

EAL BASES DOCUMENT Rev. 0 Page 120 of 258

ATTACHMENT 1 EAL Bases RCS level in the Refueling mode is normally monitored using the following instruments:

• LI 10596

• LI10577 A and B

• Clear tubing used for manometer level indication at the RCS cold legs.

RCS level indications (LI 10596, LI 10577A and B and cold leg tubing) provide accurate indication of water level when the RCS is at atmospheric pressure. The reactor vessel flange is at 577 ft. 7 in. (80 in. indicated) (ref. 1).

In this EAL, all RCS water level indication would be unavailable for greater than 15-minutes, and the RCS inventory loss must be detected by indirect leakage indications (ref. 1). Sump and tank level increases must be evaluated against other potential sources of leakage. If the make-up rate to the RCS unexplainably rises above the pre-established rate, a loss of RCS inventory may be occurring even if the source of the leakage cannot be immediately identified.

Visual observation of leakage from systems connected to the RCS that cannot be isolated, or a sump sample verified by Chemistry, could also be indicative of a loss of RCS inventory.

Basis Reference(s):

1. DB-OP-06904 Shutdown Operations

2. NEI 99-01 CA1 EAL BASES DOCUMENT Rev. 0 Page 121 of 258

ATTACHMENT 1 EAL Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - RCS Level Initiating Condition: UNPLANNED loss of RCS inventory for 15 minutes or longer EAL:

CU1.1 Unusual Event UNPLANNED loss of reactor coolant results in RCS level less than a required lower limit for 15 min. (Note 1)

Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Mode Applicability:

5 - Cold Shutdown, 6 - Refueling Basis:

This IC addresses the inability to restore and maintain water level to a required minimum level (or the lower limit of a level band), or a loss of the ability to monitor RCS level concurrent with indications of coolant leakage. Either of these conditions is considered to be a potential degradation of the level of safety of the plant.

RCS draining evolutions that decrease RCS water inventory are carefully planned and controlled. An UNPLANNED event that results in water level decreasing below a procedurally required limit warrants the declaration of an Unusual Event due to the reduced water inventory that is available to keep the core covered.

This EAL recognizes that the minimum required RCS level can change several times during the course of a refueling outage as different plant configurations and system lineups are implemented. This EAL is met if the minimum level, specified for the current plant conditions, cannot be maintained for 15 minutes or longer. The minimum level is typically specified in the applicable operating procedure but may be specified in another controlling document.

The 15-minute threshold duration allows sufficient time for prompt operator actions to restore and maintain the expected water level. This criterion excludes transient conditions causing a brief lowering of water level.

Continued loss of RCS inventory may result in escalation to the Alert emergency classification level via either IC CA1 or CA3.

EAL BASES DOCUMENT Rev. 0 Page 122 of 258

ATTACHMENT 1 EAL Bases With the plant in Cold Shutdown, RCS water level is normally maintained above a lower level band limit (ref. 1, 2), typically above the pressurizer low level setpoint. However, if RCS level is being controlled below the pressurizer low level setpoint, or if level is being maintained in a designated band in the reactor vessel it is the inability to maintain level above the low end of the designated control band due to a loss of inventory resulting from a leak in the RCS that is the concern.

With the plant in Refueling mode, RCS water level is normally maintained at or above the reactor vessel flange, (Technical Specification LCO 3.9.6 requires at least 23 ft. of water above the top of the reactor vessel flange in the refueling canal when irradiated fuel is moved in containment). The reactor vessel flange is at 577 ft. 7 in. (80 in. indicated) (ref. 3, 4).

In Cold Shutdown mode, the RCS will normally be intact and standard RCS level monitoring means are available. RCS level in the Refueling mode is normally monitored using the following instruments:

• LI 10596

• LI 10577 A and B

• Clear tubing used for manometer level indication at the RCS cold legs.

RCS level indications (LI 10596, LI 10577A and B and cold leg tubing) provide accurate indication of water level when the RCS is at atmospheric pressure (ref. 3).

Basis Reference(s):

1. DB-OP-02513 Pressurizer System Abnormal Operation

2. DB-OP-02004 4-2-E PZR LVL LO

3. DB-OP-06904 Shutdown Operations

4. DBNPS Technical Specifications Section 3.9.6 Refueling Canal Water Level

5. DB-OP-02527 Loss of Decay Heat Removal

6. NEI 99-01 CU1 EAL BASES DOCUMENT Rev. 0 Page 123 of 258

ATTACHMENT 1 EAL Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - RCS Level Initiating Condition: UNPLANNED loss of RCS inventory for 15 minutes or longer EAL:

CU1.2 Unusual Event RCS level cannot be monitored AND EITHER UNPLANNED increase in Containment Sumps, Auxiliary Building Sumps, BWST or RCDT due to a loss of RCS inventory Visual observation of UNISOLABLE RCS leakage Mode Applicability:

5 - Cold Shutdown, 6 - Refueling Basis:

This IC addresses the inability to restore and maintain water level to a required minimum level (or the lower limit of a level band), or a loss of the ability to monitor RCS level concurrent with indications of coolant leakage. Either of these conditions is considered to be a potential degradation of the level of safety of the plant.

RCS draining evolutions that decrease RCS water inventory are carefully planned and controlled. An UNPLANNED event that results in water level decreasing below a procedurally required limit warrants the declaration of an Unusual Event due to the reduced water inventory that is available to keep the core covered.

This EAL addresses a condition where all means to determine RCS level have been lost. In this condition, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the RCS.

Continued loss of RCS inventory may result in escalation to the Alert emergency classification level via either IC CA1 or CA3.

EAL BASES DOCUMENT Rev. 0 Page 124 of 258

ATTACHMENT 1 EAL Bases In Cold Shutdown mode, the RCS will normally be intact and standard RCS level monitoring means are available. RCS level in the Refueling mode is normally monitored using the following instruments:

• LI 10596

• LI10577 A and B

• Clear tubing used for manometer level indication at the RCS cold legs.

RCS level indications (LI 10596, LI 10577A and B and cold leg tubing) provide accurate indication of water level when the RCS is at atmospheric pressure. The reactor vessel flange is at 577 ft. 7 in. (80 in. indicated) (ref. 1).

In this EAL, all water level indication is unavailable and the RCS inventory loss must be detected by indirect leakage indications (ref. 1). 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. If the make-up rate to the RCS unexplainably rises above the pre-established rate, a loss of RCS inventory may be occurring even if the source of the leakage cannot be immediately identified. Visual observation of leakage from systems connected to the RCS that cannot be isolated, or a sump sample verified by Chemistry, could also be indicative of a loss of RCS inventory.

Basis Reference(s):

1. DB-OP-06904 Shutdown Operations

2. NEI 99-01 CU1 EAL BASES DOCUMENT Rev. 0 Page 125 of 258

ATTACHMENT 1 EAL Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 2 - Loss of Essential AC Power Initiating Condition: Loss of ALL offsite and ALL onsite AC power to essential buses for 15 minutes or longer EAL:

CA2.1 Alert Loss of ALL offsite and ALL onsite AC power capability to essential 4160V buses C1 and D1 for 15 min. (Note 1)

Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Mode Applicability:

5 - Cold Shutdown, 6 - Refueling, D - Defueled Basis:

This IC addresses a total loss of AC power that compromises the performance of ALL SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink.

When in the Cold Shutdown, Refueling, or Defueled mode, this condition is not classified as a Site Area Emergency because of the increased time available to restore an essential bus to service. Additional time is available due to the reduced core decay heat load, and the lower temperatures and pressures in various plant systems. Thus, when in these modes, this condition represents an actual or potential substantial degradation of the level of safety of the plant.

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

Escalation of the emergency classification level would be via IC CS1 or RS1.

The 4160 VAC System provides the power requirements for operation and safe shutdown of the plant. The essential switchgear are buses C1 and D1 (ref. 1).

EAL BASES DOCUMENT Rev. 0 Page 126 of 258

ATTACHMENT 1 EAL Bases The essential buses during plant operation are normally powered from the 13.8KV offsite power system through their respective 13.8KV/4160V bus tie transformers, via the Unit Auxiliary Transformer (X11). In cold operating modes, the essential buses may be back-fed via the X11 and Main Transformer provided the main generator lead disconnect links are removed (ref. 1, 2). Credit for the X11 back-feed can only be taken if already aligned, as it takes greater than 15 minutes to align.

A standby source of offsite power to each 4160V essential bus is provided from the 13.8KV offsite power system via two separate and independent 13.8KV/4160V Startup Transformers (X01 & X02). Normally each startup transformer serves as a reserve power source for only one essential bus. However, a single startup transformer can be aligned to power both essential buses (ref. 1, 2).

Each essential bus has a diesel generator (EDG1 & EDG2) to supply an onsite emergency source of power to safe shutdown loads in the event of a loss of the normal power source or loss of off-site power (ref. 1).

An Alternate AC power source, the SBO Diesel Generator, is located onsite. This onsite AC power source can be started from the Control Room and be loaded within 10 minutes of a SBO (ref. 1).

This cold condition EAL is equivalent to the hot condition loss of ALL offsite AC power EAL SS1.1.

Basis Reference(s):

1. UFSAR Section 8.0 Electric Power

2. DB-OP-02521 Loss of AC Bus Power Sources

3. NEI 99-01 CA2 EAL BASES DOCUMENT Rev. 0 Page 127 of 258

ATTACHMENT 1 EAL Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 2 - Loss of Essential AC Power Initiating Condition: Loss of ALL but one AC power source to essential buses for 15 minutes or longer EAL:

CU2.1 Unusual Event AC power capability, Table C-2, to essential 4160V buses C1 and D1 reduced to a single power source for 15 min. (Note 1)

AND ANY additional single power source failure will result in loss of ALL AC power to SAFETY SYSTEMS Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Table C-2 Offsite/Onsite AC Power Sources Offsite:

X01 X02 X11 (back-fed via Main Transformer if already aligned)

Onsite:

EDG1 EDG2 SBODG Mode Applicability:

5 - Cold Shutdown, 6 - Refueling, D - Defueled EAL BASES DOCUMENT Rev. 0 Page 128 of 258

ATTACHMENT 1 EAL Bases Basis:

This IC describes a significant degradation of offsite and onsite AC power sources such that any additional single failure would result in a loss of ALL AC power to SAFETY SYSTEMS. In this condition, the sole AC power source may be powering one, or more than one, train of safety-related equipment.

When in the Cold Shutdown, Refueling, or Defueled mode, this condition is not classified as an Alert because of the increased time available to restore another power source to service.

Additional time is available due to the reduced core decay heat load, and the lower temperatures and pressures in various plant systems. Thus, when in these modes, this condition is considered to be a potential degradation of the level of safety of the plant.

An AC power source is a source recognized in AOPs and EOPs, and capable of supplying required power to an essential bus. Some examples of this condition are presented below.

A loss of ALL offsite power with a concurrent failure of ALL but one essential power source (e.g., an onsite diesel generator)

A loss of essential power sources (e.g., onsite diesel generators) with a single train of essential buses being fed from an offsite power source Fifteen minutes was selected as a threshold to exclude transient or momentary losses of power.

The subsequent loss of the remaining single power source would escalate the event to an Alert in accordance with IC CA2.

The 4160 VAC System provides the power requirements for operation and safe shutdown of the plant. The essential switchgear are buses C1 and D1 (ref. 1).

The essential buses during plant operation are normally powered from the 13.8KV offsite power system through their respective 13.8KV/4160V bus tie transformers, via the Unit Auxiliary Transformer (X11). In cold operating modes, the essential buses may be back-fed via the X11 and Main Transformer provided the main generator lead disconnect links are removed (ref. 1, 2). Credit for the X11 back-feed can only be taken if already aligned, as it takes greater than 15 minutes to align.

A standby source of offsite power to each 4160V essential bus is provided from the 13.8KV offsite power system via two separate and independent 13.8KV/4160V Startup Transformers (X01 & X02). Normally each startup transformer serves as a reserve power source for only one essential bus. However, a single startup transformer can be aligned to power both essential buses (ref. 1, 2).

Each essential bus has a diesel generator (EDG1 & EDG2) to supply an onsite emergency source of power to safe shutdown loads in the event of a loss of the normal power source or loss of off-site power (ref. 1).

EAL BASES DOCUMENT Rev. 0 Page 129 of 258

ATTACHMENT 1 EAL Bases An Alternate AC power source, the SBO Diesel Generator, is located onsite. This onsite AC power source can be started from the Control Room and be loaded within 10 minutes of a SBO (ref. 1).

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

Basis Reference(s):

1. UFSAR Section 8.0 Electric Power

2. DB-OP-02521 Loss of AC Bus Power Sources

3. NEI 99-01 CU2 EAL BASES DOCUMENT Rev. 0 Page 130 of 258

ATTACHMENT 1 EAL Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 3 - RCS Temperature Initiating Condition: Inability to maintain plant in cold shutdown EAL:

CA3.1 Alert UNPLANNED increase in RCS temperature to > 200°F for > Table C-3 duration (Note 1, 10)

OR UNPLANNED RCS pressure increase > 10 psig due to a loss of RCS cooling (This EAL does not apply during water-solid plant conditions).

Note 1: The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.

Note: 10 In the absence of reliable RCS temperature indication caused by the loss of decay heat removal capability, classification should be based on the RCS pressure increase criteria when the RCS is intact in Mode 5 or based on time to boil data when in Mode 6 or the RCS is not intact in Mode 5.

Table C-3: RCS Heat-up Duration Thresholds Containment Closure RCS Status Heat-up Duration Status Intact (but not reduced N/A 60 min.*

inventory)

Not intact established 20 min.*

OR not established 0 min.

At reduced inventory

• If an RCS heat removal system is in operation within this time frame and RCS temperature is being reduced, the EAL is not applicable.

Mode Applicability:

5 - Cold Shutdown, 6 - Refueling Basis:

This IC addresses conditions involving a loss of decay heat removal capability or an addition of heat to the RCS in excess of that which can currently be removed. Either condition represents an actual or potential substantial degradation of the level of safety of the plant.

EAL BASES DOCUMENT Rev. 0 Page 131 of 258

ATTACHMENT 1 EAL Bases A momentary UNPLANNED excursion above the Technical Specification cold shutdown temperature limit when the heat removal function is available does not warrant a classification.

The RCS Heat-up Duration Thresholds table addresses an increase in RCS temperature when CONTAINMENT CLOSURE is established but the RCS is not intact, or RCS inventory is reduced (e.g., mid-loop operation). The 20-minute criterion was included to allow time for operator action to address the temperature increase.

The RCS Heat-up Duration Thresholds table also addresses an increase in RCS temperature with the RCS INTACT. The status of CONTAINMENT CLOSURE is not crucial in this condition since the intact RCS is providing a high pressure barrier to a fission product release.

The 60-minute time frame should allow sufficient time to address the temperature increase without a substantial degradation in plant safety.

Finally, in the case where there is an increase in RCS temperature, the RCS is not intact or is at reduced inventory, and CONTAINMENT CLOSURE is not established, no heat-up duration is allowed (i.e., 0 minutes). This is because 1) the evaporated reactor coolant may be released directly into the containment atmosphere and subsequently to the environment, and

2) there is reduced reactor coolant inventory above the top of irradiated fuel.

The RCS pressure increase threshold provides a pressure-based indication of RCS heat-up in the absence of RCS temperature monitoring capability.

Escalation of the emergency classification level would be via IC CS1 or RS1.

In the absence of reliable RCS temperature indication caused by the loss of decay heat removal capability, classification should be based on the RCS pressure increase criteria when the RCS is intact in Mode 5 or based on time to boil data when in Mode 6 or the RCS is not intact in Mode 5.

Several instruments are capable of providing indication of RCS temperature with respect to the Technical Specification cold shutdown temperature limit (200°F, ref. 1) including:

Selected Incore thermocouples TIRC4B2 Reactor Coolant T-Cold Wide Range (Loop 1)

TIRC4A2 Reactor Coolant T-Cold Wide Range (Loop 2)

TIRC3B5 or TIRCB6 Reactor Coolant T-Hot (Loop 1)

TIRC3A5 or TIRCA6 Reactor Coolant T-Hot (Loop 2)

DHR Display on SPDS EAL BASES DOCUMENT Rev. 0 Page 132 of 258

ATTACHMENT 1 EAL Bases A 10 psig RCS pressure increase can be read on various instruments such as (ref. 2):

PI 6365B, Loop 1 Pressure PI 6365A, Loop 2 Pressure PRS RC2A1, Loop 2 Wide Range Pressure PI RC2A6, Low Range Pressure The RCS is considered in reduced inventory if (ref. 3):

The Reactor Coolant System is not full (loops not filled and the RCS is incapable of natural circulation), OR Both steam generators are not available as a heat sink, OR The refueling canal level is less than a stable level specified in the Shutdown Defense in Depth Review, typically 23 feet, with the head removed.

Basis Reference(s):

1. DBNPS Technical Specifications Table 1.1-1

2. DB-OP-02527 Loss of Decay Heat Removal

3. NG-DB-00117 Shutdown Defense In Depth Assessment

4. NEI 99-01 CA3 EAL BASES DOCUMENT Rev. 0 Page 133 of 258

ATTACHMENT 1 EAL Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 3 - RCS Temperature Initiating Condition: UNPLANNED increase in RCS temperature EAL:

CU3.1 Unusual Event UNPLANNED increase in RCS temperature to > 200°F due to loss of decay heat removal capability Mode Applicability:

5 - Cold Shutdown, 6 - Refueling Basis:

This IC addresses an UNPLANNED increase in RCS temperature above the Technical Specification cold shutdown temperature limit and represents a potential degradation of the level of safety of the plant. If the RCS is not intact and CONTAINMENT CLOSURE is not established during this event, the Emergency Director should also refer to IC CA3.

A momentary UNPLANNED excursion above the Technical Specification cold shutdown temperature limit when the heat removal function is available does not warrant a classification.

This EAL involves a loss of decay heat removal capability, or an addition of heat to the RCS in excess of that which can currently be removed, such that reactor coolant temperature cannot be maintained below the cold shutdown temperature limit specified in Technical Specifications.

During this condition, there is no immediate threat of fuel damage because the core decay heat load has been reduced since the cessation of power operation.

During an outage, the level in the reactor vessel will normally be maintained at or above the reactor vessel flange. Refueling evolutions that lower water level below the reactor vessel flange are carefully planned and controlled. A loss of forced decay heat removal at reduced inventory may result in a rapid increase in reactor coolant temperature depending on the time after shutdown.

Escalation to Alert would be via IC CA1 based on an inventory loss or IC CA3 based on exceeding plant configuration-specific time criteria.

EAL BASES DOCUMENT Rev. 0 Page 134 of 258

ATTACHMENT 1 EAL Bases Several instruments are capable of providing indication of RCS temperature with respect to the Technical Specification cold shutdown temperature limit (200°F, ref. 1) including:

Selected Incore thermocouples TIRC4B2 Reactor Coolant T-Cold Wide Range (Loop 1)

TIRC4A2 Reactor Coolant T-Cold Wide Range (Loop 2)

TIRC3B5 or TIRCB6 Reactor Coolant T-Hot (Loop 1)

TIRC3A5 or TIRCA6 Reactor Coolant T-Hot (Loop 2)

DHR Display on SPDS In the absence of reliable RCS temperature indication caused by a loss of decay heat removal capability, classification should be based on EAL CU3.2 should RCS level indication be subsequently lost.

Basis Reference(s):

1. DBNPS Technical Specifications Table 1.1-1 Modes

2. DB-OP-02527 Loss of Decay Heat Removal

3. NEI 99-01 CU3 EAL BASES DOCUMENT Rev. 0 Page 135 of 258

ATTACHMENT 1 EAL Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 3 - RCS Temperature Initiating Condition: UNPLANNED increase in RCS temperature EAL:

CU3.2 Unusual Event Loss of ALL RCS temperature and RCS level indication for 15 min. (Note 1)

Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Mode Applicability:

5 - Cold Shutdown, 6- Refueling Basis:

This EAL addresses the inability to determine RCS temperature and level, and represents a potential degradation of the level of safety of the plant. If the RCS is not intact and CONTAINMENT CLOSURE is not established during this event, the Emergency Director should also refer to IC CA3.

This EAL reflects a condition where there has been a significant loss of instrumentation capability necessary to monitor RCS conditions and operators would be unable to monitor key parameters necessary to assure core decay heat removal. During this condition, there is no immediate threat of fuel damage because the core decay heat load has been reduced since the cessation of power operation.

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

Escalation to Alert would be via IC CA1 based on an inventory loss or IC CA3 based on exceeding plant configuration-specific time criteria.

EAL BASES DOCUMENT Rev. 0 Page 136 of 258

ATTACHMENT 1 EAL Bases Several instruments are capable of providing indication of RCS temperature with respect to the Technical Specification cold shutdown temperature limit (200°F, ref. 1) including:

Selected Incore thermocouples TIRC4B2 Reactor Coolant T-Cold Wide Range (Loop 1)

TIRC4A2 Reactor Coolant T-Cold Wide Range (Loop 2)

TIRC3B5 or TIRCB6 Reactor Coolant T-Hot (Loop 1)

TIRC3A5 or TIRCA6 Reactor Coolant T-Hot (Loop 2)

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

RCS level in the Refueling mode is normally monitored using the following instruments:

• LI 10596

• LI10577 A and B

• Clear tubing used for manometer level indication at the RCS cold legs.

RCS level indications (LI 10596, LI 10577A and B and cold leg tubing) provide accurate indication of water level when the RCS is at atmospheric pressure. However, there is no redundant means of RCS level indication when Mode 6 is entered (ref. 2).

Basis Reference(s):

1. DBNPS Technical Specifications Table 1.1-1

2. DB-OP-02527 Loss of Decay Heat Removal

3. NEI 99-01 CU3 EAL BASES DOCUMENT Rev. 0 Page 137 of 258

ATTACHMENT 1 EAL Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 4 - Loss of Essential DC Power Initiating Condition: Loss of essential DC power for 15 minutes or longer EAL:

CU4.1 Unusual Event

< 105 VDC voltage indications on Technical Specification required essential 125 VDC distribution panels for 15 min. (Note 1)

Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Mode Applicability:

5 - Cold Shutdown, 6 - Refueling Basis This IC addresses a loss of essential DC power, which compromises the ability to monitor and control operable SAFETY SYSTEMS when the plant is in the Cold Shutdown or Refueling mode. In these modes, the core decay heat load has been significantly reduced, and coolant system temperatures and pressures are lower; these conditions increase the time available to restore an essential DC bus to service. Thus, this condition is considered to be a potential degradation of the level of safety of the plant.

As used in this EAL, required means the essential DC buses necessary to support operation of the in-service, or operable, train or trains of SAFETY SYSTEM equipment. For example, if Train A is out-of-service (inoperable) for scheduled outage maintenance work and Train B is in-service (operable), then a loss of essential DC power affecting Train B would require the declaration of an Unusual Event. A loss of essential DC power to Train A would not warrant an emergency classification.

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

Depending upon the event, escalation of the emergency classification level would be via IC CA1 or CA3, or an IC in Recognition Category R.

EAL BASES DOCUMENT Rev. 0 Page 138 of 258

ATTACHMENT 1 EAL Bases The DC electrical distribution subsystem consists of two trains, designated Train 1 and Train 2.

Each train consists of a 250/125 VDC motor control center (MCC), and each 250/125 VDC MCC consists of two 125 VDC buses (one positive and one negative) that can be powered from a polarity specific battery or battery charger. The two DC buses then supply a specific Essential DC Distribution Panel (D1P and D1N for DCMCC 1, and D2P and D2N for DCMCC2). The four Essential DC Distribution Panels independently supply DC electrical power to the four inverters, which in turn power the 120 VAC essential buses (ref. 1, 2, 3, 4, 5, 6).

The Class 1E DC loads have an operating voltage range of 105 to 135 volts. The minimum battery discharge voltage (requiring opening the degraded battery output breaker) is 105 VDC (ref. 1, 6).

This EAL is the cold condition equivalent of the hot condition loss of DC power EAL SS2.1.

Basis Reference(s):

1. DBNPS UFSAR Section 8.0 Electrical Power

2. DB-OP-02537 Loss of D1P and DAP

3. DB-OP-02538 Loss of D2P and DBP

4. DB-OP-02539 Loss of D1N and DAN

5. DB-OP-02540 Loss of D2N and DBN

6. System Description for 125/250 VDC and 120 V Instrumentation AC System

7. NEI 99-01 CU4 EAL BASES DOCUMENT Rev. 0 Page 139 of 258

ATTACHMENT 1 EAL Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 5 - Loss of Communications Initiating Condition: Loss of ALL onsite or offsite communications capabilities EAL:

CU5.1 Unusual Event Loss of ALL Table C-4 onsite communication methods Table C-4 Communication Methods System Onsite ORO NRC Public Address (Gaitronics) X Onsite Radios X Plant Telephones X X X Commercial Telephones X X X 4-Way Ringdown Circuit X NRC Emergency Telephone System (ETS) X Mode Applicability:

5 - Cold Shutdown, 6 - Refueling, D - Defueled Basis:

This IC addresses a significant loss of on-site or offsite communications capabilities. While not a direct challenge to plant or personnel safety, this event warrants prompt notifications to OROs and the NRC.

This IC should be assessed only when extraordinary means are being utilized to make communications possible (e.g., use of non-plant, privately owned equipment, relaying of on-site information via individuals or multiple radio transmission points, individuals being sent to offsite locations, etc.).

This EAL addresses a total loss of the communications methods used in support of routine plant operations.

EAL BASES DOCUMENT Rev. 0 Page 140 of 258

ATTACHMENT 1 EAL Bases Onsite/Offsite Response Organization (ORO)/NRC communications include one or more of the systems listed in Table C-4 (ref. 1, 2).

Public Address (Gaitronics) System The DBNPS public address system provides paging and party line communications between stations located throughout the plant. Inside and outside type wall and desk-mounted stations are used to communicate between roaming personnel and fixed work locations. Plant-wide instructions are issued using the paging feature.

On-site Radio System Radio systems can be used for communication among operators, off-site monitoring teams, the control room, security, TSC and EOF.

Plant Telephone System The DBNPS plant telephone system provides communication capability between telephone stations located within the plant by dialing the four-digit telephone station code as well as external or offsite calling capability.

Commercial Telephones Commercial telephone lines, which supply public telephone communications, are employed by DBNPS. The local service provider provides primary and secondary power for their lines at the Central Office.

4-Way Ringdown Circuit Dedicated ring down line that includes the State and County EOCs, the Ohio Highway Patrol Office, the Lucas County and Ottawa County Sheriff's dispatcher offices, the Emergency Operations Facility, and the Control Room.

NRC Emergency Telephone System The NRC uses a DBNPS dedicated telephone line, which allows direct telephone communications from the plant to NRC regional and national offices. The DBNPS communications line provides a link independent of the local public telephone network.

Telephones connected to this network are located in the DBNPS Control Room, Technical Support Center, and Emergency Operations Facility and can be used to establish NRC Emergency Notification System (ENS) and Health Physics Network (HPN) capability.

This EAL is the cold condition equivalent of the hot condition EAL SU7.1.

Basis Reference(s):

1. DBNPS UFSAR Section 9.5.2 Communications Systems

2. DBNPS Emergency Plan 7.6 Communications Systems

3. NEI 99-01 CU5 EAL BASES DOCUMENT Rev. 0 Page 141 of 258

ATTACHMENT 1 EAL Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 5 - Loss of Communications Initiating Condition: Loss of ALL onsite or offsite communications capabilities EAL:

CU5.2 Unusual Event Loss of ALL Table C-4 ORO communication methods Table C-4 Communication Methods System Onsite ORO NRC Public Address (Gaitronics) X Onsite Radios X Plant Telephones X X X Commercial Telephones X X X 4-Way Ringdown Circuit X NRC Emergency Telephone System (ETS) X Mode Applicability:

5 - Cold Shutdown, 6 - Refueling, D - Defueled Basis:

This IC addresses a significant loss of on-site or offsite communications capabilities. While not a direct challenge to plant or personnel safety, this event warrants prompt notifications to OROs and the NRC.

This IC should be assessed only when extraordinary means are being utilized to make communications possible (e.g., use of non-plant, privately owned equipment, relaying of on-site information via individuals or multiple radio transmission points, individuals being sent to offsite locations, etc.).

This EAL addresses a total loss of the communications methods used to notify all OROs of an emergency declaration. The OROs referred to here are the State, Lucas and Ottawa County EOCs.

EAL BASES DOCUMENT Rev. 0 Page 142 of 258

ATTACHMENT 1 EAL Bases Onsite/Offsite Response Organization (ORO)/NRC communications include one or more of the systems listed in Table C-4 (ref. 1, 2).

Public Address (Gaitronics) System The DBNPS public address system provides paging and party line communications between stations located throughout the plant. Inside and outside type wall and desk-mounted stations are used to communicate between roaming personnel and fixed work locations. Plant-wide instructions are issued using the paging feature.

On-site Radio System Radio systems can be used for communication among operators, off-site monitoring teams, the control room, security, TSC and EOF.

Plant Telephone System The DBNPS plant telephone system provides communication capability between telephone stations located within the plant by dialing the four-digit telephone station code as well as external or offsite calling capability.

Commercial Telephones Commercial telephone lines, which supply public telephone communications, are employed by DBNPS. The local service provider provides primary and secondary power for their lines at the Central Office.

4-Way Ringdown Circuit Dedicated ring down line that includes the State and County EOCs, the Ohio Highway Patrol Office, the Lucas County and Ottawa County Sheriff's dispatcher offices, the Emergency Operations Facility, and the Control Room.

NRC Emergency Telephone System The NRC uses a DBNPS dedicated telephone line, which allows direct telephone communications from the plant to NRC regional and national offices. The DBNPS communications line provides a link independent of the local public telephone network.

Telephones connected to this network are located in the DBNPS Control Room, Technical Support Center, and Emergency Operations Facility and can be used to establish NRC Emergency Notification System (ENS) and Health Physics Network (HPN) capability.

This EAL is the cold condition equivalent of the hot condition EAL SU7.2.

Basis Reference(s):

1. DBNPS UFSAR Section 9.5.2 Communications Systems

2. DBNPS Emergency Plan 7.6 Communications Systems

3. NEI 99-01 CU5 EAL BASES DOCUMENT Rev. 0 Page 143 of 258

ATTACHMENT 1 EAL Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 5 - Loss of Communications Initiating Condition: Loss of ALL onsite or offsite communications capabilities EAL:

CU5.3 Unusual Event Loss of ALL Table C-4 NRC communication methods Table C-4 Communication Methods System Onsite ORO NRC Public Address (Gaitronics) X Onsite Radios X Plant Telephones X X X Commercial Telephones X X X 4-Way Ringdown Circuit X NRC Emergency Telephone System (ETS) X Mode Applicability:

5 - Cold Shutdown, 6 - Refueling, D - Defueled Basis:

This IC addresses a significant loss of on-site or offsite communications capabilities. While not a direct challenge to plant or personnel safety, this event warrants prompt notifications to OROs and the NRC.

This IC should be assessed only when extraordinary means are being utilized to make communications possible (e.g., use of non-plant, privately owned equipment, relaying of on-site information via individuals or multiple radio transmission points, individuals being sent to offsite locations, etc.).

This EAL addresses a total loss of the communications methods used to notify the NRC of an emergency declaration.

EAL BASES DOCUMENT Rev. 0 Page 144 of 258

ATTACHMENT 1 EAL Bases Onsite/Offsite Response Organization (ORO)/NRC communications include one or more of the systems listed in Table C-4 (ref. 1, 2).

Public Address (Gaitronics) System The DBNPS public address system provides paging and party line communications between stations located throughout the plant. Inside and outside type wall and desk-mounted stations are used to communicate between roaming personnel and fixed work locations. Plant-wide instructions are issued using the paging feature.

On-site Radio System Radio systems can be used for communication among operators, off-site monitoring teams, the control room, security, TSC and EOF.

Plant Telephone System The DBNPS plant telephone system provides communication capability between telephone stations located within the plant by dialing the four-digit telephone station code as well as external or offsite calling capability.

Commercial Telephones Commercial telephone lines, which supply public telephone communications, are employed by DBNPS. The local service provider provides primary and secondary power for their lines at the Central Office.

4-Way Ringdown Circuit Dedicated ring down line that includes the State and County EOCs, the Ohio Highway Patrol Office, the Lucas County and Ottawa County Sheriff's dispatcher offices, the Emergency Operations Facility, and the Control Room.

NRC Emergency Telephone System The NRC uses a DBNPS dedicated telephone line, which allows direct telephone communications from the plant to NRC regional and national offices. The DBNPS communications line provides a link independent of the local public telephone network.

Telephones connected to this network are located in the DBNPS Control Room, Technical Support Center, and Emergency Operations Facility and can be used to establish NRC Emergency Notification System (ENS) and Health Physics Network (HPN) capability.

This EAL is the cold condition equivalent of the hot condition EAL SU7.3.

Basis Reference(s):

1. DBNPS UFSAR Section 9.5.2 Communications Systems

2. DBNPS Emergency Plan 7.6 Communications Systems

3. NEI 99-01 CU5 EAL BASES DOCUMENT Rev. 0 Page 145 of 258

ATTACHMENT 1 EAL Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 6 - Hazardous Event Affecting SAFETY SYSTEMS Initiating Condition: Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode EAL:

CA6.1 Alert The occurrence of ANY Table C-5 Hazardous Event AND EITHER:

Event has caused indications of degraded performance in at least one train of a SAFETY SYSTEM required for the current operating mode The event has caused VISIBLE DAMAGE to a SAFETY SYSTEM component or structure required for the current operating mode Table C-5 Hazardous Events Seismic event (earthquake)

Internal or external FLOODING event High winds or tornado strike FIRE EXPLOSION Other events with similar hazard characteristics as determined by the Emergency Director Mode Applicability:

5 - Cold Shutdown, 6 - Refueling Basis:

This IC addresses a hazardous event that causes damage to a SAFETY SYSTEM, or a structure containing SAFETY SYSTEM components, needed for the current operating mode.

This condition significantly reduces the margin to a loss or potential loss of a fission product barrier, and therefore represents an actual or potential substantial degradation of the level of safety of the plant.

EAL BASES DOCUMENT Rev. 0 Page 146 of 258

ATTACHMENT 1 EAL Bases The first conditional addresses damage to a SAFETY SYSTEM train that is in service/operation since indications for it will be readily available. The indications of degraded performance should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train.

The second conditional addresses damage to a SAFETY SYSTEM component that is not in service/operation or readily apparent through indications alone, or to a structure containing SAFETY SYSTEM components. Operators will make this determination based on the totality of available event and damage report information. This is intended to be a brief assessment not requiring lengthy analysis or quantification of the damage.

Escalation of the emergency classification level would be via IC CS1 or RS1.

Ground motion acceleration of 0.08g horizontal or 0.053g vertical is the Maximum Probable Earthquake as is considered generically as the Operating Basis Earthquake for DBNPS (ref. 8). Control room alarm indication of an earthquake greater than OBE is indicated on the seismic control panel (C5764A). RA-EP-02820 Earthquake provides the guidance for determining any required response actions if the OBE earthquake threshold is exceeded (ref. 1). The significance of seismic events is discussed under EAL HU2.1.

Internal FLOODING occurs from breaches of water systems that are located inside plant buildings and are connected to large water sources such as Intake Forebay or tanks (ref. 2).

External FLOODING may be due to high lake level. DBNPS flood emergency elevation is 578 ft. Site access would be limited to rail, boat or helicopter (ref. 3).

Seismic Category I structures are analyzed to withstand a sustained, design wind velocity of at least 90 mph. (ref. 4).

Areas containing functions and systems required for safe shutdown of the plant are identified by fire area in the fire abnormal procedure (ref. 5).

An EXPLOSION that degrades the performance of a SAFETY SYSTEM train or visibly damages a SAFETY SYSTEM component or structure would be classified under this EAL (ref. 6).

EAL BASES DOCUMENT Rev. 0 Page 147 of 258

ATTACHMENT 1 EAL Bases Basis Reference(s):

1. RA-EP-02820 Earthquake

2. RA-EP-02880 Internal Flooding

3. RA-EP-02830 Flooding

4. DBNPS UFSAR Section 3.3.1 Wind Criteria

5. DB-OP-02501 Serious Station Fire

6. RA-EP-02840 Explosion

7. NEI 99-01 CA6

8. Updated FSAR Section 3.1 Seismic Design EAL BASES DOCUMENT Rev. 0 Page 148 of 258

ATTACHMENT 1 EAL Bases Category S - System Malfunction EAL Group: Hot Conditions (RCS temperature > 200ºF); EALs in this category are applicable only in one or more hot operating modes.

Numerous system-related equipment failure events that warrant emergency classification have been identified in this category. They may pose actual or potential threats to plant safety.

The events of this category pertain to the following subcategories:

1. Loss of Essential 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 V essential buses.

2. Loss of Essential DC Power Loss of emergency electrical power can compromise plant SAFETY SYSTEM operability including decay heat removal and emergency core cooling systems, which may be necessary to ensure fission product barrier integrity. This category includes loss of essential plant 125 VDC power sources.

3. Loss of Control Room Indications Certain events that degrade plant operators ability to assess plant conditions within the plant warrant emergency classification. Losses of indicators are in this subcategory.

4. RCS Activity During normal operation, reactor coolant fission product activity is very low. Small concentrations of fission products in the coolant are primarily from the fission of tramp uranium in the fuel clad or minor perforations in the clad itself. Any significant increase from these base-line levels (2% - 5% clad failures) is indicative of fuel failures and is covered under the Fission Product Barrier Degradation category. However, lesser amounts of clad damage may result in coolant activity exceeding Technical Specification limits.

These fission products will be circulated with the reactor coolant and can be detected by coolant sampling.

EAL BASES DOCUMENT Rev. 0 Page 149 of 258

ATTACHMENT 1 EAL Bases

5. RCS Leakage The reactor vessel provides a volume for the coolant that covers the reactor core. The reactor pressure vessel and associated pressure piping (reactor coolant system) together provide a barrier to limit the release of radioactive material should the reactor fuel clad integrity fail. Excessive RCS leakage greater than Technical Specification limits indicates potential pipe cracks that may propagate to an extent threatening fuel clad, RCS and containment integrity.

6. RPS Failure This subcategory includes events related to failure of the Reactor Protection System (RPS) to initiate and complete reactor trips. In the plant licensing basis, postulated failures of the RPS to complete a reactor trip comprise a specific set of analyzed events referred to as Anticipated Transient Without Scram (ATWS) events. For EAL classification, however, ATWS is intended to mean ANY trip failure event that does not achieve reactor shutdown.

If RPS actuation fails to assure reactor shutdown, positive control of reactivity is at risk and could cause a threat to fuel clad, RCS and containment integrity.

7. Loss of Communications Certain events that degrade plant operators ability to communicate with essential personnel within or external to the plant warrant emergency classification.

8. Containment Failure Failure of containment isolation capability (under conditions in which the containment is not currently challenged) warrants emergency classification. Failure of containment pressure control capability also warrants emergency classification.

9. Hazardous Event Affecting SAFETY SYSTEMS Various natural and technological events that result in degraded plant SAFETY SYSTEM performance or significant visible damage warrant emergency classification under this subcategory.

EAL BASES DOCUMENT Rev. 0 Page 150 of 258

ATTACHMENT 1 EAL Bases Category: S - System Malfunction Subcategory: 1 - Loss of Essential AC Power Initiating Condition: Prolonged loss of ALL offsite and ALL onsite AC power to essential buses EAL:

SG1.1 General Emergency Loss of ALL offsite and ALL onsite AC power capability to essential 4160V buses C1 and D1 AND EITHER:

Restoration of at least one essential bus in < 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is not likely (Note 1)

Calculated Clad Temperature in Region 3 (DB-OP-02000 Figure 2)

Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Basis:

This IC addresses a prolonged loss of ALL power sources to AC essential buses. A loss of ALL AC power compromises the performance of ALL SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink. A prolonged loss of these buses will lead to a loss of one or more fission product barriers. In addition, fission product barrier monitoring capabilities may be degraded under these conditions.

The EAL should require declaration of a General Emergency prior to meeting the thresholds for IC FG1. This will allow additional time for implementation of offsite protective actions.

Escalation of the emergency classification from Site Area Emergency will occur if it is projected that power cannot be restored to at least one AC essential bus by the end of the analyzed station blackout coping period. Beyond this time, plant responses and event trajectory are subject to greater uncertainty, and there is an increased likelihood of challenges to multiple fission product barriers.

The estimate for restoring at least one essential bus should be based on a realistic appraisal of the situation. Mitigation actions with a low probability of success should not be used as a basis for delaying a classification upgrade. The goal is to maximize the time available to prepare for, and implement, protective actions for the public.

EAL BASES DOCUMENT Rev. 0 Page 151 of 258

ATTACHMENT 1 EAL Bases The EAL will also require a General Emergency declaration if the loss of AC power results in parameters that indicate an inability to adequately remove decay heat from the core.

This EAL is indicated by the extended loss of ALL offsite and onsite AC power capability to 4160V essential buses C1 and D1 either for greater than 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or that has resulted in indications of an actual loss of adequate core cooling.

Indication of continuing core cooling degradation is manifested by Calculated Clad Temperature in Region 3 (DB-OP-02000 Figure 2) (ref. 3).

The 4160 VAC System provides the power requirements for operation and safe shutdown of the plant. The essential switchgear are buses C1 and D1 (ref. 1).

The essential buses during plant operation are normally powered from the 13.8KV offsite power system through their respective 13.8KV/4160V bus tie transformers, via the Unit Auxiliary Transformer (X11). In non-power operating modes, the essential buses may be back-fed via the X11 and Main Transformer provided the main generator lead disconnect links are removed (ref. 1, 2).

A standby source of offsite power to each 4160V essential bus is provided from the 13.8KV offsite power system via two separate and independent 13.8KV/4160V Startup Transformers (X01 & X02). Normally each startup transformer serves as a reserve power source for only one essential bus. However, a single startup transformer can be aligned to power both essential buses (ref. 1, 2).

Each essential bus has a diesel generator (EDG1 & EDG2) to supply an onsite emergency source of power to safe shutdown loads in the event of a loss of the normal power source or loss of off-site power (ref. 1).

An Alternate AC power source, the Station Black Out (SBO) Diesel Generator, is located onsite. This onsite AC power source can be started from the Control Room and be loaded within 10 minutes of a SBO (ref. 1).

The SBODG fuel oil supply is separate from fuel oil supply for the station's EDGs. The SBODG fuel oil supply tank capacity is based on an eight-hour supply with the SBODG at rated load. A minimum supply of four hours must be stored in this tank to meet the site's station blackout duration analysis conditions (ref 1).

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 Calculated Clad Temperature in Region 3 (DB-OP-02000 Figure 2). Figure 2, Incore T/C Temperature vs. RCS Pressure for ICC, provides indication of how serious core conditions are based upon combinations of RCS pressure and incore thermocouple temperatures. If the RCS P-T point is in Region 3, the cladding temperatures in the high power regions of the Core may be 1400ºF or higher. This is a serious condition, which could lead to significant amounts of H2 production; Core damage may be unavoidable (ref. 3, 4).

EAL BASES DOCUMENT Rev. 0 Page 152 of 258

ATTACHMENT 1 EAL Bases Basis Reference(s):

1. UFSAR Section 8.0 Electric Power

2. DB-OP-02521 Loss of AC Bus Power Sources

3. DB-OP-02000 Figure 2, Incore T/C Temperature vs. RC Pressure for Inadequate Core Cooling

4. Bases and Deviation Document for DB-OP-02000

5. NEI 99-01 SG1 EAL BASES DOCUMENT Rev. 0 Page 153 of 258

ATTACHMENT 1 EAL Bases Category: S - System Malfunction Subcategory: 1 - Loss of Essential AC Power Initiating Condition: Loss of ALL essential AC and DC power sources for 15 minutes or longer EAL:

SG1.2 General Emergency Loss of ALL offsite and ALL onsite AC power capability to essential 4160V buses C1 and D1 for 15 min.

AND Loss of ALL 125 VDC power based on battery bus voltage indications < 105 VDC on ALL essential DC distribution panels D1P, D1N, D2P and D2N for 15 min.

(Note 1)

Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Basis:

This IC addresses a concurrent and prolonged loss of both essential AC AND DC power. A loss of ALL essential AC power compromises the performance of ALL SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink. A loss of essential DC power compromises the ability to monitor and control SAFETY SYSTEMS. A sustained loss of both AC and DC power will lead to multiple challenges to fission product barriers.

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

The 15-minute emergency declaration clock begins at the point when both EAL thresholds are met.

This EAL is indicated by the loss of ALL offsite and onsite essential AC power capability to 4160V essential buses C1 and D1 for greater than 15 minutes in combination with degraded DC power voltage. This EAL addresses operating experience from the March 2011 accident at Fukushima Daiichi.

EAL BASES DOCUMENT Rev. 0 Page 154 of 258

ATTACHMENT 1 EAL Bases The 4160 VAC System provides the power requirements for operation and safe shutdown of the plant. The essential switchgear are buses C1 and D1 (ref. 1).

The essential buses during plant operation are normally powered from the 13.8KV offsite power system through their respective 13.8KV/4160V bus tie transformers, via the Unit Auxiliary Transformer (X11). In non-power operating modes, the essential buses may be back-fed via the X11 and Main Transformer provided the main generator lead disconnect links are removed (ref. 1, 2). Credit for the X11 back-feed can only be taken if already aligned, as it takes greater than 15 minutes to align.

A standby source of offsite power to each 4160V essential bus is provided from the 13.8KV offsite power system via two separate and independent 13.8KV/4160V Startup Transformers (X01 & X02). Normally each startup transformer serves as a reserve power source for only one essential bus. However, a single startup transformer can be aligned to power both essential buses (ref. 1, 2).

Each essential bus has a diesel generator (EDG1 & EDG2) to supply an onsite emergency source of power to safe shutdown loads in the event of a loss of the normal power source or loss of off-site power (ref. 1).

An Alternate AC power source, the Station Black Out (SBO) Diesel Generator, is located onsite. This onsite AC power source can be started from the Control Room and be loaded within 10 minutes of a SBO (ref. 1).

The DC electrical distribution subsystem consists of two trains, designated Train 1 and Train 2.

Each train consists of a 250/125 VDC motor control center (MCC), and each 250/125 VDC MCC consists of two 125 VDC buses (one positive and one negative) that can be powered from a polarity specific battery or battery charger. The two DC buses then supply a specific Essential DC Distribution Panel (D1P and D1N for DCMCC 1, and D2P and D2N for DCMCC 2). The four Essential DC Distribution Panels independently supply DC electrical power to the four inverters, which in turn power the 120 VAC essential buses (ref. 1, 3, 4, 5, 6, 7).

The Class 1E DC loads have an operating voltage range of 105 to 135 volts. The minimum battery discharge voltage (requiring opening the degraded battery output breaker) is 105 VDC (ref. 1, 7).

Basis Reference(s):

1. UFSAR Section 8.0 Electric Power

2. DB-OP-02521 Loss of AC Bus Power Sources

3. DB-OP-02537 Loss of D1P and DAP

4. DB-OP-02538 Loss of D2P and DBP

5. DB-OP-02539 Loss of D1N and DAN

6. DB-OP-02540 Loss of D2N and DBN

7. System Description for 125/250 VDC and 120 V Instrumentation AC System

8. NEI 99-01 SG8 EAL BASES DOCUMENT Rev. 0 Page 155 of 258

ATTACHMENT 1 EAL Bases Category: S - System Malfunction Subcategory: 1 - Loss of Essential AC Power Initiating Condition: Loss of ALL offsite power and ALL onsite AC power to essential buses for 15 minutes or longer EAL:

SS1.1 Site Area Emergency Loss of ALL offsite and ALL onsite AC power capability to essential 4160V buses C1 and D1 for 15 min. (Note 1)

Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Basis:

This IC addresses a total loss of AC power that compromises the performance of ALL SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink.

In addition, fission product barrier monitoring capabilities may be degraded under these conditions. This IC represents a condition that involves actual or likely major failures of plant functions needed for the protection of the public.

Escalation of the emergency classification level would be via ICs RG1, FG1 or SG1.

This EAL is indicated by the loss of ALL offsite and onsite AC power capability (Table S-1) to 4160V essential buses C1 and D1. The essential switchgear are buses C1 (Train A) and D1 (Train B) (ref. 1). For emergency classification purposes, capability means that an AC power source is available to the essential buses, whether or not the buses are powered from it.

The 4160 VAC System provides the power requirements for operation and safe shutdown of the plant. The essential switchgear are buses C1 and D1 (ref. 1).

The essential buses during plant operation are normally powered from the 13.8KV offsite power system through their respective 13.8KV/4160V bus tie transformers, via the Unit Auxiliary Transformer (X11). In non-power operating modes, the essential buses may be back-fed via the X11 and Main Transformer provided the main generator lead disconnect links are removed (ref. 1, 2). Credit for the X11 back-feed can only be taken if already aligned, as it takes greater than 15 minutes to align.

EAL BASES DOCUMENT Rev. 0 Page 156 of 258

ATTACHMENT 1 EAL Bases A standby source of offsite power to each 4160V essential bus is provided from the 13.8KV offsite power system via two separate and independent 13.8KV/4160V Startup Transformers (X01 & X02). Normally each startup transformer serves as a reserve power source for only one essential bus. However, a single startup transformer can be aligned to power both essential buses (ref. 1, 2).

Each essential bus has a diesel generator (EDG1 & EDG2) to supply an onsite emergency source of power to safe shutdown loads in the event of a loss of the normal power source or loss of off-site power (ref. 1).

An Alternate AC power source, the Station Black Out (SBO) Diesel Generator, is located onsite. This onsite AC power source can be started from the Control Room and be loaded within 10 minutes of a SBO (ref. 1).

The 15-minute interval was selected as a threshold to exclude transient or momentary power losses. The interval begins when both offsite and onsite AC power capability are lost.

Basis Reference(s):

1. UFSAR Section 8.0 Electric Power

2. DB-OP-02521 Loss of AC Bus Power Sources

3. NEI 99-01 SS1 EAL BASES DOCUMENT Rev. 0 Page 157 of 258

ATTACHMENT 1 EAL Bases Category: S - System Malfunction Subcategory: 1 - Loss of Essential AC Power Initiating Condition: Loss of ALL but one AC power source to essential buses for 15 minutes or longer EAL:

SA1.1 Alert AC power capability, Table S-1, to essential 4160V buses C1 and D1 reduced to a single power source for 15 min. (Note 1)

AND ANY additional single power source failure will result in loss of ALL AC power to SAFETY SYSTEMS Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Table S-1 Offsite/Onsite AC Power Sources Offsite:

X11 X11 (back-fed via Main Transformer if already aligned)

X01 X02 Onsite:

EDG1 EDG2 SBODG Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4- Hot Shutdown Basis:

This IC describes a significant degradation of offsite and onsite AC power sources such that ANY additional single failure would result in a loss of ALL AC power to SAFETY SYSTEMS. In this condition, the sole AC power source may be powering one, or more than one, train of safety-related equipment. This IC provides an escalation path from IC SU1.

EAL BASES DOCUMENT Rev. 0 Page 158 of 258

ATTACHMENT 1 EAL Bases An AC power source is a source recognized in AOPs and EOPs, and capable of supplying required power to an essential bus. Some examples of this condition are presented below.

A loss of ALL offsite power with a concurrent failure of ALL but one emergency power source (e.g., an onsite diesel generator)

A loss of ALL offsite power and loss of ALL emergency power sources (e.g., onsite diesel generators) with a single train of essential buses being back-fed from the unit main generator A loss of emergency power sources (e.g., onsite diesel generators) with a single train of essential buses being fed from an offsite power source Escalation of the emergency classification level would be via IC SS1.

For emergency classification purposes, capability means that an AC power source is available to the essential buses, whether or not the buses are powered from it.

The 4160 VAC System provides the power requirements for operation and safe shutdown of the plant. The essential switchgear are buses C1 and D1 (ref. 1).

The essential buses during plant operation are normally powered from the 13.8KV offsite power system through their respective 13.8KV/4160V bus tie transformers, via the Unit Auxiliary Transformer (X11). In non-power operating modes, the essential buses may be back-fed via the X11 and Main Transformer provided the main generator lead disconnect links are removed (ref. 1, 2). Credit for the X11 back-feed can only be taken if already aligned, as it takes greater than 15 minutes to align.

A standby source of offsite power to each 4160V essential bus is provided from the 13.8KV offsite power system via two separate and independent 13.8KV/4160V Startup Transformers (X01 & X02). Normally each startup transformer serves as a reserve power source for only one essential bus. However, a single startup transformer can be aligned to power both essential buses (ref. 1, 2).

Each essential bus has a diesel generator (EDG1 & EDG2) to supply an onsite emergency source of power to safe shutdown loads in the event of a loss of the normal power source or loss of off-site power (ref. 1).

An Alternate AC power source, the Station Black Out (SBO) Diesel Generator, is located onsite. This onsite AC power source can be started from the Control Room and be loaded within 10 minutes of a SBO (ref. 1).

The 15-minute interval was selected as a threshold to exclude transient or momentary power losses. If the capability of a second source of essential bus power is not restored within 15 minutes, an Alert is declared under this EAL.

EAL BASES DOCUMENT Rev. 0 Page 159 of 258

ATTACHMENT 1 EAL Bases Basis Reference(s):

1. UFSAR Section 8.0 Electric Power

2. DB-OP-02521 Loss of AC Bus Power Sources

3. NEI 99-01 SA1 EAL BASES DOCUMENT Rev. 0 Page 160 of 258

ATTACHMENT 1 EAL Bases Category: S - System Malfunction Subcategory: 1 - Loss of Essential AC Power Initiating Condition: Loss of ALL offsite AC power capability to essential buses for 15 minutes or longer EAL:

SU1.1 Unusual Event Loss of ALL offsite AC power capability, Table S-1, to essential 4160V buses C1 and D1 for 15 min. (Note 1)

Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Table S-1 Offsite/Onsite AC Power Sources Offsite:

X11 X11 (back-fed via Main Transformer if already aligned)

X01 X02 Onsite:

EDG1 EDG2 SBODG Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Basis:

This IC addresses a prolonged loss of offsite power. The loss of offsite power sources renders the plant more vulnerable to a complete loss of power to AC essential buses. This condition represents a potential reduction in the level of safety of the plant.

For emergency classification purposes, capability means that an offsite AC power source(s) is available to the essential buses, whether or not the buses are powered from it.

EAL BASES DOCUMENT Rev. 0 Page 161 of 258

ATTACHMENT 1 EAL Bases Fifteen minutes was selected as a threshold to exclude transient or momentary losses of offsite power.

Escalation of the emergency classification level would be via IC SA1.

For emergency classification purposes, capability means that an AC power source is available to the essential buses, whether or not the buses are powered from it.

The 4160 VAC System provides the power requirements for operation and safe shutdown of the plant. The essential switchgear are buses C1 and D1 (ref. 1).

The essential buses during plant operation are normally powered from the 13.8KV offsite power system through their respective 13.8KV/4160V bus tie transformers, via the Unit Auxiliary Transformer (X11). In non-power operating modes, the essential buses may be back-fed via the X11 and Main Transformer provided the main generator lead disconnect links are removed (ref. 1, 2). Credit for the X11 back-feed can only be taken if already aligned, as it takes greater than 15 minutes to align.

A standby source of offsite power to each 4160V essential bus is provided from the 13.8KV offsite power system via two separate and independent 13.8KV/4160V Startup Transformers (X01 & X02). Normally each startup transformer serves as a reserve power source for only one essential bus. However, a single startup transformer can be aligned to power both essential buses (ref. 1, 2).

Each essential bus has a diesel generator (EDG1 & EDG2) to supply an onsite emergency source of power to safe shutdown loads in the event of a loss of the normal power source or loss of off-site power (ref. 1).

An Alternate AC power source, the Station Black Out (SBO) Diesel Generator, is located onsite. This onsite AC power source can be started from the Control Room and be loaded within 10 minutes of a SBO (ref. 1).

Basis Reference(s):

1. UFSAR Section 8.0 Electric Power

2. DB-OP-02521 Loss of AC Bus Power Sources

3. NEI 99-01 SU1 EAL BASES DOCUMENT Rev. 0 Page 162 of 258

ATTACHMENT 1 EAL Bases Category: S - System Malfunction Subcategory: 2 - Loss of Essential DC Power Initiating Condition: Loss of ALL essential DC power for 15 minutes or longer EAL:

SS2.1 Site Area Emergency Loss of ALL 125 VDC power based on battery bus voltage indications < 105 VDC on ALL essential DC distribution panels D1P, D1N, D2P and D2N for 15 min. (Note 1)

Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Basis:

This IC addresses a loss of essential DC power, which compromises the ability to monitor and control SAFETY SYSTEMS. In modes above Cold Shutdown, this condition involves a major failure of plant functions needed for the protection of the public.

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

Escalation of the emergency classification level would be via ICs RG1, FG1 or SG1.

The DC electrical distribution subsystem consists of two trains, designated Train 1 and Train 2.

Each train consists of a 250/125 VDC motor control center (MCC), and each 250/125 VDC MCC consists of two 125 VDC buses (one positive and one negative) that can be powered from a polarity specific battery or battery charger. The two DC buses then supply a specific Essential DC Distribution Panel (D1P and D1N for DCMCC 1, and D2P and D2N for DCMCC 2). The four Essential DC Distribution Panels independently supply DC electrical power to the four inverters, which in turn power the 120 VAC essential buses (ref. 1, 2, 3, 4, 5, 6).

The Class 1E DC loads have an operating voltage range of 105 to 135 volts. The minimum battery discharge voltage (requiring opening the degraded battery output breaker) is 105 VDC (ref. 1, 6).

EAL BASES DOCUMENT Rev. 0 Page 163 of 258

ATTACHMENT 1 EAL Bases Basis Reference(s):

1. DBNPS UFSAR Section 8.0 Electrical Power

2. DB-OP-02537 Loss of D1P and DAP

3. DB-OP-02538 Loss of D2P and DBP

4. DB-OP-02539 Loss of D1N and DAN

5. DB-OP-02540 Loss of D2N and DBN

6. System Description for 125/250 VDC and 120 V Instrumentation AC System

7. NEI 99-01 SS8 EAL BASES DOCUMENT Rev. 0 Page 164 of 258

ATTACHMENT 1 EAL Bases Category: S - System Malfunction Subcategory: 3 - Loss of Control Room Indications Initiating Condition: UNPLANNED loss of Control Room indications for 15 minutes or longer with a significant transient in progress EAL:

SA3.1 Alert An UNPLANNED event results in the inability to monitor one or more Table S-2 SAFETY SYSTEM parameters from within the Control Room for 15 min. (Note 1)

AND ANY Significant Transient is in progress, Table S-3 Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Table S-2 Safety System Parameters Reactor power Pressurizer level RCS pressure In-core T/C temperature Level in at least one S/G Auxiliary or emergency feed flow Table S-3 Significant Transients Reactor trip Runback > 25% thermal power Electrical load rejection > 25%

electrical load Safety injection actuation Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown EAL BASES DOCUMENT Rev. 0 Page 165 of 258

ATTACHMENT 1 EAL Bases Basis:

This IC addresses the difficulty associated with monitoring rapidly changing plant conditions during a transient without the ability to obtain SAFETY SYSTEM parameters from within the Control Room. During this condition, the margin to a potential fission product barrier challenge is reduced. It thus represents a potential substantial degradation in the level of safety of the plant.

As used in this EAL, an inability to monitor means that values for one or more of the listed parameters cannot be determined from within the Control Room. This situation would require a loss of ALL of the Control Room sources for the given parameter(s). For example, the reactor power level cannot be determined from any analog, digital and recorder source within the Control Room.

An event involving a loss of plant indications, annunciators and/or display systems is evaluated in accordance with 10 CFR 50.72 (and associated guidance in NUREG-1022) to determine if an NRC event report is required. The event would be reported if it significantly impaired the capability to perform emergency assessments. In particular, emergency assessments necessary to implement abnormal operating procedures, emergency operating procedures, and emergency plan implementing procedures addressing emergency classification, accident assessment, or protective action decision-making.

This EAL is focused on a selected subset of plant parameters associated with the key safety functions of reactivity control, core cooling and RCS heat removal. The loss of the ability to determine one or more of these parameters from within the Control Room is considered to be more significant than simply a reportable condition. In addition, if ALL indication sources for one or more of the listed parameters are lost, then the ability to determine the values of other SAFETY SYSTEM parameters may be impacted as well. For example, if the value for RCS pressure cannot be determined from the indications and recorders on a main control board, the SPDS or the plant computer, the availability of other parameter values may be compromised as well.

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

Escalation of the emergency classification level would be via ICs FS1 or RS1 SAFETY SYSTEM parameters listed in Table S-2 are monitored in the Control Room through a combination of hard control panel indicators as well as computer based information systems.

The Plant Process Computer, which displays SPDS required information, serves as a redundant compensatory indicator, which may be utilized in lieu of normal Control Room indicators (ref. 1, 2, 3).

EAL BASES DOCUMENT Rev. 0 Page 166 of 258

ATTACHMENT 1 EAL Bases Significant transients are listed in Table S-3 and include response to automatic or manually initiated functions such as reactor trips, runbacks involving greater than 25% thermal power change, electrical load rejections of greater than 25% full electrical load or SI injection actuations.

Basis Reference(s):

1. UFSAR Section 7.5 Safety-Related Display Instrumentation

2. DB-OP-02541 Loss of YAU

3. DB-OP-02542 Loss of YBU

4. NEI 99-01 SA2 EAL BASES DOCUMENT Rev. 0 Page 167 of 258

ATTACHMENT 1 EAL Bases Category: S - System Malfunction Subcategory: 3 - Loss of Control Room Indications Initiating Condition: UNPLANNED loss of Control Room indications for 15 minutes or longer EAL:

SU3.1 Unusual Event An UNPLANNED event results in the inability to monitor one or more Table S-2 SAFETY SYSTEM parameters from within the Control Room for 15 min. (Note 1)

Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Table S-2 Safety System Parameters Reactor power Pressurizer level RCS pressure In-core T/C temperature Level in at least one S/G Auxiliary or emergency feed flow Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Basis:

This IC addresses the difficulty associated with monitoring normal plant conditions without the ability to obtain SAFETY SYSTEM parameters from within the Control Room. This condition is a precursor to a more significant event and represents a potential degradation in the level of safety of the plant.

As used in this EAL, an inability to monitor means that values for one or more of the listed parameters cannot be determined from within the Control Room. This situation would require a loss of ALL of the Control Room sources for the given parameter(s). For example, the reactor power level cannot be determined from any analog, digital and recorder source within the Control Room.

An event involving a loss of plant indications, annunciators and/or display systems is evaluated in accordance with 10 CFR 50.72 (and associated guidance in NUREG-1022) to determine if an NRC event report is required. The event would be reported if it significantly impaired the capability to perform emergency assessments. In particular, emergency assessments necessary to implement abnormal operating procedures, emergency operating procedures, EAL BASES DOCUMENT Rev. 0 Page 168 of 258

ATTACHMENT 1 EAL Bases and emergency plan implementing procedures addressing emergency classification, accident assessment, or protective action decision-making.

This EAL is focused on a selected subset of plant parameters associated with the key safety functions of reactivity control, core cooling and RCS heat removal. The loss of the ability to determine one or more of these parameters from within the Control Room is considered to be more significant than simply a reportable condition. In addition, if ALL indication sources for one or more of the listed parameters are lost, then the ability to determine the values of other SAFETY SYSTEM parameters may be impacted as well. For example, if the value for RCS pressure cannot be determined from the indications and recorders on a main control board, the SPDS or the plant computer, the availability of other parameter values may be compromised as well.

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

Escalation of the emergency classification level would be via IC SA3.

SAFETY SYSTEM parameters listed in Table S-2 are monitored in the Control Room through a combination of hard control panel indicators as well as computer based information systems.

The Plant Process Computer, which displays SPDS required information, serves as a redundant compensatory indicator, which may be utilized in lieu of normal Control Room indicators (ref. 1, 2, 3).

Basis Reference(s):

1. UFSAR Section 7.5 Safety-Related Display Instrumentation

2. DB-OP-02541 Loss of YAU

3. DB-OP-02542 Loss of YBU

4. NEI 99-01 SU2 EAL BASES DOCUMENT Rev. 0 Page 169 of 258

ATTACHMENT 1 EAL Bases Category: S - System Malfunction Subcategory: 4 - RCS Activity Initiating Condition: Reactor coolant activity greater than Technical Specification allowable limits EAL:

SU4.1 Unusual Event RCS activity > Technical Specification LCO 3.4.16 as indicated by ANY of the following:

Dose equivalent I-131 in the unacceptable region of Figure 3.4.16-1

> 1 µCi/gm dose equivalent I-131 for > 48 hrs

> 100/ µCi/gm gross specific coolant activity Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Basis:

This IC addresses a reactor coolant activity value that exceeds an allowable limit specified in Technical Specifications. This condition is a precursor to a more significant event and represents a potential degradation of the level of safety of the plant.

Escalation of the emergency classification level would be via ICs FA1 or the Recognition Category R ICs.

Technical Specification LCO 3.4.16 Condition A limits RCS System Dose Equivalent I-131 to 1.0 µCi/gm for > 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. Technical Specification Section 3.4.16 Condition B limits RCS System Dose Equivalent I-131 to within Figure 3.4.16-1 and gross specific activity to

< 100/ µCi/gm. (ref 1).

Basis Reference(s):

1. DBNPS Technical Specifications section 3.4.16 RCS Specific Activity

2. NEI 99-01 SU3 EAL BASES DOCUMENT Rev. 0 Page 170 of 258

ATTACHMENT 1 EAL Bases Category: S - System Malfunction Subcategory: 5 - RCS Leakage Initiating Condition: RCS leakage for 15 minutes or longer EAL:

SU5.1 Unusual Event RCS unidentified or pressure boundary leakage > 10 gpm for 15 min.

OR RCS identified leakage > 25 gpm for 15 min.

OR Leakage from the RCS to a location outside containment > 25 gpm for 15 min.

(Note 1)

Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Basis:

This IC addresses RCS leakage, which may be a precursor to a more significant event. In this case, RCS leakage has been detected and operators, following applicable procedures, have been unable to promptly isolate the leak. This condition is considered to be a potential degradation of the level of safety of the plant.

The first and second EAL conditions are focused on a loss of mass from the RCS due to unidentified leakage," "pressure boundary leakage" or "identified leakage (as these leakage types are defined in the plant Technical Specifications). The third condition addresses an RCS mass loss caused by an UNISOLABLE leak through an interfacing system. These conditions thus apply to leakage into the containment, a secondary-side system (e.g., steam generator tube leakage) or a location outside of containment.

The leak rate value was selected because it is usually observable with normal Control Room indications. Lesser values typically require time-consuming calculations to determine (e.g., a mass balance calculation). The first condition uses a lower value that reflects the greater significance of unidentified or pressure boundary leakage.

The release of mass from the RCS due to the as-designed/expected operation of a relief valve does not warrant an emergency classification. An emergency classification would be required if a mass loss is caused by a relief valve that is not functioning as designed/expected (e.g., a relief valve sticks open and the line flow cannot be isolated).

EAL BASES DOCUMENT Rev. 0 Page 171 of 258

ATTACHMENT 1 EAL Bases The 15-minute threshold duration allows sufficient time for prompt operator actions to isolate the leakage, if possible.

Unidentified leakage is ALL leakage (except RCP seal return) that is not identified leakage (ref. 1, 2).

Escalation of this EAL to the Alert level is via Category F, Fission Product Barrier Degradation, EAL FA1.1.

Basis Reference(s):

1. DBNPS Technical Specifications Definitions section 1.1

2. UFSAR Section 5.2.4.7 Leakage Identification

3. DB-OP-02522 Small RCS Leaks

4. NEI 99-01 SU4 EAL BASES DOCUMENT Rev. 0 Page 172 of 258

ATTACHMENT 1 EAL Bases Category: S - System Malfunction Subcategory: 6 - RPS Failure Initiating Condition: Inability to shut down the reactor causing a challenge to core cooling or RCS heat removal EAL:

SS6.1 Site Area Emergency An automatic or manual trip fails to shut down the reactor as indicated by reactor power 5%

AND ALL actions to shut down the reactor are not successful as indicated by reactor power 5%

AND EITHER:

Calculated Clad Temperature in Region 3 (DB-OP-02000 Figure 2)

MFW, AFW and MU-HPI PORV Cooling are all unavailable Mode Applicability:

1 - Power Operation Basis:

This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor trip that results in a reactor shutdown, ALL subsequent operator actions to manually shutdown the reactor are unsuccessful, and continued power generation is challenging the capability to adequately remove heat from the core and/or the RCS. This condition will lead to fuel damage if additional mitigation actions are unsuccessful and thus warrants the declaration of a Site Area Emergency.

In some instances, the emergency classification resulting from this IC/EAL may be higher than that resulting from an assessment of the plant responses and symptoms against the Recognition Category F ICs/EALs. This is appropriate in that the Recognition Category F ICs/EALs do not address the additional threat posed by a failure to shut down the reactor. The inclusion of this IC and EAL ensures the timely declaration of a Site Area Emergency in response to prolonged failure to shutdown the reactor.

A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.

Escalation of the emergency classification level would be via ICs RG1 or FG1.

EAL BASES DOCUMENT Rev. 0 Page 173 of 258

ATTACHMENT 1 EAL Bases This EAL addresses the following:

ANY automatic reactor trip signal followed by a manual trip that fails to shut down the reactor to an extent the reactor is producing energy in excess of the heat load for which the SAFETY SYSTEMS were designed (EAL SA6.1), AND Indications that either core cooling is extremely challenged or heat removal is extremely challenged The combination of failure of both front line and backup protection systems to function in response to a plant transient, along with the continued production of heat, poses a direct threat to the Fuel Clad and RCS barriers.

Reactor shutdown achieved by use of other trip actions specified in DB-OP-02000 RPS, SFAS, SFRCS Trip, or SG Tubing Rupture (locally opening Reactor Trip Breaker, emergency boration or manually driving control rods) are also credited as a successful manual trip methods provided reactor power can be reduced below 5% before indications of an extreme challenge to either core cooling or heat removal exist (ref. 1).

Each of the redundant Auxiliary Feed Pumps is sized to provide 100% of the capacity required by the SGs to remove 5% of the reactor thermal power produced at full load steam pressure conditions. The 5% power portion of this EAL threshold was chosen to be an easily recognizable, onscale indication that the reactor trip has not functioned to shutdown the reactor assuming the reactor power was in the normal operating range.

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 Calculated Clad Temperature in Region 3 (DB-OP-02000 Figure 2). Figure 2, Incore T/C Temperature vs. RCS Pressure for ICC, provides indication of how serious core conditions are based upon combinations of RCS pressure and incore thermocouple temperatures. If the RCS P-T point is in Region 3, the cladding temperatures in the high power regions of the core may be 1400°F or higher. This is a serious condition, which will lead to significant amounts of H2 production; core damage may be unavoidable (ref. 1).

An extreme challenge to heat removal is defined as a complete loss of MFW, AFW and MU-HPI PORV Cooling (ref. 2, 3).

Basis Reference(s):

1. DBNPS Technical Specifications section 3.3.1 Reactor Protection System (RPS)

Instrumentation

2. DB-OP-02000 RPS, SFAS, SFRCS Trip, or SG Tubing Rupture

3. Bases and Deviation Document for DB-OP-02000

4. NEI 99-01 SS5 EAL BASES DOCUMENT Rev. 0 Page 174 of 258

ATTACHMENT 1 EAL Bases Category: S - System Malfunction Subcategory: 6 - RPS Failure Initiating Condition: Automatic or manual trip fails to shut down the reactor AND subsequent manual actions taken at the Controls Area are not successful in shutting down the reactor EAL:

SA6.1 Alert An automatic or manual trip fails to shut down the reactor as indicated by reactor power 5%

AND Manual trip actions taken at the Controls Area (manual RPS trip pushbuttons and de-energizing E2 and F2) are not successful in shutting down the reactor as indicated by reactor power 5% (Note 8)

Note 8: A manual trip action is ANY operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron injection strategies.

Mode Applicability:

1 - Power Operation Basis:

This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor trip that results in a reactor shutdown, AND subsequent operator manual actions taken at the Controls Area to shutdown the reactor are also unsuccessful. This condition represents an actual or potential substantial degradation of the level of safety of the plant. An emergency declaration is required even if the reactor is subsequently shutdown by an action taken away from the Controls Area since this event entails a significant failure of the RPS.

A manual action at the Controls Area is ANY operator action, or set of actions, which causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor trip). This action does not include manually driving in control rods or implementation of boron injection strategies. If this action(s) is unsuccessful, operators would immediately pursue additional manual actions at locations away from the Controls Area (e.g., locally opening breakers).

Actions taken at back-panels or other locations within the Control Room, or any location outside the Control Room, are not considered to be at the Controls Area.

The plant response to the failure of an automatic or manual reactor trip will vary based upon several factors including the reactor power level prior to the event, availability of the condenser, performance of mitigation equipment and actions, other concurrent plant conditions, etc. If the failure to shut down the reactor is prolonged enough to cause a EAL BASES DOCUMENT Rev. 0 Page 175 of 258

ATTACHMENT 1 EAL Bases challenge to the core cooling or RCS heat removal safety functions, the emergency classification level will escalate to a Site Area Emergency via IC SS6. Depending upon plant responses and symptoms, escalation is also possible via IC FS1. Absent the plant conditions needed to meet either IC SS6 or FS1, an Alert declaration is appropriate for this event.

It is recognized that plant responses or symptoms may also require an Alert declaration in accordance with the Recognition Category F ICs; however, this IC and EAL are included to ensure a timely emergency declaration.

A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.

This EAL addresses ANY automatic or manual reactor trip signal that fails to shut down the reactor followed by a subsequent manual trip that fails to shut down the reactor to an extent the reactor is producing energy in excess of the heat load for which the SAFETY SYSTEMS were designed.

For the purposes of emergency classification, successful manual trip actions are those, which can be quickly performed from the Controls Area (ATCA) (i.e., manual pushbuttons, de-energizing E2 and F2, and reactor trip test key). Reactor shutdown achieved by use of other trip actions specified in DB-OP-02000 RPS, SFAS, SFRCS Trip, or SG Tubing Rupture (locally opening Reactor Trip Breaker, emergency boration or manually driving control rods) do not constitute a successful manual trip (ref. 1).

Each of the redundant Auxiliary Feed Pumps is sized to provide 100% of the capacity required by the SGs to remove 5% of the reactor thermal power produced at full load steam pressure conditions. The 5% power portion of this EAL threshold was chosen to be an easily recognizable, onscale indication that the reactor trip has not functioned to shutdown the reactor assuming the reactor power was in the normal operating range.

Escalation of this event to a Site Area Emergency would be under EAL SS6.1 or Emergency Director judgment.

Basis Reference(s):

1. DB-OP-02000 RPS, SFAS, SFRCS Trip, or SG Tubing Rupture

2. NEI 99-01 SA5 EAL BASES DOCUMENT Rev. 0 Page 176 of 258

ATTACHMENT 1 EAL Bases Category: S - System Malfunction Subcategory: 6 - RPS Failure Initiating Condition: Automatic or manual trip fails to shut down the reactor EAL:

SU6.1 Unusual Event An automatic trip did not shut down the reactor as indicated by reactor power 5% after ANY RPS setpoint is exceeded AND A subsequent manual trip action taken at the Controls Area (manual RPS trip pushbuttons or de-energizing E2 and F2) is successful in shutting down the reactor as indicated by reactor power < 5% (Note 8)

Note 8: A manual trip action is ANY operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron injection strategies.

Mode Applicability:

1 - Power Operation Basis:

This EAL addresses a failure of the RPS to initiate or complete an automatic or manual reactor trip that results in a reactor shutdown, AND a subsequent operator manual action taken at the Controls Area is successful in shutting down the reactor. This event is a precursor to a more significant condition and thus represents a potential degradation of the level of safety of the plant.

Following the failure on an automatic reactor trip, operators will promptly initiate manual actions at the Controls Area to shutdown the reactor (e.g., initiate a manual reactor trip). If these manual actions are successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plants decay heat removal systems.

If an initial manual reactor trip is unsuccessful, operators will promptly take manual action at another location(s) on the Controls Area to shutdown the reactor (e.g., initiate a manual reactor trip using a different switch, reactor trip test key). Depending upon several factors, the initial or subsequent effort to manually trip the reactor, or a concurrent plant condition, may lead to the generation of an automatic reactor trip signal. If a subsequent manual or automatic trip is successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plants decay heat removal systems.

EAL BASES DOCUMENT Rev. 0 Page 177 of 258

ATTACHMENT 1 EAL Bases A manual action at the Controls Area is ANY operator action, or set of actions, which causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor trip). This action does not include manually driving in control rods or implementation of boron injection strategies. Actions taken at back-panels or other locations within the Control Room, or any location outside the Control Room, are not considered to be at the Controls Area.

The plant response to the failure of an automatic or manual reactor trip will vary based upon several factors including the reactor power level prior to the event, availability of the condenser, performance of mitigation equipment and actions, other concurrent plant conditions, etc. If subsequent operator manual actions taken at the Controls Area are also unsuccessful in shutting down the reactor, then the emergency classification level will escalate to an Alert via IC SA6. Depending upon the plant response, escalation is also possible via IC FA1. Absent the plant conditions needed to meet either IC SA6 or FA1, an Unusual Event declaration is appropriate for this event.

A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.

Should a reactor trip signal be generated as a result of plant work (e.g., RPS setpoint testing),

the following classification guidance should be applied.

If the signal causes a plant transient that should have included an automatic reactor trip and the RPS fails to automatically shutdown the reactor, then this IC and the EALs are applicable, and should be evaluated.

If the signal does not cause a plant transient and the trip failure is determined through other means (e.g., assessment of test results), then this IC and the EALs are not applicable and no classification is warranted.

The first condition of this EAL identifies the need to cease critical reactor operations by actuation of the automatic Reactor Protection System (RPS) trip function. A reactor trip is automatically initiated by the RPS when certain continuously monitored parameters exceed predetermined setpoints (ref. 1).

Following a successful reactor trip, rapid insertion of the control rods occurs. Reactor power promptly drops to a fraction of the original power level and then decays to a level several decades less with a negative startup rate. The reactor power drop continues until reactor power reaches the point at which the influence of source neutrons on reactor power starts to be observable. A predictable post-trip response from an automatic reactor trip signal should therefore consist of a prompt drop in reactor power as sensed by the nuclear instrumentation and a lowering of power into the source range. A successful trip has therefore occurred when there is sufficient rod insertion from the trip of RPS to bring the reactor power below the immediate shutdown decay heat level of 5% (ref. 2).

EAL BASES DOCUMENT Rev. 0 Page 178 of 258

ATTACHMENT 1 EAL Bases Each of the redundant Auxiliary Feed Pumps is sized to provide 100% of the capacity required by the SGs to remove 5% of the reactor thermal power produced at full load steam pressure conditions. The 5% power portion of this EAL threshold was chosen to be an easily recognizable, onscale indication that the reactor trip has not functioned to shutdown the reactor assuming the reactor power was in the normal operating range.

For the purposes of emergency classification, successful manual trip actions are those, which can be quickly performed from the Controls Area (ATCA) (i.e., manual pushbuttons or de-energizing E2 and F2). Reactor shutdown achieved by use of other trip actions specified in DB-OP-02000 RPS, SFAS, SFRCS Trip, or SG Tubing Rupture (locally opening Reactor Trip Breaker, emergency boration or manually driving control rods) do not constitute a successful manual trip (ref. 2).

In the event that the operator identifies a reactor trip is imminent and initiates a successful manual reactor trip before the automatic RPS trip setpoint is reached, no declaration is required. The successful manual trip of the reactor before it reaches its automatic trip setpoint or reactor trip signals caused by instrumentation channel failures do not lead to a potential fission product barrier loss. However, if subsequent manual reactor trip actions fail to reduce reactor power below 5%, the event escalates to the Alert under EAL SA6.1.

If by procedure, operator actions include the initiation of an immediate manual trip following receipt of an automatic trip signal and there are no clear indications that the automatic trip failed (such as a time delay following indications that a trip setpoint was exceeded), it may be difficult to determine if the reactor was shut down because of automatic trip or manual actions.

If a subsequent review of the trip actuation indications reveals that, the automatic trip did not cause the reactor to be shut down, then consideration should be given to evaluating the fuel for potential damage, and the reporting requirements of 10 CFR 50.72 should be considered for the transient event.

Basis Reference(s):

1. DBNPS Technical Specifications section 3.3.1 Reactor Protection System (RPS)

Instrumentation

2. DB-OP-02000 RPS, SFAS, SFRCS Trip, or SG Tubing Rupture

3. NEI 99-01 SU5 EAL BASES DOCUMENT Rev. 0 Page 179 of 258

ATTACHMENT 1 EAL Bases Category: S - System Malfunction Subcategory: 6 - RPS Failure Initiating Condition: Automatic or manual trip fails to shut down the reactor EAL:

SU6.2 Unusual Event A manual trip did not shut down the reactor as indicated by reactor power 5% after ANY manual trip action was initiated AND A subsequent automatic trip or manual trip action taken at the Controls Area (manual RPS trip pushbuttons or de-energizing E2 and F2) is successful in shutting down the reactor as indicated by reactor power < 5% (Note 8)

Note 8: A manual trip action is ANY operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron injection strategies.

Mode Applicability:

1 - Power Operation Basis:

This EAL addresses a failure of the RPS to initiate or complete a manual reactor trip that results in a reactor shutdown, AND either a subsequent operator manual action taken at the Controls Area or an automatic trip is successful in shutting down the reactor. This event is a precursor to a more significant condition and thus represents a potential degradation of the level of safety of the plant.

Following the failure on an automatic reactor trip, operators will promptly initiate manual actions at the Controls Area to shutdown the reactor (e.g., initiate a manual reactor trip). If these manual actions are successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plants decay heat removal systems.

If an initial manual reactor trip is unsuccessful, operators will promptly take manual action at another location(s) on the Controls Area to shutdown the reactor (e.g., initiate a manual reactor trip) using a different switch. Depending upon several factors, the initial or subsequent effort to manually trip the reactor, or a concurrent plant condition, may lead to the generation of an automatic reactor trip signal. If a subsequent manual or automatic trip is successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plants decay heat removal systems.

EAL BASES DOCUMENT Rev. 0 Page 180 of 258

ATTACHMENT 1 EAL Bases A manual action at the Controls Area is ANY operator action, or set of actions, which causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor trip). This action does not include manually driving in control rods or implementation of boron injection strategies. Actions taken at back-panels or other locations within the Control Room, or any location outside the Control Room, are not considered to be at the Controls Area.

The plant response to the failure of an automatic or manual reactor trip will vary based upon several factors including the reactor power level prior to the event, availability of the condenser, performance of mitigation equipment and actions, other concurrent plant conditions, etc. If subsequent operator manual actions taken at the Controls Area are also unsuccessful in shutting down the reactor, then the emergency classification level will escalate to an Alert via IC SA6. Depending upon the plant response, escalation is also possible via IC FA1. Absent the plant conditions needed to meet either IC SA6 or FA1, an Unusual Event declaration is appropriate for this event.

A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.

Should a reactor trip signal be generated as a result of plant work (e.g., RPS setpoint testing),

the following classification guidance should be applied.

If the signal causes a plant transient that should have included an automatic reactor trip and the RPS fails to automatically shutdown the reactor, then this IC and the EALs are applicable, and should be evaluated.

If the signal does not cause a plant transient and the trip failure is determined through other means (e.g., assessment of test results), then this IC and the EALs are not applicable and no classification is warranted.

This EAL addresses a failure of a manually initiated trip in the absence of having exceeded an automatic RPS trip setpoint AND a subsequent automatic or manual trip is successful in shutting down the reactor (reactor power < 5%) (ref. 1).

Following a successful reactor trip, rapid insertion of the control rods occurs. Reactor power promptly drops to a fraction of the original power level and then decays to a level several decades less with a negative startup rate. The reactor power drop continues until reactor power reaches the point at which the influence of source neutrons on reactor power starts to be observable. A predictable post-trip response from a manual reactor trip signal should therefore consist of a prompt drop in reactor power as sensed by the nuclear instrumentation and a lowering of power into the source range. A successful trip has therefore occurred when there is sufficient rod insertion from the trip of RPS to bring the reactor power below the immediate shutdown decay heat level of 5% (ref. 2).

Each of the redundant Auxiliary Feed Pumps is sized to provide 100% of the capacity required by the SGs to remove 5% of the reactor thermal power produced at full load steam pressure conditions. The 5% power portion of this EAL threshold was chosen to be an easily recognizable, onscale indication that the reactor trip has not functioned to shutdown the reactor assuming the reactor power was in the normal operating range.

EAL BASES DOCUMENT Rev. 0 Page 181 of 258

ATTACHMENT 1 EAL Bases For the purposes of emergency classification, successful manual trip actions are those, which can be quickly performed from the Controls Area (ATCA) (i.e., manual pushbuttons or de-energizing E2 and F2). Reactor shutdown achieved by use of other trip actions specified in DB-OP-02000 RPS, SFAS, SFRCS Trip, or SG Tubing Rupture (locally opening Reactor Trip Breaker, emergency boration or manually driving control rods) do not constitute a successful manual trip (ref. 2).

If both subsequent automatic and subsequent manual reactor trip actions in the Control Room fail to reduce reactor power below the power associated with the SAFETY SYSTEM design

(< 5%) following a failure of an initial manual trip, the event escalates to an Alert under EAL SA6.1 Basis Reference(s):

1. DBNPS Technical Specifications section 3.3.1 Reactor Protection System (RPS)

Instrumentation

2. DB-OP-02000 RPS, SFAS, SFRCS Trip, or SG Tubing Rupture

3. NEI 99-01 SU5 EAL BASES DOCUMENT Rev. 0 Page 182 of 258

ATTACHMENT 1 EAL Bases Category: S - System Malfunction Subcategory: 7 - Loss of Communications Initiating Condition: Loss of ALL onsite or offsite communications capabilities EAL:

SU7.1 Unusual Event Loss of ALL Table S-4 onsite communication methods Table S-4 Communication Methods System Onsite ORO NRC Public Address (Gaitronics) X Onsite Radios X Plant Telephones X X X Commercial Telephones X X X 4-Way Ringdown Circuit X NRC Emergency Telephone System X Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Basis:

This IC addresses a significant loss of on-site or offsite communications capabilities. While not a direct challenge to plant or personnel safety, this event warrants prompt notifications to OROs and the NRC.

This IC should be assessed only when extraordinary means are being utilized to make communications possible (e.g., use of non-plant, privately owned equipment, relaying of on-site information via individuals or multiple radio transmission points, individuals being sent to offsite locations, etc.).

This EAL addresses a total loss of the communications methods used in support of routine plant operations.

Onsite/Offsite Response Organization (ORO)/NRC communications include one or more of the systems listed in Table S-4 (ref. 1, 2).

EAL BASES DOCUMENT Rev. 0 Page 183 of 258

ATTACHMENT 1 EAL Bases Public Address (Gaitronics) System The DBNPS public address system provides paging and party line communications between stations located throughout the plant. Inside and outside type wall and desk-mounted stations are used to communicate between roaming personnel and fixed work locations. Plant-wide instructions are issued using the paging feature.

On-site Radio System Radio systems can be used for communication among operators, off-site monitoring teams, the control room, security, TSC and EOF.

Plant Telephone System The DBNPS plant telephone system provides communication capability between telephone stations located within the plant by dialing the four-digit telephone station code as well as external or offsite calling capability.

Commercial Telephones Commercial telephone lines, which supply public telephone communications, are employed by DBNPS. The local service provider provides primary and secondary power for their lines at the Central Office.

4-Way Ringdown Circuit Dedicated ring down line that includes the State and County EOCs, the Ohio Highway Patrol Office, the Lucas County and Ottawa County Sheriff's dispatcher offices, the Emergency Operations Facility, and the Control Room.

NRC Emergency Telephone System The NRC uses a DBNPS dedicated telephone line, which allows direct telephone communications from the plant to NRC regional and national offices. The DBNPS communications line provides a link independent of the local public telephone network.

Telephones connected to this network are located in the DBNPS Control Room, Technical Support Center, and Emergency Operations Facility and can be used to establish NRC Emergency Notification System (ENS) and Health Physics Network (HPN) capability.

This EAL is the hot condition equivalent of the cold condition EAL CU5.1.

Basis Reference(s):

1. DBNPS UFSAR Section 9.5.2 Communications Systems

2. DBNPS Emergency Plan 7.6 Communications Systems

3. NEI 99-01 SU6 EAL BASES DOCUMENT Rev. 0 Page 184 of 258

ATTACHMENT 1 EAL Bases Category: S - System Malfunction Subcategory: 7 - Loss of Communications Initiating Condition: Loss of ALL onsite or offsite communications capabilities EAL:

SU7.2 Unusual Event Loss of ALL Table S-4 ORO communication methods Table S-4 Communication Methods System Onsite ORO NRC Public Address (Gaitronics) X Onsite Radios X Plant Telephones X X X Commercial Telephones X X X 4-Way Ringdown Circuit X NRC Emergency Telephone System X Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Basis:

This IC addresses a significant loss of on-site or offsite communications capabilities. While not a direct challenge to plant or personnel safety, this event warrants prompt notifications to OROs and the NRC.

This IC should be assessed only when extraordinary means are being utilized to make communications possible (e.g., use of non-plant, privately owned equipment, relaying of on-site information via individuals or multiple radio transmission points, individuals being sent to offsite locations, etc.).

This EAL addresses a total loss of the communications methods used to notify ALL OROs of an emergency declaration. The OROs referred to here are the State of Ohio, Lucas and Ottawa County EOCs.

EAL BASES DOCUMENT Rev. 0 Page 185 of 258

ATTACHMENT 1 EAL Bases Onsite/Offsite Response Organization (ORO)/NRC communications include one or more of the systems listed in Table S-4 (ref. 1, 2).

Public Address (Gaitronics) System The DBNPS public address system provides paging and party line communications between stations located throughout the plant. Inside and outside type wall and desk-mounted stations are used to communicate between roaming personnel and fixed work locations. Plant-wide instructions are issued using the paging feature.

On-site Radio System Radio systems can be used for communication among operators, off-site monitoring teams, the control room, security, TSC and EOF.

Plant Telephone System The DBNPS plant telephone system provides communication capability between telephone stations located within the plant by dialing the four-digit telephone station code as well as external or offsite calling capability.

Commercial Telephones Commercial telephone lines, which supply public telephone communications, are employed by DBNPS. The local service provider provides primary and secondary power for their lines at the Central Office.

4-Way Ringdown Circuit Dedicated ring down line that includes the State and County EOCs, the Ohio Highway Patrol Office, the Lucas County and Ottawa County Sheriff's dispatcher offices, the Emergency Operations Facility, and the Control Room.

NRC Emergency Telephone System The NRC uses a DBNPS dedicated telephone line, which allows direct telephone communications from the plant to NRC regional and national offices. The DBNPS communications line provides a link independent of the local public telephone network.

Telephones connected to this network are located in the DBNPS Control Room, Technical Support Center, and Emergency Operations Facility and can be used to establish NRC Emergency Notification System (ENS) and Health Physics Network (HPN) capability.

This EAL is the hot condition equivalent of the cold condition EAL CU5.2.

Basis Reference(s):

1. DBNPS UFSAR Section 9.5.2 Communications Systems

2. DBNPS Emergency Plan 7.6 Communications Systems

3. NEI 99-01 SU6 EAL BASES DOCUMENT Rev. 0 Page 186 of 258

ATTACHMENT 1 EAL Bases Category: S - System Malfunction Subcategory: 7 - Loss of Communications Initiating Condition: Loss of ALL onsite or offsite communications capabilities EAL:

SU7.3 Unusual Event Loss of ALL Table S-4 NRC communication methods Table S-4 Communication Methods System Onsite ORO NRC Public Address (Gaitronics) X Onsite Radios X Plant Telephones X X X Commercial Telephones X X X 4-Way Ringdown Circuit X NRC Emergency Telephone System X Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Basis:

This IC addresses a significant loss of on-site or offsite communications capabilities. While not a direct challenge to plant or personnel safety, this event warrants prompt notifications to OROs and the NRC.

This IC should be assessed only when extraordinary means are being utilized to make communications possible (e.g., use of non-plant, privately owned equipment, relaying of on-site information via individuals or multiple radio transmission points, individuals being sent to offsite locations, etc.).

This EAL addresses a total loss of the communications methods used to notify the NRC of an emergency declaration.

Onsite/Offsite Response Organization (ORO)/NRC communications include one or more of the systems listed in Table S-4 (ref. 1, 2).

EAL BASES DOCUMENT Rev. 0 Page 187 of 258

ATTACHMENT 1 EAL Bases Public Address (Gaitronics) System The DBNPS public address system provides paging and party line communications between stations located throughout the plant. Inside and outside type wall and desk-mounted stations are used to communicate between roaming personnel and fixed work locations. Plant-wide instructions are issued using the paging feature.

On-site Radio System Radio systems can be used for communication among operators, off-site monitoring teams, the control room, security, TSC and EOF.

Plant Telephone System The DBNPS plant telephone system provides communication capability between telephone stations located within the plant by dialing the four-digit telephone station code as well as external or offsite calling capability.

Commercial Telephones Commercial telephone lines, which supply public telephone communications, are employed by DBNPS. The local service provider provides primary and secondary power for their lines at the Central Office.

4-Way Ringdown Circuit Dedicated ring down line that includes the State and County EOCs, the Ohio Highway Patrol Office, the Lucas County and Ottawa County Sheriff's dispatcher offices, the Emergency Operations Facility, and the Control Room.

NRC Emergency Telephone System The NRC uses a DBNPS dedicated telephone line, which allows direct telephone communications from the plant to NRC regional and national offices. The DBNPS communications line provides a link independent of the local public telephone network.

Telephones connected to this network are located in the DBNPS Control Room, Technical Support Center, and Emergency Operations Facility and can be used to establish NRC Emergency Notification System (ENS) and Health Physics Network (HPN) capability.

This EAL is the hot condition equivalent of the cold condition EAL CU5.3.

Basis Reference(s):

1. DBNPS UFSAR Section 9.5.2 Communications Systems

2. DBNPS Emergency Plan 7.6 Communications Systems

3. NEI 99-01 SU6 EAL BASES DOCUMENT Rev. 0 Page 188 of 258

ATTACHMENT 1 EAL Bases Category: S - System Malfunction Subcategory: 8 - Containment Failure Initiating Condition: Failure to isolate containment or loss of containment pressure control EAL:

SU8.1 Unusual Event ANY penetration is not closed within 15 min. of a VALID containment isolation signal (Note 1)

Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Basis:

This EAL addresses a failure of one or more containment penetrations to automatically isolate (close) when required by an actuation signal. Absent challenges to another fission product barrier, this condition represents potential degradation of the level of safety of the plant.

For this EAL the containment isolation signal must be generated as the result on an off-normal/accident condition (e.g., a safety injection or high containment pressure); a failure resulting from testing or maintenance does not warrant classification. The determination of containment and penetration status - isolated or not isolated - should be made in accordance with the appropriate criteria contained in the plant AOPs and EOPs. The 15-minute criterion is included to allow operators time to manually isolate the required penetrations, if possible.

This event would escalate to a Site Area Emergency in accordance with IC FS1 if there were a concurrent loss or potential loss of either the Fuel Clad or RCS fission product barriers.

Successful closure of any one valve in a penetration line is sufficient to consider the penetration closed.

Basis Reference(s):

1. NEI 99-01 SU7 EAL BASES DOCUMENT Rev. 0 Page 189 of 258

ATTACHMENT 1 EAL Bases Category: S - System Malfunction Subcategory: 8 - Containment Failure Initiating Condition: Failure to isolate containment or loss of containment pressure control EAL:

SU8.2 Unusual Event Containment pressure > 40 psia with < one full train of containment cooling, Table S-6, operating per design for 15 min. (Note 1)

Table S-6 Containment Cooling Full Train CT Spray Pumps CT Cooling Fans 2 0 1 1 0 2 Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Basis:

This EAL addresses an event that results in high containment pressure with a concurrent failure of containment pressure control systems. Absent challenges to another fission product barrier, this condition represents a potential degradation of the level of safety of the plant.

This EAL addresses a condition where containment pressure is greater than the setpoint at which containment energy (heat) removal systems are designed to automatically actuate, and less than one full train of equipment is capable of operating per design. The 15-minute criterion is included to allow operators time to manually start equipment that may not have automatically started, if possible. The inability to start the required equipment indicates that containment heat removal/depressurization systems (e.g., containment sprays or ice condenser fans) are either lost or performing in a degraded manner.

This event would escalate to a Site Area Emergency in accordance with IC FS1 if there were a concurrent loss or potential loss of either the Fuel Clad or RCS fission product barriers.

EAL BASES DOCUMENT Rev. 0 Page 190 of 258

ATTACHMENT 1 EAL Bases The combination of Containment spray pumps and Containment cooling fan units considered to be a full train of containment cooling operating per design is shown in Table F-3 (ref. 1).

SFAS 2 actuation automatically initiates Containment Air Coolers upon exceeding the Containment pressure high setpoint of 18.7 psia or low RCS pressure of 1600 psig. SFAS Level 4 actuation automatically initiates Containment Spray upon exceeding the Containment pressure high-high setpoint of 40 (nominal) psia (ref. 2, 3).

Basis Reference(s):

1. UFSAR Section 6.2.2. Containment Vessel Heat Removal Systems

2. DB-OP-02000 RPS, SFAS, SFRCS Trip, or SG Tubing Rupture

3. DBNPS Technical Specifications Table 3.3.5-1 Safety Features Actuation System Instrumentation

4. NEI 99-01 SU7 EAL BASES DOCUMENT Rev. 0 Page 191 of 258

ATTACHMENT 1 EAL Bases Category: S - System Malfunction Subcategory: 9 - Hazardous Event Affecting SAFETY SYSTEMS Initiating Condition: Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode EAL:

SA9.1 Alert The occurrence of ANY Table S-5 Hazardous Event AND EITHER:

Event has caused indications of degraded performance in at least one train of a SAFETY SYSTEM required for the current operating mode The event has caused VISIBLE DAMAGE to a SAFETY SYSTEM component or structure required for the current operating mode Table S-5 Hazardous Events Seismic event (earthquake)

Internal or external FLOODING event High winds or tornado strike FIRE EXPLOSION Other events with similar hazard characteristics as determined by the Emergency Director Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Basis:

This IC addresses a hazardous event that causes damage to a SAFETY SYSTEM, or a structure containing SAFETY SYSTEM components, needed for the current operating mode.

This condition significantly reduces the margin to a loss or potential loss of a fission product barrier, and therefore represents an actual or potential substantial degradation of the level of safety of the plant.

The first conditional addresses damage to a SAFETY SYSTEM train that is in service/operation since indications for it will be readily available. The indications of degraded performance should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train.

EAL BASES DOCUMENT Rev. 0 Page 192 of 258

ATTACHMENT 1 EAL Bases The second conditional addresses damage to a SAFETY SYSTEM component that is not in service/operation or readily apparent through indications alone, or to a structure containing SAFETY SYSTEM components. Operators will make this determination based on the totality of available event and damage report information. This is intended to be a brief assessment not requiring lengthy analysis or quantification of the damage.

Escalation of the emergency classification level would be via IC FS1 or RS1.

Ground motion acceleration of 0.08g horizontal or 0.053g vertical is the Maximum Probable Earthquake as is considered generically as the Operating Basis Earthquake for DBNPS (ref. 8). Control room alarm indication of an earthquake greater than OBE is indicated on the seismic control panel (C5764A). RA-EP-02820 Earthquake provides the guidance for determining any required response actions if the OBE earthquake threshold is exceeded (ref. 1). The significance of seismic events is discussed under EAL HU2.1.

Internal FLOODING occurs from breaches of water systems that are located inside plant buildings and are connected to large water sources such as Intake Forebay or tanks (ref. 2).

External FLOODING may be due to high lake level. DBNPS flood emergency elevation is 578 ft. Site access would be limited to rail, boat or helicopter (ref. 3).

Seismic Category I structures are analyzed to withstand a sustained, design wind velocity of at least 90 mph. (ref. 4).

Areas containing functions and systems required for safe shutdown of the plant are identified by fire area in the fire abnormal procedure (ref. 5).

An EXPLOSION that degrades the performance of a SAFETY SYSTEM train or visibly damages a SAFETY SYSTEM component or structure would be classified under this EAL (ref. 6).

Basis Reference(s):

1. RA-EP-02820 Earthquake

2. RA-EP-02880 Internal Flooding

3. RA-EP-02830 Flooding

4. DBNPS UFSAR Section 3.3.1 Wind Criteria

5. DB-OP-02501 Serious Station Fire

6. RA-EP-02840 Explosion

7. NEI 99-01 SA9

8. Updated FSAR Section 3.1 Seismic Design EAL BASES DOCUMENT Rev. 0 Page 193 of 258

ATTACHMENT 1 EAL Bases Category F - Fission Product Barrier Degradation EAL Group: Hot Conditions (RCS temperature > 200ºF); EALs in this category are applicable only in one or more hot operating modes.

EALs in this category represent threats to the defense in depth design concept that precludes the release of highly radioactive fission products to the environment. This concept relies on multiple physical barriers any one of which, if maintained intact, precludes the release of significant amounts of radioactive fission products to the environment. The primary fission product barriers are:

A. Fuel Clad (FC): The Fuel Clad Barrier consists of the cladding material that contains the fuel pellets.

B. Reactor Coolant System (RCS): The RCS Barrier includes the RCS primary side and its connections up to and including the pressurizer safety and relief valves, and other connections up to and including the primary isolation valves.

C. Containment (CT): The Containment Barrier includes the containment building, connections up to and including the outermost containment isolation valves. This barrier also includes the main steam, feedwater, and blowdown line extensions outside the containment building up to and including the outermost secondary side isolation valve. Containment Barrier thresholds are used as criteria for escalation of the ECL from Alert to a Site Area Emergency or a General Emergency.

The EALs in this category require evaluation of the loss and potential loss thresholds listed in the fission product barrier matrix of Table F-1 (Attachment 2). Loss and Potential Loss signify the relative damage and threat of damage to the barrier. Loss means the barrier no longer assures containment of radioactive materials. Potential Loss means integrity of the barrier is threatened and could be lost if conditions continue to degrade. The number of barriers that are lost or potentially lost and the following criteria determine the appropriate emergency classification level:

Alert:

ANY loss or ANY potential loss of EITHER Fuel Clad or RCS Site Area Emergency:

Loss or potential loss of ANY two barriers General Emergency:

Loss of ANY two barriers AND loss or potential loss of third barrier EAL BASES DOCUMENT Rev. 0 Page 194 of 258

ATTACHMENT 1 EAL Bases The logic used for emergency classification based on fission product barrier monitoring should reflect the following considerations:

The Fuel Clad Barrier and the RCS Barrier are weighted more heavily than the Containment Barrier.

Unusual Event ICs associated with RCS and Fuel Clad Barriers are addressed under System Malfunction ICs.

For accident conditions involving a radiological release, evaluation of the FISSION PRODUCT BARRIER THRESHOLDS will need to be performed in conjunction with dose assessments to ensure correct and timely escalation of the emergency classification. For example, an evaluation of the FISSION PRODUCT BARRIER THRESHOLDS may result in a Site Area Emergency classification while a dose assessment may indicate that an EAL for General Emergency IC RG1 has been exceeded.

The FISSION PRODUCT BARRIER THRESHOLDS specified within a scheme reflect plant-specific DBNPS design and operating characteristics.

As used in this category, the term RCS leakage encompasses not just those types defined in Technical Specifications but also includes the loss of RCS mass to any location- inside the containment, an interfacing system, or outside of the containment.

The release of liquid or steam mass from the RCS due to the as-designed/expected operation of a relief valve is not considered RCS leakage.

At the Site Area Emergency level, EAL users should maintain cognizance of how far present conditions are from meeting a threshold that would require a General Emergency declaration. For example, if the Fuel Clad and RCS fission product barriers were both lost, then there should be frequent assessments of containment radioactive inventory and integrity. Alternatively, if both the Fuel Clad and RCS fission product barriers were potentially lost, the Emergency Director would have more assurance that there was no immediate need to escalate to a General Emergency.

EAL BASES DOCUMENT Rev. 0 Page 195 of 258

ATTACHMENT 1 EAL Bases Category: Fission Product Barrier Degradation Subcategory: N/A Initiating Condition: Loss of ANY two barriers AND Loss or Potential Loss of third barrier EAL:

FG1.1 General Emergency Loss of ANY two barriers AND Loss or Potential Loss of third barrier (Table F-1)

Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Basis:

Fuel Clad, RCS and Containment 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 Basis Reference(s):

1. NEI 99-01 FG1 EAL BASES DOCUMENT Rev. 0 Page 196 of 258

ATTACHMENT 1 EAL Bases 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-1)

Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Basis:

Fuel Clad, RCS and Containment comprise the fission product barriers. Table F-1 (Attachment 2) lists the FISSION PRODUCT BARRIER THRESHOLDS, bases and references.

At the Site Area Emergency classification level, each barrier is weighted equally. A Site Area Emergency is therefore appropriate for ANY combination of the following conditions:

One barrier loss AND a second barrier loss (i.e., loss - loss)

One barrier loss AND a second barrier potential loss (i.e., loss - potential loss)

One barrier potential loss AND a second barrier potential loss (i.e., potential loss -

potential loss)

At the Site Area Emergency classification level, the ability to 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.

Basis Reference(s):

1. NEI 99-01 FS1 EAL BASES DOCUMENT Rev. 0 Page 197 of 258

ATTACHMENT 1 EAL Bases 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-1)

Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Basis:

Fuel Clad, RCS and Containment 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 FS1.1 Basis Reference(s):

1. NEI 99-01 FA1 EAL BASES DOCUMENT Rev. 0 Page 198 of 258

ATTACHMENT 2 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. RCS or SG Tube Leakage B. Inadequate Heat removal C. CT Radiation / RCS Activity D. CT Integrity or Bypass E. Emergency Director 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 in Category C would be assigned CT P-Loss C.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-1, 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.

EAL BASES DOCUMENT Rev. 0 Page 199 of 258

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases If the EAL-user determines that ANY threshold has been exceeded, by definition, the barrier is lost or potentially lost - even if multiple thresholds in the same barrier column are exceeded; only that one barrier is lost or potentially lost. The EAL-user must examine each of the three fission product barriers to determine if other barrier thresholds in the category are lost or potentially lost. For example, if containment radiation is sufficiently high, a Loss of the Fuel Clad and RCS barriers and a Potential Loss of the Containment barrier can occur. Barrier Losses and Potential Losses are then applied to the algorithms given in EALs FG1.1, FS1.1, and FA1.1 to determine the appropriate emergency classification.

In the remainder of this Attachment, the Fuel Clad barrier threshold bases appear first, followed by the RCS barrier and finally the Containment barrier threshold bases. In each barrier, the bases are given according category Loss followed by category Potential Loss beginning with Category A, then B E.

EAL BASES DOCUMENT Rev. 0 Page 200 of 258

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Table F-1 Fission Product Barrier Threshold Matrix Fuel Clad (FC) Barrier Reactor Coolant System (RCS) Barrier Containment (CT) Barrier Category Loss Potential Loss Loss Potential Loss Loss Potential Loss

1. Operation of a standby Makeup

1. An automatic or manual ECCS Pump (>250 gpm) is required by A (SFAS) actuation required by EITHER:

EITHER:

UNISOLABLE RCS leakage 1. A leaking or RUPTURED SG is RCS or None None None UNISOLABLE RCS FAULTED outside of containment SG Tube SG tube leakage leakage Leakage OR SG tube RUPTURE

2. PTS requirements invoked (SR5)

1. Calculated Clad Temperature in Region 2 or higher (DB-OP-02000 1. Calculated Clad Temperature in B 1. Calculated Clad Temperature Figure 2)

1. Loss of ALL feedwater Region 3 or higher (DB-OP-02000 Figure 2)

OR Inadequate in Region 3 or higher (DB- None AND None AND Heat OP-02000 Figure 2) 2. Loss of ALL feedwater SG Cooling is required Restoration procedures not Removal AND effective within 15 min. (Note 1)

SG Cooling is required C 1. RE 4596A or B > Table F-2 column FC Loss (Note 9)

CT OR None

1. RE 4596A or B > Table F-2 None None

1. RE 4596A or B > Table F-2 column Radiation column RCS Loss (Note 9) CT Potential Loss (Note 9)

/ RCS 2. Dose equivalent I-131 coolant activity > 300 µCi/gm Activity

1. Containment isolation is required

1. Containment pressure > 50.4 psia AND EITHER:

OR Containment integrity has D been lost based on Emergency Director

2. Containment hydrogen concentration > 4%

judgment CT None None None None OR Integrity UNISOLABLE pathway from Containment to the environment 3. Containment pressure > 40 psia or Bypass exists with < one full train, Table F-3, of containment cooling OR operating per design for 15 min.

(Note 1)

2. Indications of RCS leakage outside of containment E 1. ANY condition in the opinion of the Emergency Director that

1. ANY condition in the opinion of the Emergency Director that

1. ANY condition in the opinion of 1. ANY condition in the opinion of the

1. ANY condition in the opinion of the Emergency Director that

1. ANY condition in the opinion of the Emergency Director that indicates the Emergency Director that Emergency Director that indicates ED indicates Loss of the Fuel Clad indicates Potential Loss of the indicates Loss of the Potential Loss of the Containment indicates Loss of the RCS Barrier Potential Loss of the RCS Barrier Barrier Fuel Clad Barrier Containment Barrier Barrier Judgment EAL BASES DOCUMENT Rev. 0 Page 201 of 258

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Table F-2 Containment Radiation - R/hr (RE 4596A or B)

Time After S/D RCS Loss FC Loss CT Potential Loss (Hrs.)

0-1 1.50E+01 3.03E+03 1.40E+04 1-2 1.50E+01 2.56 E+03 1.18 E+04 2-8 1.50E+01 1.61 E+03 7.46 E+03 8-16 1.50E+01 1.14 E+03 5.28 E+03 16-24 1.50E+01 8.66 E+02 4.00 E+03

>24 1.50E+01 3.94 E+02 1.82 E+03 Table F-3 Containment Cooling Full Train Spray Coolers 2 0 1 1 0 2 EAL BASES DOCUMENT Rev. 0 Page 202 of 258

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: A. RCS or SG Tube Leakage Degradation Threat: Loss Threshold:

None EAL BASES DOCUMENT Rev. 0 Page 203 of 258

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: A. RCS or SG Tube Leakage Degradation Threat: Potential Loss Threshold:

None EAL BASES DOCUMENT Rev. 0 Page 204 of 258

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: B. Inadequate Heat Removal Degradation Threat: Loss Threshold:

1. Calculated Clad Temperature in Region 3 or higher (DB-OP-02000 Figure 2)

Basis:

This reading indicates temperatures within the core have caused significant superheating of reactor coolant.

Indication of severe core cooling degradation is manifested by Calculated Clad Temperature in Region 3 or higher (DB-OP-02000 Figure 2). Figure 2, Incore T/C Temperature vs. RCS Pressure for ICC, provides indication of how serious core conditions are based upon combinations of RCS pressure and incore thermocouple temperatures. If the RCS P-T point is in Region 3, the cladding temperatures in the core may be 1400°F or higher. This is a very serious condition and may lead to significant amounts of H2 production; core damage may be unavoidable as this represents a very serious inadequate core cooling condition (ref. 1, 2).

WCAP-14969-A states, "Analyses performed for the WOG ERGs for indication of inadequate core cooling concluded that the temperature indicated by the core exit thermocouples, especially during transient heatup conditions, is always several hundred degrees lower than the fuel rod cladding temperatures. Thus, an indicated temperature of 1200°F can be translated to a peak cladding temperature on the order of 1400°F (ref. 4).

Basis Reference(s):

1. DB-OP-02000 RPS, SFAS, SFRCS Trip, or SG Tubing Rupture

2. Bases and Deviation Document for DB-OP-02000

3. NEI 99-01 Inadequate Heat Removal Fuel Clad Loss 2.A

4. WCAP-14969-A, Westinghouse Owners Group Core Damage Assessment Guidance EAL BASES DOCUMENT Rev. 0 Page 205 of 258

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: B. Inadequate Heat Removal Degradation Threat: Potential Loss Threshold:

1. Calculated Clad Temperature in Region 2 or higher (DB-OP-02000 Figure 2)

Basis:

This reading indicates a reduction in reactor vessel water level sufficient to allow the onset of heat-induced cladding damage.

The average incore thermocouple temperature and RCS pressure is used to determine whether Calculated Clad Temperature is in Region 2. This corresponds to a loss of RCS subcooling with clad temperatures remaining below the point where damage is immediately likely (Tclad approximately 900° to 1100° F).

Basis Reference(s):

1. DB-OP-02000 RPS, SFAS, SFRCS Trip, or SG Tubing Rupture

2. Bases and Deviation Document for DB-OP-02000

3. NEI 99-01 Inadequate Heat Removal Fuel Clad Potential Loss 2.A EAL BASES DOCUMENT Rev. 0 Page 206 of 258

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: B. Inadequate Heat Removal Degradation Threat: Potential Loss Threshold:

2. Loss of ALL feedwater AND SG cooling is required Basis:

This condition indicates an extreme challenge to the ability to remove RCS heat using the steam generators (i.e., loss of an effective secondary-side heat sink). This condition represents a potential loss of the Fuel Clad Barrier. In accordance with EOPs, there may be unusual accident conditions during which operators intentionally reduce the heat removal capability of the steam generators; during these conditions, classification using threshold is not warranted.

In combination with RCS Potential Loss B.1, meeting this threshold would result in a Site Area Emergency.

Loss of ALL feedwater cooling heat transfer capability when SG cooling is required indicates the ultimate heat sink function is under extreme challenge AND that the RCS barrier is also challenged (ref. 1).

The phrase AND SG cooling is required precludes the need for classification for conditions in which RCS pressure is less than SG pressure. For large LOCA events inside the Containment, the SGs are moot because heat removal through the containment heat removal systems takes place. Therefore, SG cooling should not be required and, should not be assessed for EAL classification because a LOCA event alone should not require higher than an Alert classification (ref. 1, 2).

Basis Reference(s):

1. DB-OP-02000 RPS, SFAS, SFRCS Trip, or SG Tubing Rupture

2. Bases and Deviation Document for DB-OP-02000

3. NEI 99-01 Inadequate Heat Removal Fuel Clad Loss 2.B EAL BASES DOCUMENT Rev. 0 Page 207 of 258

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: C. CT Radiation / RCS Activity Degradation Threat: Loss Threshold:

1. RE 4596A or B > Table F-2 column FC Loss (Note 9)

Table F-2 Containment Radiation - R/hr (RE 4596A or B)

Time After S/D RCS Loss FC Loss CT Potential Loss (Hrs.)

0-1 1.50E+01 3.03E+03 1.40E+04 1-2 1.50E+01 2.56 E+03 1.18 E+04 2-8 1.50E+01 1.61 E+03 7.46 E+03 8-16 1.50E+01 1.14 E+03 5.28 E+03 16-24 1.50E+01 8.66 E+02 4.00 E+03

>24 1.50E+01 3.94 E+02 1.82 E+03 Note 9: During a main steam line break in containment or LOCA with temperature >170F, there is a potential to induce transient errors into the output of RE4596A and B during the peak rate of temperature change.

Consult alternate indications. If the main steam line break is accompanied by core damage this error is insignificant (ref. 4, 5, 6, 7, 8).

Basis:

The radiation monitor reading corresponds to an instantaneous release of ALL reactor coolant mass into the containment, assuming that reactor coolant activity equals 300 µCi/gm dose equivalent I-131. Reactor coolant activity above this level is greater than that expected for iodine spikes and corresponds to an approximate range of 2% to 5% fuel clad damage. Since this condition indicates that a significant amount of fuel clad damage has occurred, it represents a loss of the Fuel Clad Barrier.

The radiation monitor reading in this threshold is higher than that specified for RCS Barrier Loss threshold C.1 since it indicates a loss of both the Fuel Clad Barrier and the RCS Barrier.

Note that a combination of the two monitor readings appropriately escalates the ECL to a Site Area Emergency.

The containment high range monitors, RE 4596A & B., monitor the gamma dose rate resulting from a postulated loss of coolant accident (LOCA). RE 4596A & B are located inside containment. The detector range is approximately 1 to 1E8 R/hr (logarithmic scale) (ref. 1).

The Table F-2 values, column FC Loss represents, based on Calculation EP-EALCALC-DB-0701, the expected containment high range radiation monitor (RE 4596A & B) response based EAL BASES DOCUMENT Rev. 0 Page 208 of 258

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases on a LOCA, for periods of 1, 2, 8, 16, 24 and 48 (>24) hours after shutdown with ~4.33% fuel failure (ref. 2).

When evaluating fission product barrier integrity values in Table F-2, time after shutdown should be confirmed with the Control Room as the time that the reactor is tripped, and reactor power is verified to be lowering on the Intermediate Range (ref. 3). If time after shutdown is less than one hour (or the reactor is still critical), the 0-1 hour after shutdown value should be chosen. This is conservative as it represents sufficient time for plant conditions to deteriorate to the point that core damage may occur, and the activity released from the RCS into the containment atmosphere to reach equilibrium mixing throughout containment.

During a main steam line break in containment or LOCA with temperature > 170ºF, there is a potential to induce transient errors (positive and negative) into the output of RE4596A & B during the peak rate of temperature change. Consult alternate indications. If the main steam line break or LOCA is accompanied by core damage this error is, however, insignificant (ref. 4).

Basis Reference(s):

1. UFSAR Section 7.13.3.1 Containment High Radiation Monitors

2. EP-EALCALC-DB-0701 Containment Radiation Monitor Readings Following Clad Damage (FC2 and CT2 Potential Loss)

3. DB-OP-02000 RPS, SFAS, SFRCS Trip, or SG Tubing Rupture

4. NRC Information Notice 97-45 Supplement 1 Environmental Qualification Deficiency for Cables and Containment Penetration Pigtails

5. Condition Report 09-53277, Containment High Range Rad Monitor Function and Current EALS

6. Condition Report 09-53278, Containment High Range Rad Monitor Function and NEI 99-01 EAL Submittal

7. Condition Report 07-31108, Potential for Thermally Induced Currents In Containment HRRM

8. Condition Report 09-55171, Containment High Range Rad Monitors Engineering Assistance Requested

9. NEI 99-01 CMT Radiation / RCS Activity Fuel Clad Loss 3.A EAL BASES DOCUMENT Rev. 0 Page 209 of 258

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: C. CT Radiation / RCS Activity Degradation Threat: Loss Threshold:

2. Dose equivalent I-131 coolant activity > 300 µCi/gm Basis:

This threshold indicates that RCS radioactivity concentration is greater than 300 µCi/gm dose equivalent I-131. Reactor coolant activity above this level is greater than that expected for iodine spikes and corresponds to an approximate range of 2% to 5% fuel clad damage. Since this condition indicates that a significant amount of fuel clad damage has occurred, it represents a loss of the Fuel Clad Barrier.

Elevated reactor coolant activity represents a potential degradation in the level of safety of the plant and a potential precursor of more serious problems. The threshold dose equivalent I-131 concentration is well above that expected for iodine spikes and corresponds to about 4.33% fuel clad damage. When reactor coolant activity reaches this level the Fuel Clad barrier is considered lost (ref. 1).

There is no Potential Loss threshold associated with RCS Activity / Containment Radiation.

Basis Reference(s):

1. EP-EALCALC-DB-0701 Containment Radiation Monitor Readings Following Clad Damage (FC2 and CT2 Potential Loss)

2. NEI 99-01 CMT Radiation / RCS Activity Fuel Clad Loss 3.B EAL BASES DOCUMENT Rev. 0 Page 210 of 258

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: C. CT Radiation / RCS Activity Degradation Threat: Potential Loss Threshold:

None EAL BASES DOCUMENT Rev. 0 Page 211 of 258

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: D. CT Integrity or Bypass Degradation Threat: Loss Threshold:

None EAL BASES DOCUMENT Rev. 0 Page 212 of 258

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: D. CT Integrity or Bypass Degradation Threat: Potential Loss Threshold:

None EAL BASES DOCUMENT Rev. 0 Page 213 of 258

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: E. Emergency Director Judgment Degradation Threat: Loss Threshold:

1. ANY condition in the opinion of the Emergency Director that indicates Loss of the Fuel Clad Barrier 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.

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 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.

Basis Reference(s):

1. NEI 99-01 Emergency Director Judgment Fuel Clad Loss 6.A EAL BASES DOCUMENT Rev. 0 Page 214 of 258

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: E. Emergency Director Judgment Degradation Threat: Potential Loss Threshold:

1. ANY condition in the opinion of the Emergency Director that indicates Potential Loss of the Fuel Clad Barrier 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. The Emergency Director should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored.

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 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.

Basis Reference(s):

1. NEI 99-01 Emergency Director Judgment Potential Fuel Clad Loss 6.A EAL BASES DOCUMENT Rev. 0 Page 215 of 258

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category: A. RCS or SG Tube Leakage Degradation Threat: Loss Threshold:

1. An automatic or manual ECCS (SFAS) actuation required by EITHER:

UNISOLABLE RCS leakage SG tube RUPTURE Basis:

This threshold is based on an UNISOLABLE RCS leak of sufficient size to require an automatic or manual actuation of the Emergency Core Cooling System (ECCS). This condition clearly represents a loss of the RCS Barrier.

This threshold is applicable to unidentified and pressure boundary leakage, as well as identified leakage. It is also applicable to UNISOLABLE RCS leakage through an interfacing system. The mass loss may be into any location - inside containment, to the secondary-side (i.e., steam generator tube leakage) or outside of containment.

A steam generator with primary-to-secondary leakage of sufficient magnitude to require a safety injection is considered to be RUPTURED. If a RUPTURED steam generator is also FAULTED outside of containment, the declaration escalates to a Site Area Emergency since the Containment Barrier Loss threshold A.1 will also be met.

ECCS (SFAS) actuation is caused by (ref. 1):

Low RCS pressure High Containment pressure Basis Reference(s):

1. DB-OP-02000 RPS, SFAS, SFRCS Trip, or SG Tubing Rupture

2. NEI 99-01 RCS or SG Tube Leakage Reactor Coolant System Loss 1.A EAL BASES DOCUMENT Rev. 0 Page 216 of 258

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category: A. RCS or SG Tube Leakage Degradation Threat: Potential Loss Threshold:

1. Operation of a standby Makeup Pump is required (> 250 gpm) by EITHER:

UNISOLABLE RCS leakage SG tube leakage Basis:

This threshold is based on an UNISOLABLE RCS leak that results in the inability to maintain pressurizer level within specified limits by operation of a normally used charging (makeup) pump, but an ECCS (SI) actuation has not occurred. The threshold is met when an operating procedure, or operating crew supervision, directs that a standby charging (makeup) pump be placed in service to restore and maintain pressurizer level.

This threshold is applicable to unidentified and pressure boundary leakage, as well as identified leakage. It is also applicable to UNISOLABLE RCS leakage through an interfacing system. The mass loss may be into any location - inside containment, to the secondary-side (i.e., steam generator tube leakage) or outside of containment.

If a leaking steam generator is also FAULTED outside of containment, the declaration escalates to a Site Area Emergency since the Containment Barrier Loss threshold 1.A will also be met.

The Makeup and Purification System includes two centrifugal makeup pumps, which take suction from the Makeup Tank, and return cooled, purified reactor coolant to the RCS. One of the two makeup pumps handles normal charging flow. Makeup pump capacity of a single makeup injection line is ~250 gpm. A second makeup pump being required is indicative of a substantial RCS leak (ref. 1, 2).

Basis Reference(s):

1. UFSAR Table 9.3-8 Makeup and Purification System Component Data

2. DB-OP-02000 RPS, SFAS, SFRCS Trip, or SG Tubing Rupture

3. NEI 99-01 RCS or SG Tube Leakage Reactor Coolant System Potential Loss 1.A EAL BASES DOCUMENT Rev. 0 Page 217 of 258

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category: A. RCS or SG Tube Leakage Degradation Threat: Potential Loss Threshold:

2. PTS requirements invoked (SR5)

Basis:

This condition indicates an extreme challenge to the integrity of the RCS pressure boundary due to pressurized thermal shock - a transient that causes rapid RCS cooldown while the RCS is in Mode 3 or higher (i.e., hot and pressurized).

With an extended overcooling, thermal shock becomes a concern. Pressurized Thermal Shock (PTS) limits must be invoked if the criteria specified in Specific Rule 5 are met. The RCS pressure must be controlled to ensure that PTS limits are not violated. This requires action on the part of the operator to control RCS pressure and temperature (ref. 1).

The "Potential Loss" threshold is defined by the PTS limits as specified in DB-OP-02000 RPS, SFAS, SFRCS Trip, or SG Tubing Rupture Specific Rule 5 being invoked. This indicates an extreme challenge to the RCS barrier (ref. 1, 2).

Basis Reference(s):

1. DB-OP-02000 RPS, SFAS, SFRCS Trip, or SG Tubing Rupture

2. Bases and Deviation Document for DB-OP-02000

3. NEI 99-01 RCS or SG Tube Leakage Reactor Coolant System Potential Loss 1.B EAL BASES DOCUMENT Rev. 0 Page 218 of 258

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category: B. Inadequate Heat Removal Degradation Threat: Loss Threshold:

None EAL BASES DOCUMENT Rev. 0 Page 219 of 258

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category: B. Inadequate Heat Removal Degradation Threat: Potential Loss Threshold:

1. Loss of ALL feedwater AND SG cooling is required Basis:

This condition indicates an extreme challenge to the ability to remove RCS heat using the steam generators (i.e., loss of an effective secondary-side heat sink). This condition represents a potential loss of the RCS Barrier. In accordance with EOPs, there may be unusual accident conditions during which operators intentionally reduce the heat removal capability of the steam generators; during these conditions, classification using threshold is not warranted.

Meeting this threshold results in a Site Area Emergency because this threshold is identical to Fuel Clad Barrier Potential Loss threshold B.2; both will be met. This condition warrants a Site Area Emergency declaration because inadequate RCS heat removal may result in fuel heat-up sufficient to damage the cladding and increase RCS pressure to the point where mass will be lost from the system.

Loss of ALL feedwater cooling heat transfer capability when SG cooling is required indicates the ultimate heat sink function is under extreme challenge and that the RCS barrier is also challenged (ref. 1).

The phrase AND SG cooling required precludes the need for classification for conditions in which RCS pressure is less than SG pressure. For large LOCA events inside the Containment, the SGs are moot because heat removal through the containment heat removal systems takes place. Therefore, SG cooling should not be required and, should not be assessed for EAL classification because a LOCA event alone should not require higher than an Alert classification (ref. 1, 2).

Basis Reference(s):

1. DB-OP-02000 RPS, SFAS, SFRCS Trip, or SG Tubing Rupture

2. Bases and Deviation Document for DB-OP-02000

3. NEI 99-01 Inadequate Heat Removal RCS Loss 2.B EAL BASES DOCUMENT Rev. 0 Page 220 of 258

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category: C. CT Radiation/ RCS Activity Degradation Threat: Loss Threshold:

1. RE 4596A or B > Table F-2 column RCS Loss (Note 9)

Table F-2 Containment Radiation - R/hr (RE 4596A or B)

Time After S/D RCS Loss FC Loss CT Potential Loss (Hrs.)

0-1 1.50E+01 3.03E+03 1.40E+04 1-2 1.50E+01 2.56 E+03 1.18 E+04 2-8 1.50E+01 1.61 E+03 7.46 E+03 8-16 1.50E+01 1.14 E+03 5.28 E+03 16-24 1.50E+01 8.66 E+02 4.00 E+03

>24 1.50E+01 3.94 E+02 1.82 E+03 Note 9: During a main steam line break in containment or LOCA with temperature >170F, there is a potential to induce transient errors into the output of RE4596A and B during the peak rate of temperature change.

Consult alternate indications. If the main steam line break is accompanied by core damage this error is insignificant (ref. 5, 6, 7, 8, 9).

Basis:

The radiation monitor reading corresponds to an instantaneous release of ALL reactor coolant mass into the containment, assuming that reactor coolant activity equals Technical Specification allowable limits. This value is lower than that specified for Fuel Clad Barrier Loss threshold C.1 since it indicates a loss of the RCS Barrier only.

There is no Potential Loss threshold associated with RCS Activity / Containment Radiation.

The containment high range monitors, RE 4596A & B., monitor the gamma dose rate resulting from a postulated loss of coolant accident (LOCA). RE 4596A & B are located inside containment. The detector range is approximately 1 to 1E8 R/hr (logarithmic scale) (ref. 1).

The Table F-2 values, column RCS Loss represents, based on Calculation EP-EALCALC-DB-0702, the expected containment high range radiation monitor (RE 4596A &

B) response based on a LOCA, using the USAR maximum RCS activity (no core damage)

(ref. 2, 3).

EAL BASES DOCUMENT Rev. 0 Page 221 of 258

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases During a main steam line break in containment or LOCA with temperature > 170ºF, there is a potential to induce transient errors (positive and negative) into the output of RE4596A & B during the peak rate of temperature change. Consult alternate indications. If the main steam line break or LOCA is accompanied by core damage this error is, however, insignificant (ref. 5).

Since fuel cladding degradation and/or failures could also result in high Containment Area Radiation levels, a reading of >15 R/hr may be obtained without a physical loss of the RCS barrier. However, this threshold should be declared as being met if the Containment Area Radiation levels are >15 R/hr even if there are no other indications of a RCS leak or physical loss. In this case it would also be prudent to evaluate the fuel clad FISSION PRODUCT BARRIER THRESHOLDS.

Basis Reference(s):

1. UFSAR Section 7.13.3.1 Containment High Radiation Monitors

2. EP-EALCALC-DB-0702 Containment Radiation Monitor Readings Following a LOCA (RC2 Loss)

3. USAR Table 15A-4 Maximum Fission Product Activity in Reactor Coolant

4. DB-OP-02000 RPS, SFAS, SFRCS Trip, or SG Tubing Rupture

5. NRC Information Notice 97-45 Supplement 1 Environmental Qualification Deficiency for Cables and Containment Penetration Pigtails

6. Condition Report 09-53277, Containment High Range Rad Monitor Function and Current EALS

7. Condition Report 09-53278, Containment High Range Rad Monitor Function and NEI 99-01 EAL Submittal

8. Condition Report 07-31108, Potential for Thermally Induced Currents In Containment HRRM

9. Condition Report 09-55171, Containment High Range Rad Monitors Engineering Assistance Requested

10. NEI 99-01 CMT Radiation / RCS Activity RCS Loss 3.A EAL BASES DOCUMENT Rev. 0 Page 222 of 258

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category: B. CT Radiation/ RCS Activity Degradation Threat: Potential Loss Threshold:

None EAL BASES DOCUMENT Rev. 0 Page 223 of 258

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category: D. CT Integrity or Bypass Degradation Threat: Loss Threshold:

None EAL BASES DOCUMENT Rev. 0 Page 224 of 258

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category: D. CT Integrity or Bypass Degradation Threat: Potential Loss Threshold:

None EAL BASES DOCUMENT Rev. 0 Page 225 of 258

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category: E. Emergency Director Judgment Degradation Threat: Loss Threshold:

1. ANY condition in the opinion of the Emergency Director that indicates Loss of the RCS Barrier Basis:

This threshold addresses any other factors that may be used by the Emergency Director in determining whether the RCS Barrier is lost.

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 if the 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.

Basis Reference(s):

1. NEI 99-01 Emergency Director Judgment RCS Loss 6.A EAL BASES DOCUMENT Rev. 0 Page 226 of 258

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category: E. Emergency Director Judgment Degradation Threat: Potential Loss Threshold:

1. ANY condition in the opinion of the Emergency Director that indicates Potential Loss of the RCS Barrier Basis:

This threshold addresses any other factors that may be used by the Emergency Director in determining whether the RCS Barrier is potentially lost. The Emergency Director should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored.

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.

Basis Reference(s):

1. NEI 99-01 Emergency Director Judgment RCS Potential Loss 6.A EAL BASES DOCUMENT Rev. 0 Page 227 of 258

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category: A. RCS or SG Tube Leakage Degradation Threat: Loss Threshold:

1. A leaking or RUPTURED SG is FAULTED outside of Containment Basis:

This threshold addresses a leaking or RUPTURED Steam Generator (SG) that is also FAULTED outside of containment. The condition of the SG, whether leaking or RUPTURED, is determined in accordance with the thresholds for RCS Barrier Potential Loss A.1 and Loss A.1, respectively. This condition represents a bypass of the containment barrier.

FAULTED is a defined term within the NEI 99-01 methodology; this determination is not necessarily dependent upon entry into, or diagnostic steps within, an EOP. For example, if the pressure in a steam generator is decreasing uncontrollably (part of the FAULTED definition) and the FAULTED steam generator isolation procedure is not entered because EOP user rules are dictating implementation of another procedure to address a higher priority condition; the steam generator is still considered FAULTED for emergency classification purposes.

The FAULTED criterion establishes an appropriate lower bound on the size of a steam release that may require an emergency classification. Steam releases of this size are readily observable with normal Control Room indications. The lower bound for this aspect of the containment barrier is analogous to the lower bound criteria specified in IC SU4 for the fuel clad barrier (i.e., RCS activity values) and IC SU5 for the RCS barrier (i.e., RCS leak rate values).

This threshold also applies to prolonged steam releases necessitated by operational considerations such as the forced steaming of a leaking or RUPTURED steam generator directly to atmosphere to cooldown the plant, or to drive an auxiliary (emergency) feed water pump. These types of conditions will result in a significant and sustained release of radioactive steam to the environment (and are thus similar to a FAULTED condition). The inability to isolate the steam flow without an adverse effect on plant cooldown meets the intent of a loss of containment.

Steam releases associated with the expected operation of a SG power operated relief valve or safety relief valve do not meet the intent of this threshold. Such releases may occur intermittently for a short period of time following a reactor trip as operators process through emergency operating procedures to bring the plant to a stable condition and prepare to initiate a plant cooldown. Steam releases associated with the unexpected operation of a valve (e.g., a stuck-open safety valve) do meet this threshold.

EAL BASES DOCUMENT Rev. 0 Page 228 of 258

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Following an SG tube leak or rupture, there may be minor radiological releases through a secondary-side system component (e.g., air ejectors, glad seal exhausters, valve packing, etc.). These types of releases do not constitute a loss or potential loss of containment but should be evaluated using the Recognition Category R ICs.

The ECLs resulting from primary-to-secondary leakage, with or without a steam release from the FAULTED SG, are summarized below.

Affected SG is FAULTED Outside of Containment?

P-to-S Leak Rate Yes No Less than or equal to 10 gpm No classification No classification Greater than 10 gpm for 15 Unusual Event per SU5 Unusual Event per SU5 minutes or longer Requires operation of a standby Site Area Emergency per charging (makeup) pump (RCS Alert per FA1 FS1 Barrier Potential Loss)

Requires an automatic or manual Site Area Emergency per ECCS (SI) actuation (RCS Barrier Alert per FA1 FS1 Loss)

There is no Potential Loss threshold associated with RCS or SG Tube Leakage.

Basis Reference(s):

1. DB-OP-02000 RPS, SFAS, SFRCS Trip, or SG Tubing Rupture

2. NEI 99-01 RCS or SG Tube Leakage Containment Loss 1.A EAL BASES DOCUMENT Rev. 0 Page 229 of 258

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category: A. RCS or SG Tube Leakage Degradation Threat: Potential Loss Threshold:

None EAL BASES DOCUMENT Rev. 0 Page 230 of 258

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category: B. Inadequate heat Removal Degradation Threat: Loss Threshold:

None EAL BASES DOCUMENT Rev. 0 Page 231 of 258

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category: B. Inadequate heat Removal Degradation Threat: Potential Loss Threshold:

1. Calculated Clad Temperature in Region 3 or higher (DB-OP-02000 Figure 2)

AND Restoration procedures not effective within 15 min. (Note 1)

Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Basis:

This condition represents an IMMINENT core melt sequence, which, if not corrected, could lead to vessel failure and an increased potential for containment failure. For this condition to occur there must already have been a loss of the RCS Barrier AND the Fuel Clad Barrier. If implementation of a procedure(s) to restore adequate core cooling is not effective (successful) within 15 minutes, it is assumed that the event trajectory will likely lead to core melting and a subsequent challenge of the Containment Barrier.

The restoration procedure is considered effective if core exit thermocouple readings are decreasing and/or if reactor vessel level is increasing. Whether or not the procedure(s) will be effective should be apparent within 15 minutes. The Emergency Director should escalate the emergency classification level as soon as it is determined that the procedure(s) will not be effective.

Severe accident analyses (e.g., NUREG-1150) have concluded that function restoration procedures can arrest core degradation in a significant fraction of core damage scenarios, and that the likelihood of containment failure is very small in these events. Given this, it is appropriate to provide 15 minutes beyond the required entry point to determine if procedural actions can reverse the core melt sequence.

Indication of severe core cooling degradation is manifested by Calculated Clad Temperature in Region 3 (DB-OP-02000 Figure 2). Figure 2, Incore T/C Temperature vs. RCS Pressure for ICC, provides indication of how serious core conditions are based upon combinations of RCS pressure AND incore thermocouple temperatures. If the RCS P-T point is in Region 3, the cladding temperatures in the core may be 1400°F or higher. This is a very serious condition and may lead to significant amounts of H2 production; core damage may be unavoidable as this represents a very serious inadequate core cooling condition (ref. 1, 2).

EAL BASES DOCUMENT Rev. 0 Page 232 of 258

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases The function restoration procedures are those emergency operating procedures that address the recovery of core cooling functions. The procedure is considered effective if the clad temperature is decreasing or if RCS water level is increasing (ref. 1, 2).

Basis Reference(s):

1. DB-OP-02000 RPS, SFAS, SFRCS Trip, or SG Tubing Rupture

2. Bases and Deviation Document for DB-OP-02000

3. NEI 99-01 Inadequate Heat Removal Containment Potential Loss 2.A EAL BASES DOCUMENT Rev. 0 Page 233 of 258

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category: C. CT Radiation/RCS Activity Degradation Threat: Loss Threshold:

None EAL BASES DOCUMENT Rev. 0 Page 234 of 258

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category: C. CT Radiation/RCS Activity Degradation Threat: Potential Loss Threshold:

1. RE 4596A or B > Table F-2 column CT Potential Loss (Note 9)

Table F-2 Containment Radiation - R/hr (RE 4596A or B)

Time After S/D RCS Loss FC Loss CT Potential Loss (Hrs.)

0-1 1.50E+01 3.03E+03 1.40E+04 1-2 1.50E+01 2.56 E+03 1.18 E+04 2-8 1.50E+01 1.61 E+03 7.46 E+03 8-16 1.50E+01 1.14 E+03 5.28 E+03 16-24 1.50E+01 8.66 E+02 4.00 E+03

>24 1.50E+01 3.94 E+02 1.82 E+03 Note 9: During a main steam line break in containment or LOCA with temperature >170F, there is a potential to induce transient errors into the output of RE4596A and B during the peak rate of temperature change.

Consult alternate indications. If the main steam line break is accompanied by core damage, this error is insignificant.

Basis:

The radiation monitor reading corresponds to an instantaneous release of ALL reactor coolant mass into the containment, assuming that 20% of the fuel cladding has failed. This level of fuel clad failure is well above that used to determine the analogous Fuel Clad Barrier Loss and RCS Barrier Loss thresholds.

NUREG-1228, Source Estimations During Incident Response to Severe Nuclear Power Plant Accidents, indicates the fuel clad failure must be greater than approximately 20% in order for there to be a major release of radioactivity requiring offsite protective actions. For this condition to exist there must already have been a loss of the RCS Barrier and the Fuel Clad Barrier. It is therefore prudent to treat this condition as a potential loss of containment, which would then escalate the ECL to a General Emergency.

The containment high range monitors, RE 4596A & B., monitor the gamma dose rate resulting from a postulated loss of coolant accident (LOCA). RE 4596A & B are located inside containment. The detector range is approximately 1 to 1E8 R/hr (logarithmic scale) (ref. 1).

EAL BASES DOCUMENT Rev. 0 Page 235 of 258

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases The Table F-2 values, column CT Potential Loss represents, based on Calculation EP-EALCALC-DB-0701, the expected containment high range radiation monitor (RE 4596A & B) response based on a LOCA, for periods of 1, 2, 8, 16, 24 and 48 (>24) hours after shutdown with ~20% fuel failure (ref. 2).

When evaluating fission product barrier integrity values in Table F-2, time after shutdown should be confirmed with the Control Room as the time that the reactor is tripped, and reactor power is verified to be lowering on the Intermediate Range (ref. 3). If time after shutdown is less than one hour (or the reactor is still critical), the 0-1 hour after shutdown value should be chosen. This is conservative as it represents sufficient time for plant conditions to deteriorate to the point that core damage may occur, and the activity released from the RCS into the containment atmosphere to reach equilibrium mixing throughout containment.

During a main steam line break in containment or LOCA with temperature > 170ºF, there is a potential to induce transient errors (positive and negative) into the output of RE4596A & B during the peak rate of temperature change. Consult alternate indications. If the main steam line break or LOCA is accompanied by core damage this error is, however, insignificant (ref. 4).

Basis Reference(s):

1. UFSAR Section 7.13.3.1 Containment High Radiation Monitors

2. EP-EALCALC-DB-0701 Containment Radiation Monitor Readings Following Clad Damage (FC2 and CT2 Potential Loss)

3. DB-OP-02000 RPS, SFAS, SFRCS Trip, or SG Tubing Rupture

4. NRC Information Notice 97-45 Supplement 1 Environmental Qualification Deficiency for Cables and Containment Penetration Pigtails

5. NEI 99-01 CMT Radiation / RCS Activity Containment Potential Loss 3.A EAL BASES DOCUMENT Rev. 0 Page 236 of 258

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category: D. CT Integrity or Bypass Degradation Threat: Loss Threshold:

1. Containment isolation is required AND EITHER:

Containment integrity has been lost based on Emergency Director judgment UNISOLABLE pathway from Containment to the environment exists Basis:

These thresholds address a situation where containment isolation is required and one of two conditions exists as discussed below. Users are reminded that there may be accident and release conditions that simultaneously meet both bulleted thresholds.

First Threshold - Containment integrity has been lost, i.e., the actual containment atmospheric leak rate likely exceeds that associated with allowable leakage (or sometimes referred to as design leakage). Following the release of RCS mass into containment, containment pressure will fluctuate based on a variety of factors; a loss of containment integrity condition may (or may not) be accompanied by a noticeable drop in containment pressure. Recognizing the inherent difficulties in determining a containment leak rate during accident conditions, it is expected that the Emergency Director will assess this threshold using judgment, and with due consideration given to current plant conditions, and available operational and radiological data (e.g., containment pressure, readings on radiation monitors outside containment, operating status of containment pressure control equipment, etc.).

Refer to the middle piping run of Figure 1. Two simplified examples are provided. One is leakage from a penetration and the other is leakage from an in-service system valve.

Depending upon radiation monitor locations and sensitivities, the leakage could be detected by any of the four monitors depicted in the figure.

Another example would be a loss or potential loss of the RCS barrier, and the simultaneous occurrence of two FAULTED locations on a steam generator where one fault is located inside containment (e.g., on a steam or feedwater line) and the other outside of containment. In this case, the associated steam line provides a pathway for the containment atmosphere to escape to an area outside the containment.

EAL BASES DOCUMENT Rev. 0 Page 237 of 258

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Following the leakage of RCS mass into containment and a rise in containment pressure, there may be minor radiological releases associated with allowable (design) containment leakage through various penetrations or system components. These releases do not constitute a loss or potential loss of containment but should be evaluated using the Recognition Category R ICs.

Second Threshold - Conditions are such that there is an UNISOLABLE pathway for the migration of radioactive material from the containment atmosphere to the environment. As used here, the term environment includes the atmosphere of a room or area, outside the containment, that may, in turn, communicate with the outside-the-plant atmosphere (e.g.,

through discharge of a ventilation system or atmospheric leakage). Depending upon a variety of factors, this condition may or may not be accompanied by a noticeable drop in containment pressure.

Refer to the top piping run of Figure 1. In this simplified example, the inboard and outboard isolation valves remained open after a containment isolation was required (i.e., containment isolation was not successful). There is now an UNISOLABLE pathway from the containment to the environment.

The existence of a filter is not considered in the threshold assessment. Filters do not remove fission product noble gases. In addition, a filter could become ineffective due to iodine and/or particulate loading beyond design limits (i.e., retention ability has been exceeded) or water saturation from steam/high humidity in the release stream.

Leakage between two interfacing liquid systems, by itself, does not meet this threshold.

Refer to the bottom piping run of Figure 1. In this simplified example, leakage in an RCP seal cooler is allowing radioactive material to enter the Auxiliary Building. The radioactivity would be detected by the Process Monitor. If there is no leakage from the closed water cooling system to the Auxiliary Building, then no threshold has been met. If the pump developed a leak that allowed steam/water to enter the Auxiliary Building, then second threshold would be met. Depending upon radiation monitor locations and sensitivities, this leakage could be detected by any of the four monitors depicted in the figure and cause the first threshold to be met as well.

Following the leakage of RCS mass into containment and a rise in containment pressure, there may be minor radiological releases associated with allowable containment leakage through various penetrations or system components. Minor releases may also occur if a containment isolation valve(s) fails to close but the containment atmosphere escapes to an enclosed system. These releases do not constitute a loss or potential loss of containment but should be evaluated using the Recognition Category R ICs.

The status of the containment barrier during an event involving steam generator tube leakage is assessed using Loss Threshold A.1.

Basis Reference(s):

1. NEI 99-01 CMT Integrity or Bypass Containment Loss 4.A EAL BASES DOCUMENT Rev. 0 Page 238 of 258

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category: D. CT Integrity or Bypass Degradation Threat: Loss Threshold:

2. Indications of RCS leakage outside of Containment Basis:

Containment sump level, temperature, pressure and/or radiation levels will increase if reactor coolant mass is leaking into the containment. If these parameters have not increased, then the reactor coolant mass may be leaking outside of containment (i.e., a containment bypass sequence). Increases in sump level, temperature, pressure, flow and/or radiation level readings outside of the containment may indicate that the RCS mass is being lost outside of containment.

Unexpected elevated readings and alarms on radiation monitors with detectors outside containment should be corroborated with other available indications to confirm that the source is a loss of RCS mass outside of containment. If the fuel clad barrier has not been lost, radiation monitor readings outside of containment may not increase significantly; however, other unexpected changes in sump levels, area temperatures or pressures, flow rates, etc.

should be sufficient to determine if RCS mass is being lost outside of the containment.

Refer to the middle piping run of Figure 1. In this simplified example, a leak has occurred at a reducer on a pipe carrying reactor coolant in the Auxiliary Building. Depending upon radiation monitor locations and sensitivities, the leakage could be detected by any of the four monitors depicted in the figure and cause threshold D.1 to be met as well.

To ensure proper escalation of the emergency classification, the RCS leakage outside of containment must be related to the mass loss that is causing the RCS Loss and/or Potential Loss threshold A.1 to be met.

Potential RCS leak pathways outside containment include (ref. 1, 2):

Decay Heat Removal ECCS (Safety Injection)

Makeup and Purification RC pump seals RCS sample lines RCS drain lines EAL BASES DOCUMENT Rev. 0 Page 239 of 258

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Basis Reference(s):

1. UFSAR Section 5.2.4.7 Leakage Identification

2. DB-OP-02522 Small RCS Leaks

3. NEI 99-01 CMT Integrity or Bypass Containment Loss EAL BASES DOCUMENT Rev. 0 Page 240 of 258

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Figure 1: Containment Integrity or Bypass Examples Inside Reactor Building RCP Seal Cooling IEAL BASES DOCUMENT Rev. O Page 241 of 2581

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category: D. CT Integrity or Bypass Degradation Threat: Potential Loss Threshold:

1. Containment pressure > 50.4 psia Basis:

If containment pressure exceeds the design pressure, there exists a potential to lose the Containment Barrier. To reach this level, there must be an inadequate core cooling condition for an extended period of time; therefore, the RCS and Fuel Clad barriers would already be lost. Thus, this threshold is a discriminator between a Site Area Emergency and General Emergency since there is now a potential to lose the third barrier.

50.4 psia (36 psig + elevation adjusted atmospheric pressure of 14.4 psia) is based on the containment design pressure (ref.1).

Basis Reference(s):

1. UFSAR Section 3.8.2.1 Containment Vessel

2. NEI 99-01 CMT Integrity or Bypass Containment Potential Loss 4.A EAL BASES DOCUMENT Rev. 0 Page 242 of 258

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category: D. CMT Integrity or Bypass Degradation Threat: Potential Loss Threshold:

2. Containment Hydrogen concentration > 4%

Basis:

The existence of an explosive mixture means, at a minimum, that the containment atmospheric hydrogen concentration is sufficient to support a hydrogen burn (i.e., at the lower deflagration limit). A hydrogen burn will raise containment pressure and could result in collateral equipment damage leading to a loss of containment integrity. It therefore represents a potential loss of the Containment Barrier.

Following a design basis accident, hydrogen gas may be generated inside the containment by reactions such as zirconium metal with water, corrosion of materials of construction and radiolysis of aqueous solution in the core and sump (ref. 1).

The Containment Hydrogen Monitoring System is used to monitor the hydrogen concentration inside containment after a severe accident involving core damage. The containment hydrogen monitors (AI 5027 & AI 5028) are not required to be operated in the continuous mode, however, the system is required to be started up 30 minutes after containment sprays have been initiated (ref. 2).

The lower limit for the occurrence of an in-containment hydrogen burn is approximately 4%

(ref. 1).

To generate such levels of combustible gas, loss of the Fuel Clad and RCS barriers must have occurred. With the Potential Loss of the containment barrier, the threshold hydrogen concentration, therefore, will likely warrant declaration of a General Emergency.

Basis Reference(s):

1. DBSAMG-TBD Davis-Besse Severe Accident Management Guidelines Technical Bases Document

2. UFSAR Section 7.13.3.4 Containment Hydrogen Monitors

3. NEI 99-01 CMT Integrity or Bypass Containment Potential Loss 4.B EAL BASES DOCUMENT Rev. 0 Page 243 of 258

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category: D. CMT Integrity or Bypass Degradation Threat: Potential Loss Threshold:

3. Containment pressure > 40 psia with < one full train of containment cooling, Table F-3, operating per design for 15 min. (Note 1)

Table F-3 Containment Cooling Full Train CT Spray Pumps CT Cooling Fans 2 0 1 1 0 2 Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Basis:

This threshold describes a condition where containment pressure is greater than the setpoint at which containment energy (heat) removal systems are designed to automatically actuate, and less than one full train of equipment is capable of operating per design. The 15-minute criterion is included to allow operators time to manually start equipment that may not have automatically started, if possible. This threshold represents a potential loss of containment in that containment heat removal/depressurization systems (e.g., containment sprays, ice condenser fans, etc., but not including containment venting strategies) are either lost or performing in a degraded manner.

The combination of Containment spray pumps and Containment cooling fan units considered to be a full train of containment cooling operating per design is shown in Table F-3 (ref. 1).

SFAS 2 actuation automatically initiates Containment Air Coolers upon exceeding the Containment pressure high setpoint of 18.7 psia or low RCS pressure of 1600 psig. SFAS Level 4 actuation automatically initiates Containment Spray upon exceeding the Containment pressure high-high setpoint of 40 (nominal) psia (ref. 2, 3).

EAL BASES DOCUMENT Rev. 0 Page 244 of 258

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Basis Reference(s):

1. UFSAR Section 6.2.2. Containment Vessel Heat Removal Systems

2. DB-OP-02000 RPS, SFAS, SFRCS Trip, or SG Tubing Rupture

3. DBNPS Technical Specifications Table 3.3.5-1 Safety Features Actuation System Instrumentation

4. NEI 99-01 CMT Integrity or Bypass Containment Potential Loss 4.C EAL BASES DOCUMENT Rev. 0 Page 245 of 258

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category: F. Emergency Director Judgment Degradation Threat: Loss Threshold:

1. ANY condition in the opinion of the Emergency Director that indicates Loss of the Containment Barrier Basis:

This threshold addresses any other factors that may be used by the Emergency Director in determining whether the Containment Barrier is lost.

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 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.

Basis Reference(s):

1. NEI 99-01 Emergency Director Judgment PC Loss 6.A EAL BASES DOCUMENT Rev. 0 Page 246 of 258

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category: F. Emergency Director Judgment Degradation Threat: Potential Loss Threshold:

1. ANY condition in the opinion of the Emergency Director that indicates Potential Loss of the Containment Barrier Basis:

This threshold addresses any other factors that may be used by the Emergency Director in determining whether the Containment Barrier is lost.

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.

Basis Reference(s):

1. NEI 99-01 Emergency Director Judgment PC Potential Loss 6.A EAL BASES DOCUMENT Rev. 0 Page 247 of 258

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

Background

NEI 99-01 Revision 6 ICs AA3 and HA5 prescribe declaration of an Alert based on IMPEDED access to rooms or areas (due to either area radiation levels or hazardous gas concentrations) where equipment necessary for normal plant operations, cooldown or shutdown is located.

These areas are intended to be plant operating mode dependent. Specifically the Developers Notes for AA3 and HA5 states:

The site-specific list of plant rooms or areas with entry-related mode applicability identified should specify those rooms or areas that contain equipment which require a manual/local action as specified in operating procedures used for normal plant operation, cooldown and shutdown. Do not include rooms or areas in which actions of a contingent or emergency nature would be performed (e.g., an action to address an off-normal or emergency condition such as emergency repairs, corrective measures or emergency operations). In addition, the list should specify the plant mode(s) during which entry would be required for each room or area.

The list should not include rooms or areas for which entry is required solely to perform actions of an administrative or record keeping nature (e.g., normal rounds or routine inspections).

Further, as specified in IC HA5:

The list need not include the Control Room if adequate engineered safety/design features are in place to preclude a Control Room evacuation due to the release of a hazardous gas.

Such features may include, but are not limited to, capability to draw air from multiple air intakes at different and separate locations, inner and outer atmospheric boundaries, or the capability to acquire and maintain positive pressure within the Control Room envelope.

EAL BASES DOCUMENT Rev. 0 Page 248 of 258

ATTACHMENT 3 Safe Shutdown Rooms/Areas Tables R-2 & H-2 Bases DBNPS Table R-2 and H-2 Bases NEI 99-01 Rev 06 addresses elevated radiation levels in certain plant rooms/areas sufficient to preclude or IMPEDE personnel from performing actions necessary to maintain normal plant operation, or to perform a normal plant cooldown and shutdown.

DB-OP-06902, Power Operations Rev 46 was reviewed to determine what actions are necessary to maintain power operations. It was determined that over reasonable periods of time (days vice years) there are no actions outside the Control Room that are required to be performed to maintain normal operations. Eventually, a shutdown would be required if Technical Specification surveillance testing was not completed and you complied with the associated LCOs or based on consumable supplies being depleted. For the purpose of this table, no actions were determined to be required.

The following table lists the locations that an operator may be dispatched in order perform a normal plant cooldown and shutdown. The review was completed using the following procedures as the controlling documents:

DB-OP-06902, Power Operations R46 DB-OP-06903, Plant Cooldown R47 DB-OP-02504, Rapid Shutdown R20 In addition, DB-OP-06012 was reviewed to ensure the Decay Heat Removal System is aligned.

At Davis-Besse, RCS Cooldown starts once both Steam Generators reach Low Level Limits during a power reduction (approximately 30% power). As a result, this review started with DB-OP-06902, Power Operations, Section 8, Turbine and Reactor Shutdown and then transitioned to DB-OP-06903, Plant Cooldown. Each step in the controlling procedures was evaluated to determine if the action was performed in the Control Room or in the plant. In-plant actions were evaluated and a determination was made whether or not the actions, if not performed, would prevent achieving cold shutdown. The following generic assumptions were applied:

Steps involving optional degasing of the RCS were not included since degasing the RCS is not required to reach cold shutdown.

Steps involving supplying Auxiliary Steam were not included since AFW and AVVs can be used to reach cold shutdown if Condenser vacuum is lost.

Steps involving Main Feedwater Pumps were not included since AFW and AVVs can be used to reach cold shutdown if Main Feedwater is not available.

Travel paths to the locations where the equipment is operated are not part of the determination, only the rooms where the equipment is actually operated are considered as the affected room. Travel paths were not included because most locations can be reached via alternate travel paths if required due to a localized issue.

EAL BASES DOCUMENT Rev. 0 Page 249 of 258

ATTACHMENT 3 Safe Shutdown Rooms/Areas Tables R-2 & H-2 Bases No assumptions made about which LPI Train is aligned for DHR Operation. Locations could be reduced by preselecting one train (typically Train 2) to provide this function. It is assumed that both trains are in a Standby LPI mode at the start of the event.

The minimum set of in-plant actions, associated locations, and operating modes to shut down and cool down the reactor are highlighted. The locations where those actions are performed comprise the rooms/areas in EAL Tables R-2 and H-2.

The control room was not included in Table H-2 evaluation because the control room is governed by H6 series for Control Room Evacuation.

UFSAR Section 6.4.2 Toxic Gas Protection Provisions states that no toxic or explosive materials are stored in volumes or locations, which pose a control room habitability hazard that exceeds emergency system capabilities.

EAL RA3.1 addresses control room habitability relative to area radiation levels.

EAL BASES DOCUMENT Rev. 0 Page 250 of 258

ATTACHMENT 3 Safe Shutdown Rooms/Areas Tables R-2 & H-2 Bases In Plant Task - If action not performed, does this Step Action Building Elevation Room Mode Procedure prevents cooldown shutdown?

and Step 06902 Align CWRT to RCS No - Inventory and Boration requirements 8.2.2 Makeup. can be met from Control Room using N/A N/A N/A N/A BWST and BAATs.

06902 Turbine Overspeed trip No - This testing is not required. Testing 8.2.3 testing if required. could be performed on subsequent restart. N/A N/A N/A N/A 06902 Set condenser pressure to No - This action is not required to perform 8.3.9 Bullet 1.5 to 2.5 inches HgA using Shutdown - Cooldown. N/A N/A N/A N/A PCV1061.

06902 Attachment 8, Drain and No - This only affects efficiency or 8.3.9 Bullet Steam Trap Alignment for improves moisture removal. N/A N/A N/A N/A Shutdown 06902 Fill and Vent the SG No - This action facilitates SG Fill, Soak, 8.3.9 Bullet Blowdown lines, and Drains, but is not required to complete N/A N/A N/A N/A Shutdown - Cooldown.

06902 Place both Instrument Air No - Step improves response of air 8.3.9 Bullet dryers in service in parallel, system, but is not required. N/A N/A N/A N/A 06902 MDFP is in Standby in the No - MDFP could be used in AFW Mode 8.3.9 Bullet Main Feedwater mode and to reach Cold Shutdown. N/A N/A N/A N/A warm up is complete.

06902 Place the Auxiliary Boiler in No - Aux Steam will continue to be 8.3.9 Bullet service. supplied from Main Steam. If vacuum and N/A N/A N/A N/A therefore condenser is lost, steam can be dumped via AVVs.

6902 8.3.9 Transfer Auxiliary Steam No - Aux Steam will continue to be Bullet Loads from the Main Steam supplied from Main Steam. If vacuum and N/A N/A N/A N/A Reducing Station to the therefore condenser is lost, steam can be Auxiliary Boiler. dumped via AVVs.

EAL BASES DOCUMENT Rev. 0 Page 251 of 258

ATTACHMENT 3 Safe Shutdown Rooms/Areas Tables R-2 & H-2 Bases In Plant Task - If action not performed, does this Step Action Building Elevation Room Mode Procedure prevents cooldown shutdown?

and Step 6902 8.3.10 Perform Attachment 10, No - MDFP could be used in AFW Mode Bullet MDFP Operation. to reach Cold Shutdown. N/A N/A N/A N/A 6902 8.3.10 Closed the actuator cylinder No - This action facilitates SG Fill, Soak, Bullet equalizing valve for and Drains, but is not required to complete N/A N/A N/A N/A MS4531/MS4532 Shutdown - Cooldown.

6902 8.3.10 Place the SG Blowdown No - This action facilitates SG Fill, Soak, Bullet Lines in service. and Drains, but is not required to complete N/A N/A N/A N/A Shutdown - Cooldown.

6902 8.3.16 IF Main Turbine Overspeed No - This testing is not required. Testing Testing is required, could be performed on subsequent restart. N/A N/A N/A N/A 6902 8.3.17 Perform Attachment 13, No - This action completes Turbine Turbine Shutdown. Shutdown, but is not required to meet N/A N/A N/A N/A Cold Shutdown.

6902 8.3.30 Open the air isolation valve No - Since performance of Attachment 10 that was closed in is not required, neither is restoration.

N/A N/A N/A N/A Attachment 10, MDFP Operation:

6902 8.3.31 Complete the shutdown of No - desired to complete MFPT the running MFPT to shutdown, but will not prevent reaching N/A N/A N/A N/A Turning Gear operation. cold shutdown.

6902 8.3.32 Perform Attachment 8, Drain No - This only affects efficiency or and Steam Trap Alignment improves moisture removal. N/A N/A N/A N/A for Shutdown, step 3.0.

6902 8.3.38 Transition to DB-OP-06903, No - This is a procedure routing step.

Plant Cooldown N/A N/A N/A N/A 6903 3.3.10 Perform the following to No - Action not required to reach cold read Reactor Vessel Head shutdown. N/A N/A N/A N/A O-Ring pressure.

EAL BASES DOCUMENT Rev. 0 Page 252 of 258

ATTACHMENT 3 Safe Shutdown Rooms/Areas Tables R-2 & H-2 Bases In Plant Task - If action not performed, does this Step Action Building Elevation Room Mode Procedure prevents cooldown shutdown?

and Step 6903 3.12.1 Verify the requirements No - Action not required to reach cold pertaining to containment shutdown.

closure control and N/A N/A N/A N/A protected equipment are initiated.

6903 3.12.2 Initiate Attachment 13, No - Action not required to reach cold Preparation of MU Filter 1 shutdown.

N/A N/A N/A N/A for Hydrogen Peroxide Addition.

6903 3.13 Direct I & C Department to No - Action not required to reach cold perform Attachment 8, shutdown.

Radiation Monitor N/A N/A N/A N/A Preparations for Plant Shutdown 6903 3.16 IF the RCS will be opened to No - Opening the RCS is not required to atmosphere, THEN begin to reach cold shutdown.

reduce the RCS H2 N/A N/A N/A N/A concentration to less than 15 cc/kg.

6903 3.17 IF the Circulating Water No - Draining Circ Water is not required to System is to be drained, reach cold shutdown.

THEN notify Chemistry to N/A N/A N/A N/A chemically shock it prior to bypassing the Cooling Tower.

6903 3.18 Remove TPCW pumps and No - Controlling temperature and heat exchangers from Shutdown of TPCW is not required to service, REFER TO DB-OP- reach cold shutdown. N/A N/A N/A N/A 06263, Turbine Plant Cooling Water 6903 3.21 Verify one of the Clean No - The BWST and the BAATs can be Waste Receiver Tanks is used from the Control Room for RCS N/A N/A N/A N/A aligned inventory and Boration.

6903 3.25 Begin making preparations No - Placing CTMT Purge in service is not to Start-up Containment required to reach cold shutdown. N/A N/A N/A N/A Vessel Purge.

EAL BASES DOCUMENT Rev. 0 Page 253 of 258

ATTACHMENT 3 Safe Shutdown Rooms/Areas Tables R-2 & H-2 Bases In Plant Task - If action not performed, does this Step Action Building Elevation Room Mode Procedure prevents cooldown shutdown?

and Step 6903 Place/remove a second seal No - Controlling a second seal return 3.25.2.a/d.3 return cooler in/from service. cooler not required to reach cold N/A N/A N/A N/A shutdown.

6903 3.35.2 Begin performing No - This action facilitates SG Fill, Soak, Attachment 5 SG Fill, Soak, and Drains, but is not required to reach N/A N/A N/A N/A and Drain. cold shutdown.

6903 3.47 Shutdown one RCP. No - The controlling procedure provides direction when CTMT is not accessible. N/A N/A N/A N/A That direction would be used.

6903 3.53 Isolate Core Flood Tank 1 Yes - Action is required to reduce RCS perform the following: Pressure. CFT 1 Outlet Valve power Rm 304 - Corridor Aux 585 1, 2, 3 restored at E11B to allow closure. Outside #3 MPR 6903 3.53 Isolate Core Flood Tank 2 Yes - Action is required to reduce RCS perform the following: Pressure. CFT 2 Outlet Valve power Rm 427 - #2 Electrical Aux 603 1, 2, 3 restored at F11A to allow closure. Penetration Room 06903 3.63 WHEN RCS pressure is No - The controlling procedure provides between 450 psig and 425 direction when CTMT is not accessible.

N/A N/A N/A N/A psig THEN establish a 0/2 or That direction would be used.

2/0 RCP combination.

6903 3.64 Control Letdown Flow - No - This action aids in RCS cleanup, Open or throttle MU 83 which is not required to reach cold N/A N/A N/A N/A shutdown.

6903 3.64.1 IF its necessary to reduce No - This action allow lower Makeup the total makeup flow to the Flow, which would only be required in a RCS by throttling MU 58A, very slow cooldown. This action is not N/A N/A N/A N/A NORMAL MAKEUP FLOW required to reach cold shutdown.

FE MU58 SOURCE/NEEDLE, 6903 3.66 WHEN Feedwater flow No - This action would reduce FW flow via reduces to the point where the MFW header, which would only be level control using the required in a very slow cooldown. This N/A N/A N/A N/A Feedwater Startup Valves is action is not required to reach cold difficult, Throttle FW 139 shutdown.

EAL BASES DOCUMENT Rev. 0 Page 254 of 258

ATTACHMENT 3 Safe Shutdown Rooms/Areas Tables R-2 & H-2 Bases In Plant Task - If action not performed, does this Step Action Building Elevation Room Mode Procedure prevents cooldown shutdown?

and Step and/or FW44.

6903 3.68 Start actions to reduce No - This action will allow cooldown to feedwater temperature to proceed quicker if on Main Feedwater, but N/A N/A N/A N/A 200F. is not required to reach cold shutdown.

6903 4.17.1 To prevent a transfer of Yes - Action is required to align LPI Train Room 236 water from the RCS to the 1 for DHR Operations. Aux 565 #2 Mechanical 1, 2, 3 BWST, close DH10. Penetration Room 6903 4.17.1 To prevent a transfer of Yes - Action is required to align LPI Train Room 236 water from the RCS to the 2 for DHR Operations. Aux 565 #2 Mechanical 1, 2, 3 BWST, close DH26. Penetration Room 6903 4.17.2 Close breaker BF 1130 in Yes - Action is required to align either LPI F11A, for DH 11 Train 1 or 2 for DHR Operations. Rm 427 - #2 Electrical Aux 603 1, 2, 3 Penetration Room 6903 4.17.2 Close breaker BE 1183 in Yes - Action is required to align either LPI E11B, for DH 12. Train 1 or 2 for DHR Operations. Rm 304 - Corridor Aux 585 1, 2, 3 Outside #3 MPR 6903 If DH12 does not open Yes - Action is required to align either LPI 4.17.8.d (which is generally does Train 1 or 2 for DHR Operations. Rm 304 - Corridor Aux 585 1, 2, 3 not), Install jumper. Outside #3 MPR 6903 4.20 Begin reducing Deaerator No - This action will allow cooldown to and Feedwater temperature proceed quicker if on Main Feedwater, but N/A N/A N/A N/A to less than 120F. is not required to reach cold shutdown.

6903 4.21 WHEN RCS temperature is No - This action will prevent low temp-less than 280ºF, THEN high pressure conditions in RCS, but is not N/A N/A N/A N/A disable HPI by racking out required to reach cold shutdown.

breakers.

6903 4.22.3 IF MSIVs are to be stroked No - Action is not required to reach cold or pinned, THEN stroke test shutdown. N/A N/A N/A N/A MS100 and MS101.

EAL BASES DOCUMENT Rev. 0 Page 255 of 258

ATTACHMENT 3 Safe Shutdown Rooms/Areas Tables R-2 & H-2 Bases In Plant Task - If action not performed, does this Step Action Building Elevation Room Mode Procedure prevents cooldown shutdown?

and Step 6903 4.25 Place a LPI Train in service Yes - This action is performed per DB-as a Decay Heat Removal OP-06012. However, since numerous N/A N/A N/A N/A Train, actions are required each action in that procedure was assessed separately.

6903 4.28 Remove the Auxiliary No - Action is not required to reach cold Feedwater System from shutdown. N/A N/A N/A N/A service.

6903 4.30 IF SG 1 level control No - Action is not required to reach cold becomes difficult due to shutdown. N/A N/A N/A N/A excess flow through SP7B, adjust FW161/139.

6903 4.32 IF SG 2 level control No - Action is not required to reach cold becomes difficult due to shutdown. N/A N/A N/A N/A excess flow through SP7A, adjust FW162/44.

6903 4.34 WHEN RCS temperature is No - Cold shutdown has been reached.

less than 200ºF, End of review using DB-OP-06903. N/A N/A N/A N/A 6012 3.5.5 Verify the CLOSE power No - Just prevents inadvertent start of LPI fuses for AC 112, DECAY pump 1 while transferring suction. SFAS N/A N/A N/A N/A HT PUMP 1-1, are removed. is already blocked.

6012 3.5.6 Verify BE 1187 (E11E), MV No - Just prevents possible valve motor DH64 LPI-HPI CROSS overload if stroked with RCS Suction N/A N/A N/A N/A CONN ISO VLV 1, is open. Source.

6012 3.5.10 Open BE 1121 (E11A), MV No - Just prevents inadvertent transfer of 2733 DH PUMP 1 SUCT BWST inventory to RCS if the valve N/A N/A N/A N/A VLV FRM BWST. opened. SFAS is already blocked.

6012 3.5.13 Close DH10*, DH PUMP 1 Yes - This action is required to align LPI Room 236 MINIMUM COOLDOWN Train 1 or 2 for DHR Operations. Aux 565 #2 Mechanical 1, 2, 3 ISOLATION. Penetration Room 6012 3.5.17 Open BE 1126 (E11D), MV No - Just prevent inadvertent loss of DHR 1517 DH NORM SUCT Train 1 suction from RCS is valve is N/A N/A N/A N/A LINE 1 ISO VLV. stroked closed.

EAL BASES DOCUMENT Rev. 0 Page 256 of 258

ATTACHMENT 3 Safe Shutdown Rooms/Areas Tables R-2 & H-2 Bases In Plant Task - If action not performed, does this Step Action Building Elevation Room Mode Procedure prevents cooldown shutdown?

and Step 6012 3.6.5 Verify the CLOSE power No - Just prevents inadvertent start of LPI fuses for AD 112, DECAY pump 2 while transferring suction. SFAS N/A N/A N/A N/A HT PUMP 1-2, are removed. is already blocked.

6012 3.6.6 Verify BE 1195 (F11E), MV No - Just prevents possible valve motor DH63 LPI-HPI CROSS overload if stroked with RCS Suction N/A N/A N/A N/A CONN ISO VLV 2, is open. Source.

6012 3.6.11 Open BF 1134 (F11C), MV No - Just prevents inadvertent transfer of 2734 DH PMP 2 SUCT VLV BWST inventory to RCS if the valve N/A N/A N/A N/A FRM BWST. opened. SFAS is already blocked.

6012 3.6.14 Close DH10*, DH PUMP 1 Yes - Action is required to align LPI Train Room 236 MINIMUM COOLDOWN 1 or 2 for DHR Operations. Aux 565 #2 Mechanical 1, 2, 3 ISOLATION. Penetration Room 6012 3.6.19 Open BF1129 (F11C), MV No - Prevents inadvertent loss of DHR 1518 DH NORM SUCT Train 2 suction from RCS is valve is N/A N/A N/A N/A LINE 2 ISO VLV. stroked closed.

6012 3.6.27 Verify the CLOSE power No - If not initially removed, then fuses for AD 112. reinstalling will not be required. N/A N/A N/A N/A 6012 3.7.8.a Station an operator at DH No - Operator only stationed for Pump 1. monitoring function. Action is not required N/A N/A N/A N/A to reach Cold Shutdown.

6012 3.7.8.d Verify DH59, DH PUMP 1 No - Opening this valve provides the DISCHARGE SAMPLE capability to sample RCS inventory from N/A N/A N/A N/A ISOL, is open. the DHR system.

6012 3.8.8.a Station an operator at DH No - Operator only stationed for Pump 2. monitoring function. Action is not required N/A N/A N/A N/A to reach Cold Shutdown.

6012 3.8.8.d Verify DH60, DH PUMP 2 No - Opening this valve provides the DISCHARGE SAMPLE capability to sample RCS inventory from N/A N/A N/A N/A ISOL, is open. the DHR system.

Note: The information in the above table is included for historical reference information only and based upon the procedure revision numbers referenced in the DBNPS EAL BASES DOCUMENT Rev. 0 Page 257 of 258

ATTACHMENT 3 Safe Shutdown Rooms/Areas Tables R-2 & H-2 Bases Table R-2 and H-2 Bases summary.

Table R-2 & H-2 Results Table R-2 & H-2 Safe Shutdown Rooms/Areas Room/Area Mode Applicability Aux Bldg. 565 ele. Room 236 #2 Mechanical Penetration Room 1, 2, 3 Aux Bldg. 585 ele. Room 304 corridor outside #3 Mechanical Penetration Room 1, 2, 3 Aux Bldg. 603 ele. Room 427 - #2 Electrical Penetration Room 1, 2, 3 EAL BASES DOCUMENT Rev. 0 Page 258 of 258