L-12-142, Supplemental Information to a Request for Licensing Action on a Proposed Emergency Preparedness Plan Revision

Supplemental Information to a Request for Licensing Action on a Proposed Emergency Preparedness Plan Revision

ML121840082
Person / Time
Site:	Beaver Valley
Issue date:	06/29/2012
From:	Harden P FirstEnergy Nuclear Operating Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
L-12-142, TAC ME7823, TAC ME7824
Download: ML121840082 (64)

Text

FENOC Beaver Valley Power Station NP.O.

Box 4 FirstEnergy Nuclear Operating Company Shippingport, PA 15077 Paul A. Harden 724-682-5234 Site Vice President Fax: 724-643-8069 June 29, 2012 L-12-142 10 CFR 50, Appendix E ATTN: Document Control Desk U. S. Nuclear Regulatory Commission Washington, D.C. 20555-0001

SUBJECT:

Beaver Valley Power Station, Unit Nos. 1 and 2 Docket No. 50-334, License No. DPR-66 Docket No. 50-412, License No. NPF-73 Supplemental Information to a Request for Licensing Action on a Proposed Emergency Preparedness Plan Revision (TAC Nos. ME7823 and ME7824)

FirstEnergy Nuclear Operating Company (FENOC), by letter dated December 21, 2011, (Accession Number ML11362A317) requested Nuclear Regulatory Commission (NRC) review of a proposed revision of the Beaver Valley Power Station, Unit Nos. 1 and 2 Emergency Preparedness Plan (EPP). The proposed revision converted the EPP from the current EAL scheme to the Nuclear Energy Institute (NEI) 99-01, "Methodology for Development of Emergency Action Levels," Revision 5 scheme, as clarified by a series of NRC-accepted frequently asked questions (FAQs) on the NEI methodology.

Minor administrative issues were identified in Appendix 2, "Proposed Beaver Valley Power Station, Unit Nos. 1 and 2 Emergency Preparedness Plan Revision to Section 4, 'Emergency Conditions';" Appendix 4, "Beaver Valley Power Station Unit No. 1 EAL Evaluation;" Appendix 5, "Beaver Valley Power Station Unit No. 2 EAL Evaluation;"

and Appendix 6, "Proposed Beaver Valley Power Station, Unit Nos. 1 and 2 EAL Wallboards."

In Appendix 2, changes were made to pages 4-i, 4-9, 4-10, 4-12 through 4-17, 4-19 through 4-22, 4-24, 4-25, 4-31, 4-39, 4-44, 4-85, 4-95, 4-112, 4-123, 4-126, 4-157, 4-158, 4-173 through 4-175, 4-177 through 4-179, 4-181 through 4-184, 4-187, 4-193, 4-201, 4-206, 4-248, 4-258, 4-275, 4-285, and 4-288. On page 4-i, the page number reflecting the start of the Beaver Valley Power Station Unit No. 2 section was incorrect.

On pages 4-9 and 4-173 the numerical values on the x-axis of the graphs did not align with the large hash marks on the x-axis of the graphs. The title for the x-axis of the graphs on pages 4-9, 4-31, 4-44, 4-173, 4-193, and 4-206 were either incomplete or

Beaver Valley Power Station, Unit Nos. 1 and 2 L-12-142 Page 2 inconsistent with each other. On pages 4-9, 4-12, 4-14 through 4-17, 4-19, 4-22, 4-173 through 4-175, 4-177 through 4-179, 4-181, and 4-184, a note at the bottom of the page was deleted because it was not referenced in the body of the page. On pages 4-10, 4-39, 4-174, and 4-201, for Subcategory RC5, the engineering units for several temperatures were added. On page 4-13, for Initiating Conditions (IC) RAI and RU1, unneeded asterisks were deleted. On pages 4-24, 4-157, and 4-158, under IC CU8, an instrument designator was corrected. On pages 4-25 and 4-187, under IC CA10, the grammar in Emergency Action Level (EAL) Number 1 was corrected. On page 4-95, the word "EXPLOSION" was added into the second paragraph. On pages 4-85 and 4-248, the safety system and component verbiage, in the second paragraph, was changed from singular to plural. On pages 4-95 and 4-258, the safety system and component verbiage, in the third paragraph, was changed from singular to plural.

On pages 4-20, 4-21, 4-112, 4-123, 4-126, 4-182, 4-183, 4-275, 4-285, and 4-288, the referenced Emergency Director note was modified to better reflect a similar note contained in NEI 99-01, Revision 5.

In Appendix 4, changes were made on pages 4, 19, 24, 47, and 48. On page 4, Bullet 4 was made consistent with changes made to the note at the bottom of the pages in Appendix 2. Page 19 was made consistent with the change in RC5 in Appendix 2.

Page 24 had a duplicate phase deleted. Page 47 was made consistent with the change in CU8 in Appendix 2. Page 48 was made consistent with the change in CA10 in Appendix 2.

In Appendix 5, changes were made on pages 4, 19, and 47. On page 4, Bullet 4 was made consistent with changes made to the note at the bottom of the pages in Appendix 2. Page 19 was made consistent with the change in RC5 in Appendix 2.

Page 47 was made consistent with the change in CA1 0 in Appendix 2.

In Appendix 6, the wallboards were revised to reflect the changes made to Appendix 2 with respect to the changes to the graphs and the EALs. Additionally, an Emergency Director note was made consistent with the wording in Appendix 2.

Replacement pages with these changes incorporated are attached for inclusion into the submittal dated December 21, 2011.

There are no regulatory commitments contained in this letter. If there are any questions or if additional information is required, please contact Mr. Phil H. Lashley, Supervisor -

Fleet Licensing, at (330) 315-6808.

Sincer Paul A. Harden Y,,/A 4

lk't'LO&O

Beaver Valley Power Station, Unit Nos. 1 and 2 L-12-142 Page 3

Attachment:

Replacement Pages for Appendices 2, 4, 5, and 6 of the Proposed Beaver Valley Power Station, Unit Nos. 1 and 2 Emergency Preparedness Plan Using Nuclear Energy Institute 99-01 Revision 5 Methodology cc:

NRC Region I Administrator NRC Resident Inspector NRC Project Manager Director BRP/DEP Site BRP/DEP Representative

Attachment L-12-142 Replacement Pages for Appendices 2, 4, 5, and 6 of the Proposed Beaver Valley Power Station, Unit Nos. 1 and 2 Emergency Preparedness Plan Using Nuclear Energy Institute 99-01 Revision 5 Methodology (Sixty pages follow)

Appendix 2 Proposed Beaver Valley Power Station, Unit Nos. 1 and 2 Emergency Preparedness Plan Revision to Section 4, "Emergency Conditions"

Section 4 EMERGENCY CONDITIONS Table of Contents Page No.

4.1 BVPS Unit No. 1 (BVPS-1) EMERGENCY ACTION LEVELS and Basis 4-1 4.2 BVPS Unit No. 2 (BVPS-2) EMERGENCY ACTION LEVELS and Basis 4-165 4-i Rev. Proposed

PlAARALI *A*

IA* A A *I*R

• A*

A

• A *mALI ISIIUN PKRUuCT I EIRKiK UDIEGRAUAI OUN Modes: 1 - Power Operation, 2 - Startup, 3 - Hot Standoy, 4 - Hot Shutoown, 5 - Cold Snutdown, 6 - Refueli*, U - ueOuelea

1. Loss of any two barers and loss or potential loss of the

1. Loss or potential loss of any two baniers.

1. Any loss or any potential loss of either fuel clad or RCS.

1. Any loss or any potential loss of containment.

third barrier.

I Graph F-i: Radiation Reading Barrier Thresholds 1.0E+04 1.0 E + 0 3 I-0 C4 1.OE+02 0a o 1.OE+O1

()

0----------

-- FC2 RC2 1.OE+00 4--

0 2

4 6

8 10 12 14 16 18 20 22 24 26 28 30 32 34 36 38 40 42 44 46 48 Post LOCA Time (Hours after shutdown) 4-9 Rev. Proposed

Section 4 EMERGENCY ACTION LEVEL Bases Emergency Preparedness Plan FISSION PRODUCT BARRIER DEGRADATION Modes: 1 - Power Operation, 2 - Startup, 3 - Hot Standby. 4 - Hot Shutdown, 5 - Cold Shutdown, 6 - Refueling, D - Defueled FG1 I

U]JIr FS1 Nr*jrj FA1 rMr]J r FUl rLNL[4]

1.

Loss of any two barriers and loss or potential loss of the

1.

Loss or potential loss of any two barriers.

1.

Any loss or any potential loss of either fuel clad or RCS.

1.

Any loss or any potential loss of containment.

third barrier.

FC - Fuel Clad RC - Reactor Coolant System CT - Containment Sub-Category Loss Potential Loss Loss Potential Loss Loss Potential Loss

1.

Critical Safety

1. Core Cooling - Red entry

1. Core Cooling - Orange entry

1. RCS Integrity-Red entry

1.

Containment - Red entry Function Status conditions met.

conditions met.

conditions met.

conditions met.

OR OR

2. a.

Heat Sink - Red entry

2.

a.

Heat Sink - Red entry conditions met.

conditions met.

AND AND

b.

Heat Sink is required.

b.

Heat sink is required.

2.. Containment

1. Containment Radiation Monitor

1. Containment Radiation Monitor

1.

Containment Radiation Monitor Rad Monitoring (RM-1RM-219A or B)

(RM-1RM-219A or B) > 8 R/hr (RM-1RM-219A or B)

> FC2 Line on Graph F-i.

(RC2 Line on Graph F-l).

> CT2 Line on Graph F-1

3. Core

1.

Five hottest core exit

1.

Five hottest core exit

1.

a. Five hottest core exit Temperature thermocouples > 12000 F.

thermocouples > 7190 F.

thermocouples > 20000 F.

AND

b.

Restoration procedures not effective within 15 minutes.

OR

2.

a.

Five hottest core exit thermocouples > 12000 F.

AND

b.

RVLIS Full Range < 40%

with no RCPs running.

AND

c.

Restoration procedures not effective within 15 minutes

4.

RCS Level

1.

RCS level < Table F-1.

5.

RCS Leak Rate

1.

RCS leak rate greater than

1. UNISOLABLE RCS leak Table F-I: RVLIS Thresholds available makeup capacity as exceeding the capacity of one RVLIS RCPs Indication indicated by RCS subcooling charging pump (129 gpm) in the Full Range 0

40%

< 180 F normal containment or normal charging mode.

1 25%

< 330 F adverse containment.

6.

SG Tube Dynamic Range 2

33%

1.

RUPTURED SG results in an SI Note:

Leakage I 06%

actuation.

A prolonged release is greater than 4 Rupture hours.

1.

RUPTURED SG is also FAULTED outside of containment.

OR

2.

a.

Primary-to-Secondary leak rate > 10 gpm.

AND 4-10 Rev. Proposed

Section 4 EMERGENCY ACTION LEVEL Bases Emergency Preparedness Plan FISSION PRODUCT BARRIER DEGRADATION.

Modes: 1 - Power Operation. 2.- Startup, 3 - Hot Standby, 4 - Hot Shutdown, 5 - Cold Shutdown, 6 - Refueling, D - Defueled FG1 U1]j]jt][

FS1 4

FAI W gJRJ FU1 t]-][j]J

1. Loss of any two barriers and loss or potential loss of the

1. Loss or potential loss of any two barriers.

1. Any loss or any potential loss of either fuel clad or RCS.

1. Any loss or any potential loss of containment.

third barrier.

FC - Fuel Clad RC - Reactor Coolant System CT - Containment Sub-Category Loss Potential Loss Loss Potential Loss Loss Potential Loss

7. RCS Activity

1. Coolant activity > 300 pCi/gm dose equivalent 1-131.

8.

Containment

1. A containment pressure rise

1. Containment pressure > 45 psig Pressure followed by a rapid UNPLANNED and rising.

drop in containment pressure.

OR OR

2. Containment hydrogen > 4%.

2. Containment pressure or sump OR level response not consistent

3. a. Containment pressure >11 with LOCA conditions.

Paloi AND

b.

Less than one full train of depressurization equipment operating.

9.

Containment Note:

Isolation Failure Direct pathways include filtered pathways (e.g., SLCRS).

1

a.

Failure of ALL valves in any one line to close.

AND

b.

Direct downstream pathway to the environment exists after containment isolation I

I__

I signal.

10. EMERGENCY

1. Any condition in the opinion of

1. Any condition in the opinion of

1. Any condition in the opinion of

1. Any condition in the opinion of

1. Any condition in the opinion of

1. Any condition in the opinion of DIRECTOR the EMERGENCY DIRECTOR the EMERGENCY DIRECTOR the EMERGENCY DIRECTOR the EMERGENCY DIRECTOR the EMERGENCY DIRECTOR the EMERGENCY DIRECTOR Judgment that indicates loss of the fuel clad that indicates potential loss of that indicates loss of the RCS that indicates potential loss of that indicates loss of the that indicates potential loss of

barrier, the fuel clad barrier,

barrier, the RCS barrier, containment barrier, the containment barrier.

4-12 Rev. Proposed

Section 4 EMERGENCY ACTION LEVEL Bases Emergency Preparedness Plan RADIOLOGICAL EFFLLU.ENT I ABNORMAL RADIATION LEVELS Modes: 1 - Power Operation. 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown, 5 - Cold Shutdown, 6 - Refueling, D - Defueled GENRA EMRGNC SIT ARE EMEGN

-AER LINSUL I

'a.20) 0 RG1 E5R-OFFSITE dose resulting from an actual or IMMINENT release of gaseous radioactivity greater than 1000 mRem TEDE or 5000 mRem CDE Child Thyroid for the actual or projected duration of the release using actual meteorology.

EALs:

Note:

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

1.

ANY of the following gaseous effluent monitors greater than the reading shown for 15 minutes* or longer:

SLCRS Vent (RM-1VS-110 Ch 7).7.66E+02 cpm Ventilation Vent (RM-1VS-109 Ch 7)..6.42E+02 cpm OR

2.

Dose assessment using actual meteorology indicates doses at or beyond the site boundary of EITHER of the following

> 1000 mRem TEDE.

> 5000 mRem CDE Child Thyroid.

OR

3.

Field survey results at or beyond the site boundary indicate EITHER of the following:

Gamma (closed window) dose rate > 1000 mR/hr for 60 minutes* or longer.

Air sample analysis > 5000 mRem CDE Child Thyroid for one hour of inhalation.

RS1 Douj Bub MLUJ

'MA I

U UZ 0L4JLýJ U LUJ I OFFSITE dose resulting from an actual or IMMINENT release of gaseous radioactivity greater than 100 mRem TEDE or 500 mRem CDE Child Thyroid for the actual or projected duration of the release using actual meteorology.

EALs:

Note:

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

1. ANY of the following gaseous effluent monitors greater than the reading shown for 15 minutes* or longer:

SLCRS Vent (RM-1VS-110 Ch 7).7.66E+01 cpm Ventilation Vent (RM-1VS-109 Ch 7). 6.42E+01 cpm OR

2.

Dose assessment using actual meteorology indicates doses at or beyond the site boundary of EITHER of the following:

> 100 mRem TEDE.

> 500 mRem CDE Child Thyroid.

OR

3.

Field survey results at or beyond the site boundary indicate EITHER of the following:

Gamma (closed window) dose rate> 100 mRlhr for 60 minutes* or longer.

Air sample analysis > 500 mRem CDE Child Thyroid for one hour of inhalation.

Any release of gaseous or liquid radioactivity to the environment greater than 200 times the ODCM limit for 15 minutes or longer.

EALs:

Note:

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

1. ANY of the following gaseous effluent monitors greater than the reading shown for 15 minutes or longer:

SLCRS Vent (RM-1VS-110 Ch 5)...... 6.76E+05 cpm Ventilation Vent (RM-1VS-109 Ch 5). 2.94E+05 cpm OR

2.

ANY of the following liquid effluent monitors > 200 times the High-High alarm setpoint, not to exceed 8.5E+05 cpm, established by a current radioactivity discharge permit for 15 minutes or longer:

Liquid Waste Effluent Monitor (RM-1LW-104)

Laundry and Contaminated Shower Drains Monitor (RM-1 LW-1 16)

OR

3.

Confirmed sample analysis for gaseous or liquid releases > 200 times the ODCM limit for 15 minutes or lonoer.

I ~

~

~mn~

I I RU1 ME3*1 [11 A]

• risin Any release of gaseous or liquid radioactivity to the environment greater than 2 times the ODCM limit for 60 minutes or longer.

EALs:

Note:

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

1.

ANY of the following gaseous effluent monitors greater than the reading shown for 60 minutes or longer:

SLCRS Vent (RM-1VS-110 Ch 5).6.76E+03 cpm Ventilation Vent (RM-1VS-109 Ch 5).2.94E+03 cpm OR

2.

ANY of the following liquid effluent monitors > 2 times the High-High alarm setpoint established by a current radioactivity discharge permit for 60 minutes or longer:

Liquid Waste Effluent Monitor (RM-1LW-104)

Laundry and Contaminated Shower Drains Monitor (RM-1LW-116)

OR

3.

Confirmed sample analysis for gaseous or liquid releases > 2 times the ODCM limit for 60 minutes or longer.

I RA2

'D21flRE f'-

RU2 DEM MEi i Damage to irradiated fuel or loss of water level that has UNPLANNED rise in plant radiation levels.

Iresulted or will result in the uncovering of irradiated fuel EALs."

outside the reactor vessel.

EALs:

1.

a.

UNPLANNED water level drop in the spent fuel pool, transfer canal or reactor cavity as indicated by level 1.

A water level drop in the spent fuel pool, transfer canal

< Tech Spec Minimum (23 feet).

0 or reactor cavity that will result in irradiated fuel AND becoming uncovered.

OR

b.

Area radiation monitor rise resulting in a High-High alarm on ANY of the following:

E

2.

> 1000 mR/hr reading on ANY of the following due to Manipulator Crane Area Monitor (RM-1 RM-203)

O damage to irradiated fuel or loss of water level:

Manipulator Crane Area Monitor (RM-1RM-203)

OR

2.

UNPLANNED area radiation monitor or radiation survey

> 1000 times NORMAL LEVELS.

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

4-13 Rev. Proposed

Section 4 EMERGENCY ACTION LEVEL Bases RADIOLOGICAL EFFLUENT I ABNORMAL RADIATION LEVELS Emergency Preparedness Plan JulRLMEGEC SIEAE MREC Modes: 1 - Power Operation. 2 - Startup, 3 - Hot Standby, 4 -,Hot Shutdown, 5 - Cold Shutdown, 6 - Refueling, D - Defueled RA3 B))INN@

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

EALs:

1.

Dose rate > 15 mR/hr in ANY of the following areas requiring continuous occupancy to maintain plant safety functions:

CONTROL ROOM Central Alarm Station Secondary Alarm Station 4-14 Rev. Proposed

Section 4 EMERGENCY ACTION LEVEL Bases Emergency Preparedness Plan HAZARDS AND OTHER CONDIT!ONS AFFECTING PLANT.SAFETY.

Modes: 1 - Power.Operation. 2-Startup, 3 - Hot Standby, 4 - Hot Shutdown, 5 - Cold Shutdown, 5 - Ref,.lino, 0 - [efieled

_ll

/J1i i i I

U)

Sr HG1 M]RM[-

HOSTILE ACTION resulting in loss of physical control of the facility.

EALs:

1.

A HOSTILE ACTION has occurred such that plant personnel are unable to operate equipment required to maintain safety functions listed below:

Reactivity Control (ability to shut down the reactor and keep it shut down)

RCS inventory (ability to cool the core)

Secondary heat removal (ability to maintain a heat sink)

OR

2.

A HOSTILE ACTION has caused failure of spent fuel cooling systems and IMMINENT fuel damage is likely.

HSI DEMMEl][]

HOSTILE ACTION within the PROTECTED AREA.

EALs:

1. A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by the Security Shift Supervisor.

HAI EMENE*

HOSTILE ACTION within the OWNER CONTROLLED AREA or airborne attack threat.

EALs:

1.

A HOSTILE ACTION is occurring or has occurred within the OWNER CONTROLLED AREA as reported by the Security Shift Supervisor.

OR

2.

A validated notification from the NRC of a LARGE AtRCRAFT attack threat within 30 minutes of the site.

HU1 MERNR-'

Confirmed SECURITY CONDITION or threat which indicates a potential degradation in the level of safety of the plant.

EALs:

1.

A SECURITY CONDITION that does not involve a HOSTILE ACTION as reported by the Security Shift Supervisor.

OR

2.

A credible site specific security threat notification.

OR

3.

A validated notification from the NRC providing information of a LARGE AIRCRAFT threat.

+

-t F

C r-0 U

C..,

HS2 E[r34s 0

CONTROL ROOM evacuation has been initiated and plant control cannot be established.

EALs:

1. a.

CONTROL ROOM evacuation has been initiated.

AND

b.

Control of ANY of the following safety functions is not established from an alternate location within 15 minutes.

Reactivity Control (ability to shut down the reactor and keep it shut down)

RCS inventory (ability to cool the core)

Secondary heat removal (ability to maintain a heat sink)

HA2 DEMME CONTROL ROOM evacuation has been initiated.

EALs:

1.

CONTROL ROOM evacuation has been initiated.

4-15 Rev. Proposed

Section 4 EMERGENCY ACTION LEVEL Bases Emergency Preparedness Plan HAZARDS AND OTHER CONDITIONS AFFECTING. PLANT SAFETY Modes: 1 - Power Onereatcn. - - Startun. 3 - Hot Standbv. 4.- Hot Shutdown. 5 - Cold Shutdown. 6 - Refuelin.

0 - De.fuelrtd A-k ukIJA EVNT Table H-1 HA3 Natural or destructive phenomena affecting VITAL AREAS.

DMERNEM][]

• Cable Tunnel (CV-3)

CONTROL ROOM

• Containment Building

• Demin. Water Storage Tank (1WT-TK-10)

• Diesel Generator Building

• Fuel Building

• Intake Structure Pump Cubicles

• Safeguards (including AFW, Main Steam and Cable Vault Areas)

• Primary Auxiliary Building (except elev. 768')

• RWST (1QS-TK-1)

• Service Building (below elev. 735')

E0 0~

P)

M z

EALs:

1. a.

Seismic event > 0.06g (OBE) acceleration (as indicated by analysis of the Accelerograph Recording System or lit lamp on 2ERS-CCC-1 Seismic Instrumentation Central Control Cabinet)

AND

b. Earthquake confirmed by ANY of the following:

Earthquake felt in plant.

National Earthquake Center.

CONTROL ROOM indication of degraded performance of systems required for the safe shutdown of the plant.

OR

2. Tornado or high winds > 80 mph resulting in EITHER of the following:

VISIBLE DAMAGE to ANY structures in Table H-1 areas containing safety systems or components.

CONTROL ROOM indication of degraded performance of those safety systems.

OR

3. Internal flooding in Table H-1 areas resulting in EITHER of the following:

Electdcal shock hazard that precludes access to operate or monitor safety equipment.

CONTROL ROOM indication of degraded performance of those safety systems.

OR

4. High river level *705 feet MSL resulting in EITHER of the following:

VISIBLE DAMAGE to ANY structures in Table H-1 areas containing safety systems or components.

CONTROL ROOM indication of degraded performance of those safety systems.

HU3 MN[ERD Natural or destructive phenomena affecting the PROTECTED AREA.

EALs:

1.

a.

Seismic event > 0.01g acceleration (as indicated by initiation of the Accelerograph Recording System on All-59, Seismic Accelerograph Operation).

AND

b. Earthquake confirmed by EITHER of the following:

Earthquake felt in plant.

National Earthquake Center.

OR

2. a.

Tornado within the PROTECTED AREA.

OR

b.

High winds >80mph.

OR

3.

Internal flooding in Table H-1 areas that has the potential to affect safety related equipment required by Technical Specifications for the current operating mode.

OR

4.

High river water level > 705 feet MSL.

OR

5.

Low river water level (LR-1CW-101) < 650 feet MSL.

OR

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

15.

OR Low river level (LR-1CW-101) < 650 feet MSL resulting in CONTROL ROOM indication of degraded performance of safety systems located in Table H-1 areas.

OR

6. Turbine failure-generated PROJECTILES resulting in EITHER of the following:

VISIBLE DAMAGE to or penetration of ANY structures in Table H-1 areas containing safety systems or components.

" CONTROL ROOM indication of degraded performance of those safety systems.

4-16 Rev. Proposed

Section 4 EMERGENCY ACTION LEVEL Bases Emergency Preparedness Plan OR

7.

Vehicle crash resulting in EITHER of the following:

VISIBLE DAMAGE to ANY structures in Table H-1 areas containing safety systems or components.

CONTROL ROOM indication of degraded performance of those safety systems.

4-17 Rev. Proposed

Section 4 EMERGENCY ACTION LEVEL Bases Emergency Preparedness Plan HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY Modes: 1.- Power Operation, 2-Startup, 3 - Hot Standby. 4 - Hot Shutdown, 5 - Cold Shutdown. 6 -. Refueling. D - Defueled

,d__'n

,d -

u d_-

0 LI-

'UJ LU HG6

[fl0E2(0[

IHS6 Other conditions exist which in the judgment of the Other conditions exist which in the judgment of the EMERGENCY DIRECTOR warrant declaration of GENERAL EMERGENCY DIRECTOR warrant declaration of SITE EMERGENCY.

AREA EMERGENCY.

EALs:

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

EALs:

1. Other conditions exist which in the judgment of the EMERGENCY DIRECTOR indicate that events are in progress or have occurred which involve actual or likely major failures of plant functions needed for protection of the public or HOSTILE ACTION that results in intentional damage or malicious acts: (1) toward site personnel or equipment that could lead to the likely failure of or, (2) that prevent effective access to equipment needed for the protection of the public. Any releases are not expected to result in exposure levels which exceed EPA PROTECTIVE ACTION GUIDE exposure levels beyond thR -SitA hniind,.rv HA6 IMJ[ER i-Other conditions exist which in the judgment of the EMERGENCY DIRECTOR warrant declaration of an ALERT.

EALs:

1. Other conditions exist which in the judgment of the EMERGENCY DIRECTOR indicate that events are in progress or have occurred which involve actual or 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 GUIDE exposure levels.

HU6 09EH991-Other conditions exist which in the judgment of the EMERGENCY DIRECTOR warrant declaration of an UNUSUAL EVENT.

EALs:

1. Other conditions exist which in the judgment of the EMERGENCY DIRECTOR indicate that events are in progress or have occurred which indicate a potential degradation of the level of safety of the plant or indicate a security threat to facility protection has been initiated. No releases of radioactive material requiring OFFSITE response or monitoring are expected unless further degradation of safety systems occurs.

E-HU1 ILL I)

Damage to a loaded cask CONFINEMENT BOUNDARY.

IL

1.

Dam-Wo a loaded cask CONFINEMENT BOUNDARY.

4-19 Rev. Proposed

Section 4 EMERGENCY ACTION LEVEL Bases Emergency Preparedness Plan SYSTEM MALFUNCTIONS - HOT Modes: 1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown, 5 - Cold Shutdown, 6 - Refueling. D - Defueled

`p-pWRA E MEGN Iii~t

-ft~

'6C LRIUUULV 0

0.

0) 0

-j SG1 IME][

Prolonged toss of all OFFSITE and all ONSITE AC power to emergency busses.

EALs:

1.

a.

Loss of ALL OFFSITE and ALL ONSITE AC power to BOTH AE and DF 4KV emergency busses.

AND

b.

EITHER of the following:

Restoration of EITHER the AE 4KV emergency bus OR OF 4KV emergency bus within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is not likely.

Core Cooling - Red entry conditions met.

ss[

Loss of all OFFSITE and all ONSITE AC power to emergency busses for 15 minutes or longer.

DIEM~

EALs:

Note:

Credit cannot be taken for emergency busses being powered from the other unit's emergency diesel generators.

1. Loss of ALL OFFSITE and ALL ONSITE AC power to BOTH AE and DF 4KV emergency busses for 15 minutes* or longer.

SAl DEM]L)

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

EALs:

1. a.

AC power to AE and OF 4KV emergency busses is reduced to a single power source for 15 minutes* or longer.

AND

b.

Any additional single power source failure will result in loss of ALL AC power to BOTH AE and DF 4KV emergency busses.

Loss of all OFFSITE AC power to emergency busses for 15 minutes or longer.

EALs:

1. Loss of ALL OFFSITE AC power to BOTH AE and OF 4KV emergency busses for 15 minutes* or longer.

0 0L 0.

0 IL 0

-J SS2 MEE Loss of all vital DC power for 15 minutes or longer.

EALs:

1. Bus voltage indication on ALL safety related DC busses less than the following for 15 minutes* or longer:

< 110.4 VDC on Busses 1-1 and 1-2

<110.0 VDC on Busses 1-3 and 1-4 i

SG3 DIE Automatic trip and all manual actions failed to shutdown the reactor and indication of an extreme challenge to the ability to cool the core exists.

EALs:

SS3 EN Automatic trip and manual actions taken within the Controls Area (CA) failed to shutdown the reactor.

SA3 Wf Automatic trip failed to shutdown the reactor and the manual actions taken from the Controls Area (CA) are successful in shutting down the reactor.

SU3

[LfL Inadvertent criticality.

EALs:

1.

UNPLANNED sustained positive startup rate observed on nuclear instrumentation.

EALs:

1. a.

An automatic reactor trip failed to shutdown the reactor as indicated by reactor power > 5%.

EALs:

1. a.

1.

a.

An automatic reactor trip failed to shutdown the reactor as indicated by reactor power > 5%.

AND

b. ALL manual trip actions failed to shutdown the reactor as indicated by reactor power > 5%.

AND

c.

EITHER of the following has occurred:

Core Cooling - Red entry conditions met.

Heat Sink - Red entry conditions met.

AND

b.

Manual trip actions taken within the Controls Area (CA) failed to shutdown the reactor as indicated by reactor power > 5%.

An automatic reactor trip failed to shutdown the reactor.

AND

b.

Manual trip actions taken within the Controls Area (CA) successfully shutdown the reactor as indicated by reactor power

< 5%.

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.

4-20 Rev. Proposed

Section 4 EMERGENCY ACTION LEVEL Bases Emergency Preparedness Plan SYSTEM MALFUNCTIONS - HOT Modes: I - Power operation, 2 -- Startup, 3 - Hot Standbv, 4 - Hot Shutdown, 5 - Cold Shutdown, 6 - Refuelinq, D - Defueled S

ML FUNCTON EMEOT M

1C ST Pepr ion 2 tp o

a 45 lS 6L-ERefuelng

-Defu 114I i,tI IL],q tl Table S-1: Critical Safety Functions

" Reactivity Control (ability to shut down the reactor and keep it shut down)

RCS inventory (ability to cool the core)

Sec;ondary heat removal (ability to maintain a heat sink)

Table S-2: Significant Transients Automatic turbine runback > 25% thermal power Electrical load rejection > 25% full electrical load

" Reactor trip

" Safety Injection actuation SS4 Il Inability to monitor a significant transient in progress.

EALs:

1. a.

Loss of> 75% of EITHER of the following for E M11311I SA4 5112D14 Loss of safety system annunciation or indication in the CONTROL ROOM with either: (1) a significant transient in progress, or (2) COMPENSATORY INDICATIONS are unavailable.

Uo 0

c 15 minutes* or longer:

CONTROL ROOM Annunciator Panels (Al -

A13).

OR CONTROL ROOM critical safety function indications (Table S-1).

AND

b.

A Table S-2 significant transient is in progress.

AND

c.

COMPENSATORY INDICATIONS are unavailable.

EALs:

1. a.

SU4 Ui-199 Loss of safety system annunciation or indication in the CONTROL ROOM for 15 minutes or longer.

EALs:

1. Loss of> 75% of EITHER the following for 15 minutes*

or longer:

CONTROL ROOM Annunciator Panels (Al - A13).

OR CONTROL ROOM critical safety function indications (Table S-1).

Loss of > 75% of EITHER of the following for 15 minutes* or longer:

CONTROL ROOM Annunciator Panels (Al -

A13).

OR CONTROL ROOM critical safety function indications (Table S-1).

AND

b. EITHER of the following:

A Table S-2 significant transient is in progress.

OR COMPENSATORY INDICATIONS are in-.n SU5 DIMlE~

-Inability to reach required operating mode within Technical

.E Specification limits.

-J EALs:

1--

1. Plant is not brought to required operating mode within Technical Specification LCO action statement time.

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.

4-21 Rev. Proposed

Section 4 EMERGENCY ACTION LEVEL Bases Emergency Preparedness Plan SYSTEM MALFUNCTIONS - HOT Modes: 1 - Power Oneration, 2 - StartuD. 3 - Hot Standbv. 4 - Hot Shutdown. 5 - Cold Shutdown. 6 - Refuelina. D - Defueled

.. GEtAiI~~wf

-tt NSA VN SU6 Loss of all ONSITE or OFFSITE communications capabilities.

EALs:

LIMEElJ 1.

.2 0

E E

0.)

2.

Loss of ALL of the following ONSITE communication methods affecting the ability to perform routine operations:

Radios.

Plant page.

Plant telephone system (hardwired).

OR Loss of ALL of the following OFFSITE communications methods affecting the ability to perform OFFSITE notifications:

NRC Emergency Notification System - ENS (Red Phone).

NRC Health Physics Network - HPN.

SU om,,IrertIaI Le ep ones Sw tiar w re an, w re essi.

SU7IM RCS leakage.

EALs:

e Note:

0)M Identified, unidentified and pressure boundary RCS leakage as defined by Technical Specifications.

.-J U)

Relief valve normal operation should be excluded unless oJ it fails to close and cannot be isolated.

1. Unidentified or pressure boundary leakage > 10 gpm.

OR

2. Identified leakage > 25 gpm.

SU9

[I'E[EE "M 0 Fuel clad degradation.

EALs:

1. Letdown Monitor (RM-ICH-101A or B) > 6.OE+04 cpm.

OR

2.

RCS activity > 21 pCi/gm dose equivalent 1-131.

4-22 Rev. Proposed

Section 4 EMERGENCY ACTION LEVEL Bases Emergency Preparedness Plan SYSTEM MALIFUNCT!ONS - COLD Modes: 1 -.Power Operation. 2 - Startup, 3 -Hot Standby, 4 -Hot Shutdown, 5 -Cold Shutdown, 6 - Refueling, D - Defueled m i7-

-- -- 11 1.1-..

- --.- w.--

ý

. -- I..

-SU L V 1 CG7 I

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

EALs:

1.

a.

RCS level < 56% RVLIS Full Range (top of active fuel) for 30 minutes* or longer.

AND

b.

ANY Table C-1 containment challenge indications.

OR

2.

a.

RCS level cannot be monitored with core uncovery for 30 minutes* or longer.

AND

b.

Loss of RCS inventory as indicated by ANY of the following:

Containment Radiation Monitor (RM-1RM-219A or B) > 15 R/hr.

Erratic source range monitor indication.

UNPLANNED level rise in Containment sumps or incore instrument sump.

AND

c.

ANY Table C-1 containment challenge indications.

0) 0-J (I,

Cs7 NE Loss of RCS inventory affecting core decay heat removal capability.

EALs:

1. a.

CONTAINMENT CLOSURE not established.

AND

b.

RCS level < 64% RVLIS Full Range (6" below bottom of hot leg).

OR

2. a.

CONTAINMENT CLOSURE established.

AND

b. RCS level < 56% RVLIS Full Range (top of active fuel).

OR

3.

a. RCS level cannot be monitored for 30 minutes* or longer.

AND b

Loss of RCS inventory as indicated by ANY of the following:

Containment Radiation Monitor (RM-1RM-219A or B) > 15 R/hr.

" Erratic source range monitor indication.

UNPLANNED level rise in Containment sumps or incore instrument sump.

CA7 5] 6 Loss of RCS inventory.

EALs:

1. Loss of RCS inventory as indicated by ANY of the following:

RVLIS Full Range Level (LT-1 RC-131 1) < 65%

(bottom of hot leg).

Refueling Outage Temporary Level Instrument (LI-1RC-481C) < 16 inches (Reduced Inventory Only).

Refueling Outage Temporary Level Instrument (LI-1RC-482C) < 6 inches (Midloop Only).

OR

2. a.

RCS level cannot be monitored for 15 minutes* or longer.

AND

b.

Loss of RCS inventory as indicated by UNPLANNED level rise in Containment sumps or incore instrument sump.

CU7 RCS leakage.

EALs:

Note:

Relief valve normal operation should be excluded unless it fails to dose and cannot be isolated.

1. RCS leakage results in the inability to maintain or restore RCS level > Target Level Band for 15 minutes* or longer.

CU8

[]

UNPLANNED loss of RCS inventory.

EALs:

1. UNPLANNED RCS level drop as indicated by EITHER of-the following:

Refueling Outage Temporary Level Instrument (LI-1RC-481C) < 97 inches (vessel flange) for 15 minutes* or longer when the RCS level band is established above the vessel flange.

OR RCS water level drop below the RCS level band for 15 minutes* or longer when the RCS level band is established below the vessel flange.

OR

2.

a.

RCS level cannot be monitored.

AND

b.

Loss of RCS inventory as indicated by UNPLANNED level rise in containment sumps or incore instrument sumn.

Table C-1: Containment Challenge Indications CONTAINMENT CLOSURE not established.

" Hydrogen concentration > 4% inside containment.

UNPLANNED rise in containment pressure.

I ___________________________________

L ___________________________________

a ___________________________________

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

4-24 Rev. Proposed

Section 4 EMERGENCY ACTION LEVEL Bases Emergency Preparedness Plan SYSTEM MALFUNCTIONS - COLD

.Modes: 1 -Pow-:

Operalion, 2 - St-ortup, 3 -Hot Standby, 4-Hot Shutdown, 5 -Cold Shutdown, 6 -Refueling, D - Defijeled.

SYSTEMRMALFUNCTIONS AE CO11d:Er1I-Pw OE pratin.2-i 3otSad,4Hthudw 5-Cd 1hutown 6-Y R t

0 ee UN___

I I

I r r I Table C-2: RCS Reheat Duration Thresholds RCS Cont Closure Duration Intact with Full RCS N/A

> 60 min*

-Inventory Not Intact Established

> 20 min-OR Not Full RCS Not Established 0 min Inventory If an RCS heat removal system is in operation within this time frame and RCS temperature is being reduced, this EAL is not applicable.

(U 0,

Inability to maintain plant in cold shutdown.

EALs:

Note:

Full inventory is pressurizer level > 22% actual with loop stops either isolated or unisolated.

1. RCS temperature > 2000 F due to an UNPLANNED loss of decay heat removal capability for greater than the specified duration on Table C-2.

OR

2.

a.

RCS temperature cannot be monitored.

AND

b.

RCS pressure rise > 10 psi due to an UNPLANNED loss of decay heat removal capability (this EAL does not apply in RCS solid plant conditions).

CUI0 g C UNPLANNED Loss of decay heat removal capability.

EALs:

1. RCS temperature > 2000 F due to an UNPLANNED loss of decay heat removal capability.

OR

2.

Loss of ALL RCS temperature and RCS level indication for 15 minutes* or longer.

,'1

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

4-25 Rev. Proposed

Section 4 Emergency Preparedness Plan EMERGENCY ACTION LEVEL Bases Containment Radiation Monitoring FC2 Loss:

1.

Containment Radiation Monitor (RM-1 RM-219A or B) > FC2 Line on Graph F-I.

Graph F-1: UI FC2 Loss (CRM Reading for 1% Clad Damage)

" 1E+3 a

C.,

1E+1 0

2 4

6 8

'0 12 14 16 18 20 22 24 25 26 30 32 34 36 36 40 42 44 46 48 Post LOCA Time (Hours after shutdown)

Potential Loss:

None Basis:

Generic The site specific reading is a value which indicates the release of reactor coolant, with elevated activity indicative of fuel damage, into the containment.

The reading should be calculated assuming the instantaneous release and dispersal of the reactor coolant noble gas and iodine inventory associated with a concentration of 300 pCi/gm dose equivalent 1-131 into the containment atmosphere.

Reactor coolant concentrations of this magnitude are several times larger than the maximum concentrations (including iodine spiking) allowed within Technical Specifications and are therefore indicative of fuel damage.

This value is higher than that specified for RC2(L)1. Thus, this threshold indicates a loss of both the Fuel Clad barrier and RCS barrier that appropriately escalates the emergency classification to a SITE AREA EMERGENCY.

There is no potential loss threshold associated with this item.

4-31 Rev. Proposed

Section 4 Emergency Preparedness Plan EMERGENCY ACTION LEVEL Bases RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION RCS Leak Rate RC5 Loss:

1.

RCS leak rate greater than available makeup capacity as indicated by RCS subcooling < 180 F normal containment or < 330 F adverse containment.

Potential Loss:

1.

UNISOLABLE RCS leak exceeding the capacity of one charging pump (129 gpm) in the normal charging mode.

Basis:

Generic Loss Threshold #1 This threshold addresses conditions where leakage from the RCS is greater than available inventory control capacity such that a loss of subcooling has occurred. The loss of subcooling is the fundamental indication that the inventory control systems are inadequate in maintaining RCS pressure and inventory against the mass loss through the leak.

Potential Loss Threshold #1 This threshold is based on the apparent inability to maintain normal liquid inventory within the Reactor Coolant System (RCS) by normal operation of the Chemical and Volume Control System which is considered to be the flow rate equivalent to one charging/makeup pump discharging to the charging header. Isolating letdown is a standard abnormal operating procedure action and may prevent unnecessary classifications when a non-RCS leakage path such as a CVCS leak exists. The intent of this condition is met if attempts to isolate Letdown are NOT successful. Additional charging/makeup pumps being required is indicative of a substantial RCS leak.

Site Specific Loss Threshold #1 RCS subcooling is determined by evaluation of the saturation temperature that corresponds to the indicated reactor coolant system pressure minus the average reactor coolant loop hot leg temperature.

Potential Loss Threshold #1 This threshold is based on the capacity of a single charging pump flow of 129 GPM per UFSAR table 9.1-2.

4-39 Rev. Proposed

Section 4 Emergency Preparedness Plan EMERGENCY ACTION LEVEL Bases Containment Radiation Monitoring CT2 Loss:

None Potential Loss:

1.

Containment Radiation Monitor (RM-1 RM-219A or B) > CT2 Line on Graph F-1.

Graph F-1: U1 CT2 Potential Loss (CRM Reading for 20% Clad Damage) 1 E+5 co0'1 E+4 0

C 0

1E+3 C..)

1E+2 0

2 4

6 8

10 12 14 16 18 20 22 24 26 28 30 32 34 36 38 40 42 44 46 48 Post LOCA Time (Hours after shutdown)

Basis:

Generic There is no loss threshold associated with this item.

The site specific reading is a value which indicates significant fuel damage well in excess of the thresholds associated with both loss of fuel clad and loss of RCS barriers.

As stated in Section 3.8 of NEI 99-01 Rev 5, a major release of radioactivity requiring OFFSITE PROTECTIVE ACTIONS from core damage is not possible unless a major failure of fuel cladding allows radioactive material to be released from the core into the reactor coolant.

Regardless of whether containment is challenged, this amount of activity in containment, if released, could have such severe consequences that it is prudent to treat this as a potential loss of containment, such that a GENERAL EMERGENCY declaration is warranted.

NUREG-1228, "Source Term Estimation During Incident Response to Severe Nuclear Power Plant Accidents," indicates that such conditions do not exist when the amount of clad damage is less than 20%.

4-44 Rev. Proposed

Section 4 Emergency Preparedness Plan EMERGENCY ACTION LEVEL Bases RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HA3 (continued)

Generic These EALs escalate from HU3 in that the occurrence of the event has resulted in VISIBLE DAMAGE to plant structures or areas containing equipment necessary for a safe shutdown, or has caused damage to the safety systems in those structures evidenced by CONTROL ROOM indications of degraded system response or performance. The occurrence of VISIBLE DAMAGE and/or degraded system response is intended to discriminate against lesser events. The initial report should not be interpreted as mandating a lengthy damage assessment prior to classification. No attempt is made in this EAL to assess the actual magnitude of the damage. The significance here is not that a particular system or structure was damaged, but rather, that the event was of sufficient magnitude to cause this degradation.

EALs #2 - #6 These EALs should specify site specific structures or areas that contain safety systems or components and functions required for safe shutdown of the plant. Site specific Safe Shutdown Analysis should be consulted for equipment and plant areas required to establish or maintain safe shutdown.

EAL #1 Seismic events of this magnitude can result in a VITAL AREA being subjected to forces beyond design limits, and thus damage may be assumed to have occurred to plant safety systems.

This threshold should be based on site specific FSAR design basis. See EPRI-sponsored "Guidelines for Nuclear Plant Response to an Earthquake", dated October 1989, for information on seismic event categories.

The National Earthquake Center can confirm if an earthquake has occurred in the area of the plant.

EAL #2 This EAL is based on a tornado striking (touching down) or high winds that have caused VISIBLE DAMAGE to structures containing functions or systems required for safe shutdown of the plant.

The high wind value should be based on site specific FSAR design basis as long as it is within the range of the instrumentation available for wind speed.

EAL #3 This EAL addresses the effect of internal flooding caused by events such as component failures, equipment misalignment, or outage activity mishaps. It is based on the degraded performance of systems, or has created industrial safety hazards (e.g.,

electrical shock) that preclude necessary access to operate or monitor safety equipment. The inability to access, operate or monitor safety equipment represents an actual or substantial potential degradation of the level of safety of the plant.

4-85 Rev. Proposed

Section 4 Emergency Preparedness Plan EMERGENCY ACTION LEVEL Bases RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HA4 (continued)

The use of VISIBLE DAMAGE should not be interpreted as mandating a lengthy damage assessment prior to classification. The declaration of an ALERT and the activation of the TECHNICAL SUPPORT CENTER will provide the EMERGENCY DIRECTOR with the resources needed to perform detailed damage assessments.

The EMERGENCY DIRECTOR also needs to consider any security aspects of the EXPLOSION.

This EAL should specify site specific structures or areas that contain safety systems or components and functions required for safe shutdown of the plant. Site specific Safe Shutdown Analysis should be consulted for equipment and plant areas required to establish or maintain safe shutdown.

Site Specific Table H-1 lists areas that house equipment that is needed to ensure safe shutdown of the plant. Personnel access to those areas may be an important factor in monitoring and controlling equipment operability.

A steam line break or steam EXPLOSION that damages permanent structures or equipment in one of these areas would be classified under this EAL.

Basis Reference(s):

1.

NEI 99-01 Rev 5, HU2

2.

U1 UFSAR Table A.1-1, Category I Structures, Systems, and Components, Rev.

26 4-95 Rev. Proposed

Section 4 Emergency Preparedness Plan EMERGENCY ACTION LEVEL Bases RECOGNITION CATEGORY SYSTEM MALFUNCTIONS - HOT SA1 INITIATING CONDITION:

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

Operating Mode Applicability:

1,2,3,4 EALs:

Note: The EMERGENCY DIRECTOR should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the condition has exceeded, or will likely exceed, the applicable time.

1.

a.

AC power to AE and DF 4KV emergency busses is reduced to a single power source for 15 minutes or longer.

AND

b.

Any additional single power source failure will result in loss of ALL AC power to BOTH AE and DF 4KV emergency busses.

Basis:

Generic The condition indicated by this IC is the degradation of the OFFSITE and ONSITE AC power systems such that any additional single failure would result in a loss of all AC power to emergency buses. This condition could occur due to a loss of OFFSITE power with a concurrent failure of all but one emergency generator to supply power to its emergency busses. Another related condition could be the loss of all OFFSITE power and loss of ONSITE emergency generators with only one train of emergency busses being backfed from the unit main generator, or the loss of ONSITE emergency generators with only one train of emergency busses being backfed from OFFSITE power.

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

Site Specific None Basis Reference(s):

1.

NEI 99-01 Rev 5, SA5

2.

NEI 99-01 Rev 5, FAQ# 36 4-112 Rev. Proposed

Section 4 Emergency Preparedness Plan EMERGENCY ACTION LEVEL Bases RECOGNITION CATEGORY SYSTEM MALFUNCTIONS - HOT SS4 INITIATING CONDITION:

Inability to monitor a significant transient in progress.

Operating Mode Applicability:

1,2,3,4 EALs:

Note: The EMERGENCY DIRECTOR should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the condition has exceeded, or will likely exceed, the applicable time.

1.

a.

Loss of > 75% of EITHER of the following for 15 minutes or longer:

" CONTROL ROOM Annunciator Panels (Al - A13).

OR CONTROL ROOM critical safety function indications (Table S-1).

Table S-1: Critical Safety Functions b.

Reactivity Control (ability to shut down the reactor and keep it shut down)

RCS inventory (ability to cool the core)

Secondary heat removal (ability to maintain a heat sink)

AND A Table S-2 significant transient is in progress.

Table S-2: Significant Transients Automatic turbine runback > 25% thermal reactor power Electrical load rejection > 25% full electrical load Reactor trip Safety Injection actuation AND

c.

COMPENSATORY INDICATIONS are unavailable.

4-123 Rev. Proposed

Section 4 Emergency Preparedness Plan EMERGENCY ACTION LEVEL Bases RECOGNITION CATEGORY SYSTEM MALFUNCTIONS - HOT SA4 INITIATING CONDITION:

Loss of safety system annunciation or indication in the CONTROL ROOM with either:

(1) a significant transient in progress, or (2) COMPENSATORY INDICATIONS are unavailable.

Operating Mode Applicability:

1,2,3,4 EALs:

Note: The EMERGENCY DIRECTOR should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the condition has exceeded, or will likely to exceed, the applicable time.

1.

a.

Loss of > 75% of EITHER of the following for 15 minutes or longer:

" CONTROL ROOM Annunciator Panels (Al - A13).

OR CONTROL ROOM critical safety function indications (Table S-1).

Table S-1: Critical Safety Functions Reactivity Control (ability to shut down the reactor and keep it shut down)

RCS inventory (ability to cool the core)

• Secondary heat removal (ability to maintain a heat sink)

AND

b.

EITHER of the following:

A Table S-2 significant transient is in progress.

Table S-2: Significant Transients Automatic turbine runback > 25% thermal reactor power Electrical load rejection > 25% full electrical load Reactor trip Safety Injection actuation OR COMPENSATORY INDICATIONS are unavailable.

4-126 Rev. Proposed

Section 4 Emergency Preparedness Plan EMERGENCY ACTION LEVEL Bases RECOGNITION CATEGORY SYSTEM MALFUNCTIONS - COLD CU8 INITIATING CONDITION:

UNPLANNED loss of RCS inventory.

Operating Mode Applicability:

6 EALs:

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

1.

UNPLANNED RCS level drop as indicated by EITHER of the following:

Refueling Outage Temporary Level Instrument (LI-1 RC-481 C) < 97 inches (vessel flange) for 15 minutes or longer when the RCS level band is established above the vessel flange.

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

OR

2.

a.

RCS level cannot be monitored.

AND

b.

Loss of RCS inventory as indicated by UNPLANNED level rise in containment sumps or incore instrument sump.

Basis:

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

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

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

4-157 Rev. Proposed

Section 4 Emergency Preparedness Plan EMERGENCY ACTION LEVEL Bases RECOGNITION CATEGORY SYSTEM MALFUNCTIONS - COLD CU8 (continued)

The difference between CU7 and CU8 deals with the RCS conditions that exist between cold shutdown and refueling modes. In cold shutdown the RCS will normally be intact and standard RCS inventory and level monitoring means are available. In the refueling mode the RCS is not intact and RPV level and inventory are monitored by different means.

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

EAL #2 This EAL addresses conditions in the refueling mode when normal means of core temperature indication and RCS level indication may not be available. Redundant means of RCS level indication will normally be installed (including the ability to monitor level visually) to assure that the ability to monitor level will not be interrupted. However, if all level indication were to be lost during a loss of RCS inventory event, the operators would need to determine that RCS inventory loss was occurring by observing sump and tank level changes. Sump and tank level increases must be evaluated against other potential sources of leakage such as cooling water sources inside the containment to ensure they are indicative of RCS leakage.

Site Specific EAL #1 The Reactor Vessel flange is at 739 feet 2 3/8 inches (96.9 inches indicated). RCS level is normally monitored using the following instrument:

0 LI-1RC-481C Reactor vessel level indication (LI-1 RC-481 C) provide accurate indication of water level when the RCS is at atmospheric pressure and above the centerline of the hot leg nozzle elevation.

4-158 Rev. Proposed

Section 4 - EMERGENCY ACTION LEVEL Bases Emergency Preparedness Plan FISSION PRODUCT BARRIER DEGRADATION Modes: 1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown, 5 - Cold Shutdown, 6 - Refuelinr, D - Defueled 1.tLoss of any two barriers and loss or potential loss of the

1. Loss or potential loss of any two barriers.

1. Any loss or any potential loss of either fuel dad or RCS.

1. Any loss or any potential loss of containment.

thrd barIer Graph F-i: Radiation Reading Barrier Thresholds 1.OE+04 1.OE+03 S1.OE+02

-CT2 S----FC2 RC2

~,

F-1.OE+O 1 0

2 4

6 8

10 12 14 16 18 20 22 24 28 28 30 32 34 36 38 40 42 44 46 48 Post LOCA Time (Hours after shutdown) 4-173 Rev. Proposed

Section 4 - EMERGENCY ACTION LEVEL Bases Emergency Preparedness Plan FISSION PRODUCT BARRIER DEGRADATION Modes: 1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown, 5 - Cold Shutdown, 6 - Refueling, D - Defueled FG1-Ell[I1 FS1 E-4]l FAII

]

EIN FUl w]

m I

1. Loss of any two barriers and loss or potential loss of the

1. Loss or potential loss of any two barriers.

1. Any loss or any potential loss of either fuel clad or RCS.

1. Any loss or any potential loss of containment.

third barrier.

FC - Fuel Clad RC - Reactor Coolant System CT - Containment Sub-Category Loss Potential Loss Loss Potential Loss Loss Potential Loss

1. Critical Safety

1. Core Cooling - Red entry

1. Core Cooling - Orange entry

1. RCS Integrity - Red entry

1. Containment - Red entry Function Status conditions met.

conditions met.

conditions met.

conditions met.

OR OR

2. a.

Heat Sink - Red entry

2. a.

Heat Sink - Red entry conditions met.

conditions met.

AND AND

b. Heat Sink is required.

b.

Heat Sink is required.

2.

Containment

1. Containment Radiation Monitor

1. Containment Radiation Monitor

1. Containment Radiation Monitor Rad Monitoring (2RMR-RQ206 or 207)

(2RMR-R0206 or 207)

(2RMR-RQ206 or 207)

> FC2 Line on Graph F-i.

> 1.1E+01 R/hr (RC2 Line on

> CT2 Line on Graph F-i.

Graph F-i).

3. Core

1. Three max core exit

1. Three max core exit

1. a. Three max core exit Temperature thermocouples > 12000 F.

thermocouples > 729* F.

thermocouples > 20000 F.

AND

b.

Restoration procedures not effective within 15 minutes.

OR

2. a. Three max core exit thermocouples > 12000 F.

AND

b.

RVLIS Full Range < 40%

with no RCPs running.

AND

c.

Restoration procedures not effective within 15 minutes.

4.

RCS Level

1. RCS level < Table F-i.

5.

RCS Leak Rate

1. RCS leak rate greater than

1. UNISOLABLE RCS leak Table F-1: RVLIS Thresholds available makeup capacity as exceeding the capacity of one RVLIS RCPs Indication indicated by RCS subcooling charging pump (130 gpm) in the LFull Range 0

40%

< 19* F normal containment or normal charging mode.

S 125%

< 460 F adverse containment.

6.

SG Tube Dynamic Range 2

33

1. RUPTURED SG results in an SI Note:

Leakage /

3 60%

actuation.

A prolonged release is greater than 4 Rupture hours.

1. RUPTURED SG is also FAULTED outside of containment.

OR

2. a.

Primary-to-Secondary leak rate

> 10 gpm.

AND

b.

UNISOLABLE prolonged steam release from affected SG to the environment.

4-174 Rev. Proposed

Section 4 - EMERGENCY ACTION LEVEL Bases Emergency Preparedness Plan FISSION PRODUCT BARRIER DEGRADATION Modes: 1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown, 5 - Cold Shutdown, 6 - Refueling, D - Defueled JFG!

[flNO!*

FS1 NL2,U3 FAl Lij U LN F U 1.IF I

1. Loss of any two barriers and loss or potential loss of the

1. Loss or potential loss of any two barriers.

1. Any loss or any potential loss of either fuel clad or RCS.

1. Any loss or any potential loss of containment.

third barrier.

FC - Fuel Clad RC - Reactor Coolant System CT - Containment Sub-Category Loss Potential Loss Loss Potential Loss Loss Potential Loss

7.

RCS Activity

1. Coolant activity > 300 jjCilgm dose equivalent 1-131.

8. Containment

1. A containment pressure rise

1. Containment pressure > 45 psig Pressure followed by a rapid UNPLANNED and rising.

drop in containment pressure.

OR OR

2. Containment hydrogen > 4%.

2. Containment pressure or sump OR level response not consistent with LOCA conditions.

3.

a. Containment pressure

> 11 psig.

AND

c.

Less than one full train of depressurization equipment operating.

9. Containment Note:

Isolation Failure Direct pathways include filtered pathway (e.g., SLCRS).

1. a.

Failure of ALL valves in any one line to close.

AND

b.

Direct downstream pathway to the environment exists after containment isolation signal.

10. EMERGENCY

1. Any condition in the opinion of

1. Any condition in the opinion of

1. Any condition in the opinion of

1. Any condition in the opinion of

1. Any condition in the opinion of

1. Any condition in the opinion of DIRECTOR the EMERGENCY DIRECTOR the EMERGENCY DIRECTOR the EMERGENCY DIRECTOR the EMERGENCY DIRECTOR the EMERGENCY DIRECTOR the EMERGENCY DIRECTOR Judgment that indicates loss of the fuel clad that indicates potential loss of that indicates loss of the RCS that indicates potential loss of that indicates loss of the that indicates potential loss of

barrier, the fuel clad barrier,

barrier, the RCS barrier, containment barrier, the containment barrier.

4-175 Rev. Proposed

Section 4 - EMERGENCY ACTION LEVEL Bases Ennerclency Preparedness Plan RADIOLOGICAL EFFLUENT / ABNORMAL RADIATION LEVELS Modes: 1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown, 5 - Cold Shutdown, 6 - Refueling, D - Defueted RA3 Rise in radiation levels within the facility that impedes

-j operation of systems required to maintain plant safety functions.

EALs:

1. Dose rate > 15 mR/hr in ANY of the following areas requiring continuous occupancy to maintain plant safety functions:

E CONTROL ROOM 0

Central Alarm Station Secondary Alarm Station 4-177 Rev. Proposed

Section 4 - EMERGENCY ACTION LEVEL Bases Emergency Preparedness Plan I

HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY

!HG1 0[flE ER][

HS1 E]EEHN flD@

HOSTILE ACTION resulting in loss of physical control of the HOSTILE ACTION within the PROTECTED AREA.

facility.

EALs:

EALs:

1.

A HOSTILE ACTION is occurring or has occurred within

1. A HOSTILE ACTION has occurred such that plant the PROTECTED AREA as reported by the Security personnel are unable to operate equipment required to Shift Supervisor.

maintain safety functions listed below:

Reactivity Control (ability to shut down the reactor and keep it shut down)

RCS inventory (ability to cool the core)

Secondary heat removal (ability to maintain a heat sink)

OR

2.

A HOSTILE ACTION has caused failure of spent fuel cnnlinn s*vt,~m

.and IMMIINFNT fult dvam~ar,p is likely Modes: 1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown, 5 - Cold Shutdown, 6 - Refueling, D - Defueled HAl U][ E 3]E E[6))g HU1

-N EE4

- [] [D1 HOSTILE ACTION within the OWNER CONTROLLED Confirmed SECURITY CONDITION or threat which indicates AREA or airborne attack threat, a potential degradation in the level of safety of the plant.

EALs:

EALs:

1.

A HOSTILE ACTION is occurring or has occurred within

1.

A SECURITY CONDITION that does not involve a the OWNER CONTROLLED AREA as reported by the HOSTILE ACTION as reported by the Security Shift Security Shift Supervisor.

Supervisor.

OR OR

2.

A validated notification from the NRC of a LARGE

2.

A credible site specific security threat notification.

AIRCRAFT attack threat within 30 minutes of the site.

OR

3.

A validated notification from the NRC providing information of a LARGE AIRCRAFT threat.

a z-t-an IM IN N

fuldmn s+kl C

0 U(U w

C.,

HS2 MME]rE((rIr CONTROL ROOM evacuation has been initiated and plant control cannot be established.

EALs:

1.

a. CONTROL ROOM evacuation has been initiated.

AND

b. Control of ANY of the following safety functions is not established from an alternate location within 15 minutes.

Reactivity Control (ability to shut down the reactor and keep it shut down)

RCS inventory (ability to cool the core)

Secondary heat removal (ability to maintain a heat sink)

HA2 SM 2[ 9H)) O(-

CONTROL ROOM evacuation has been initiated.

EALs:

1. CONTROL ROOM evacuation has been initiated.

4-178 Rev. Proposed

Section 4 - EMERGENCY ACTION LEVEL Bases Emergency Preparedness Plan Section 4-EMERGENCY ACTION LEVEL Bases Emergency Preparedness Plan HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY Modes: 1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown, 5 - Cold Shutdown, 6 - Refueling, D - Defueled I~

~~~

GEEA EMREC IEAE MRENYAETUUULVN I

l..irr,~~r KTable H-I Cable Vault and Rod Control Bldg Containment Building Control Building Demin. Water Storage (2FWE-TK210)

Diesel Generator Building Fuel Handling Building Intake Structure Pump Cubicles Main Steam Valve Room Primary Aux. Building (except elev. 773')

RWST (2QSS-TK21)

Safeguards Building Service Building (except FW Reg Vlv Rm) 0 E

0

,. t C3 0-.0 o

0.

0.-

Zu Natural or destructive phenomena affecting VITAL AREAS.

EALs:

1. a.

Seismic event > 0.06g (OBE) acceleration (as indicated by lit lamp on 2ERS-CCC-1, Seismic Instrumentation Central Control Cabinet).

AND

b. Earthquake confirmed by ANY of the following:

Earthquake felt in plant.

National Earthquake Center.

CONTROL ROOM indication of degraded performance of systems required for the safe shutdown of the plant.

OR

2.

Tornado or high winds > 80 mph resulting in EITHER of the following:

VISIBLE DAMAGE to ANY structures in Table H-1 areas containing safety systems or components.

CONTROL ROOM indication of degraded performance of those safety systems.

OR

3.

Internal flooding in Table H-1 areas resulting in EITHER of the following:

Electrical shock hazard that precludes access to operate or monitor safety equipment.

CONTROL ROOM indication of degraded performance of those safety systems.

OR

4.

High river level > 705 feet MSL resulting in EITHER of the following:

VISIBLE DAMAGE to ANY structures in Table H-1 areas containing safety systems or components.

CONTROL ROOM indication of degraded performance of those safety systems.

OR

5.

Low river level (LR-1CW-1o01) < 650 feet MSL resulting in CONTROL ROOM indication of degraded performance of safety systems located in Table H-I areas.

OR

6.

Turbine failure-generated PROJECTILES resulting in EITHER of the following:

VISIBLE DAMAGE to or penetration of ANY structures in Table H-1 areas containing safety systems or components.

CONTROL ROOM indication of degraded performance of those safety systems.

OR

7.

Vehicle crash resulting in EITHER of the following:

VISIBLE DAMAGE to ANY structures in Table H-1 areas containing safety systems or components.

CONTROL ROOM indication of degraded performance of those safety systems.

HU3

-gDLrED]

Natural or destructive phenomena affecting the PROTECTED AREA.

EALs:

1. a.

Seismic event > 0.01g acceleration (as indicated by initiation of the Accelerograph Recording System on Ann A10-5H, Init of Seismic Exceed Preset and/or Spectral Accelerations).

AND

b.

Earthquake confirmed by EITHER of the following:

Earthquake felt in plant.

National Earthquake Center.

OR

2. a.

Tornado within the PROTECTED AREA.

OR

b.

High winds >80 mph.

OR I

3.

Internal flooding in Table H-1 areas that has the potential to affect safety related equipment required by Technical Specifications for the current operating mode.

OR

4.

High river water level > 705 feet MSL.

OR

5. Low river water level (LR-1 CW-101) < 650 feet MSL.

OR

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

4-179 Rev. Proposed

Section 4 - EMERGENCY ACTION LEVEL Bases Enremency Preparedness Plan Section 4-EMERGENCY ACTION LEVEL Bases Emergency Preparedness Plan HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY Modes: 1 - Power Operation. 2 - Startuo. 3 - Hot Standby. 4 - Hot Shutdown. 5 - Cold Shutdown. 6 - Refuelina. D - Defueled H

A SN O ERCO IO AF TN P NS EMds1PwApto2Sau.-ttdLn 5old SUtdon.-EfNT

• +,+i _"

• ,+LJn,..m.l tn I

E 0

JI-z UJI HG6 B-il NR[ D Other conditions exist which in the judgment of the EMERGENCY DIRECTOR warrant declaration of GENERAL EMERGENCY.

EALs:

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

I HS6..............

i..

Other conditions exist which in the judgment of the EMERGENCY DIRECTOR warrant declaration of SITE AREA EMERGENCY.

EALs:

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

f nllllI~

~

LLJ~LiP Other conditions exist which in the judgment of the EMERGENCY DIRECTOR warrant declaration of an ALERT.

EALs:

1. Other conditions exist which in the judgment of the EMERGENCY DIRECTOR indicate that events are in progress or have occurred which involve actual or 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 GUIDE exposure levels.

HU6 lFEl4][.

Other conditions exist which in the judgment of the EMERGENCY DIRECTOR warrant declaration of an UNUSUAL EVENT.

EALs:

1. Other conditions exist which in the judgment of the EMERGENCY DIRECTOR indicate that events are in progress or have occurred which indicate a potential degradation of the level of safety of the plant or indicate a security threat to facility protection has been initiated. No releases of radioactive material requiring OFFSITE response or monitoring are expected unless further degradation of safety systems occurs.

I E-HUI U..)

Damage to a loaded cask CONFINEMENT BOUNDARY.

I1 EALs:

1,Dmg o

oddcskCNIEENIONAY 4-181 Rev. Proposed

Section 4 - EMERGENCY ACTION LEVEL Bases Emergency Preparedness Plan Section 4-EMERGENCY ACTION LEVEL Bases Emerciencv Preoaredness Plan v SYSTEM MALFUNCTIONS - HOT Modes: 1 - Power Oneration. 2 - Startuo. 3 - Hot Standbv. 4 - Hot Shutdown. 5 - Cold Shutdown. 6 - Refuelino. D - Defueled GENRA EMREC SIT ARE EMREC ALR PNSAEVN 0

0 U/)

0

-I SGI MEE Prolonged loss of all OFFSITE and all ONSITE AC power to emergency busses.

EALs:

1. a.

Loss of ALL OFFSITE and ALL ONSITE AC power to BOTH AE and DF 4KV emergency busses.

AND

b.

EITHER of the following:

Restoration of EITHER the AE 4KV emergency bus OR DF 4KV emergency bus within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is not likely.

Conre Conlino - Red entry nonditions met.

SS1 Loss of all OFFSITE and all ONSITE AC power to emergency busses for 15 minutes or longer.

EALs:

Note:

Credit cannot be taken for emergency busses being powered from the other unit's emergency diesel generators.

1.

Loss of ALL OFFSITE and ALL ONSITE AC power to BOTH AE and DF 4KV emergency busses for 15 minutes* or longer.

D[21 D14]

SA1 DIE AC power capability to emergency busses reduced to a single source for 15 minutes or longer.

EALs:

UH Sul Loss of all OFFSITE AC power to emergency busses for 15 minutes or longer.

EALs:

1. Loss of ALL OFFSITE AC power to BOTH AE and DF 4KV emergency busses for 15 minutes* or longer.

1. a.,A power to Ar-and OF q*,v emergency busses is reduced to a single power source for 15 minutes* or longer.

AND

b.

Any additional single power source failure will result in loss of ALL AC power to BOTH AE and DF 4KV emergency busses.

O SS2 LjgI14 mLoss of all vital DC power for 15 minutes or longer.

0 (0 0 EALs:

0"

1. < 110.4 VDC on ALL safety related DC busses (2-1, 2-2, 2-3 and 2-4) for 15 minutes* or longer.

SG3 DIE SS3 rfri SA3

[DrE SU3 E2 Automatic trip and all manual actions failed to shutdown the Automatic trip and manual actions taken within the Controls Automatic trip failed to shutdown the reactor and the manual Inadvertent criticality.

reactor and indication of an extreme challenge to the ability Area (CA) failed to shutdown the reactor, actions taken from the Controls Area (CA) are successful in EALs:

to cool the core exists.

EALs:

shutting down the reactor.

EA~s.

EALs:

1. UNPLANNED sustained positive startup rate observed Es

1. a.

An automatic reactor trip failed to shutdown the on nuclear instrumentation.

0:

1. a. An automatic reactor trip failed to shutdown the reactor as indicated by reactor power > 5%

1. a.

An automatic reactor trip failed to shutdown the reactor as indicated by reactor power > 5%.

AND reactor.

ONDAND AND

b.

Manual trip actions taken within the Controls Area AND

b. ALL manual trip actions failed to shutdown the (CA) failed to shutdown the reactor as indicated by

b.

Manual trip actions taken within the Controls Area u.I reactor as indicated by reactor power > 5%.

reactor power > 5%.

(CA) successfully shutdown the reactor as indicated AND by reactor power < 5%.

c. EITHER of the following has occurred:

" Core Cooling - Red entry conditions met.

H, Heat Sink - Red entry conditions met.

• 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.

4-182 Rev. Proposed

Section 4 - EMERGENCY ACTION LEVEL Bases Emergency Preparedness Plan Section 4-EMERGENCY ACTION LEVEL Bases Emergency Preparedness Plan SYSTEM MALFUNCTIONS - HOT Modes: 1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown, 5 - Cold Shutdown, 6 - Refueling, 0 - Defuefed GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUALEVENT I

I I

Table S-1: Critical Safety Functions Reactivity Control (ability to shut down the reactor and keep it shut down)

RCS inventory (ability to cool the core)

Secondary heat removal (ability to maintain a heat sink)

SS4 EE[4I Inability to monitor a significant transient in progress.

EALs:

1. a.

Loss of> 75% of EITHER of the following for 15 minutes* or longer:

CONTROL ROOM Annunciator Panels (Al, A2, A4 - All).

OR I

O 25 0

Table S-2: Significant Transients Automatic turbine runback > 25% thermal power Electrical load rejection > 25% full electrical load Reactor trip Safety Injection actuation CONTROL ROOM critical safety function indications (Table S-1).

AND

b.

A Table S-2 significant transient is in progress.

AND

c.

COMPENSATORY INDICATIONS are unavailable.

SA4 iN L2]

[3] L4 Loss of safety system annunciation or indication in the CONTROL ROOM with either: (1) a significant transient in progress, or (2) COMPENSATORY INDICATIONS are unavailable.

EALs:

1. a.

Loss of > 75% of EITHER of the following for 15 minutes* or longer:

CONTROL ROOM Annunciator Panels (Al, A2, A4 - All).

OR CONTROL ROOM critical safety function indications (Table S-1).

AND

b. EITHER of the following:

" A Table S-2 significant transient is in progress.

OR

" COMPENSATORY INDICATIONS are unavailable.

ISU4 r1J L2ULI Loss of safety system annunciation or indication in the CONTROL ROOM for 15 minutes or longer.

EALs:

1. Loss of> 75% of EITHER of the following for 15 minutes* or longer:

CONTROL ROOM Annunciator Panels (Al, A2, A4 -

All).

OR CONTROL ROOM critical safety function indications (Table S-1).

I

+

+

t I

4.E

-J Ui I-SUs DEEP Inability to reach required operating mode within Technical Specification limits.

EALs:

1. Plant is not brought to required operating mode within Technical Specification LCO action statement time.

4

-I-4 1

SU6 Loss of all ONSITE or OFFSITE communications capabilities.

EALs:

DUDE][

.2E E

0Q-

1. Loss of ALL of the following ONSITE communication methods affecting the ability to perform routine operations:

Radios.

Plant page.

Plant telephone system (hardwired).

OR

2.

Loss of ALL of the following OFFSITE communications methods affecting the ability to perform OFFSITE notifications:

NRC Emergency Notification System - ENS (Red Phone).

NRC Health Physics Network - HPN.

Commercial teleDhones (hardwired and wireless'.

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.

4-183 Rev. Proposed

Section 4 - EMERGENCY ACTION LEVEL Bases Emerqencv Preparedness Plan SYSTEM MALFUNCTIONS - HOT Modes: 1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown, 5 - Cold Shutdown, 6 - Refueling, D Defueled SU7 RCS leakage.

EALs:

Note:.

D. '-Identified, unidentified and pressure boundarRCSy ROS

.,i leakage as defined by Technical.Specifications ;.,';

.e"':Relief valve normal operation should.be excluded unless.1 S'

'it fails to close and cannot be isolated..::

1.

Unidentified or pressure boundary leakage > 10 gpm.

OR

2.

Identified leakage > 25 gpm.

SU9 ME E-1"

.0 0

Fuel clad degradation.

EALs:

1.

Letdown Monitor (2CHS-RO101B) > 2.98E+03 pCI/cc.

LL 0)

LLW OR

2.

RCS activity

• 21 JCilgm oseeuvret111 4-184 Rev. Proposed

Section 4 - EMERGENCY ACTION LEVEL Bases Emergency Preparedness Plan Section 4-EMERGENCY ACTION LEVEL Bases Emergency Preparedness Plan SYSTEM MALFUNCTIONS - COLD I

GENERAL EMERGENCY SITE AREA EMERGENCY I

Table C-2: RCS Reheat Duration Thresholds RCS I Cont Closure Duration Intact with Full RCS N/A

> 60 min*

I nventory I

Not Intact Established

> 20 min-OR v e No ull RCS Not Established 0 mm

.S Inventor If an RCS heat removal system is in operation within e

this time frame and RCS temperature is being X

reduced, this EAL is not applicable.

Modes: 1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown, 5 - Cold Shutdown, 6 - Refueling, D - Defueled CA10 gJ CU10

[5 Ee]

Inability to maintain plant in cold shutdown.

UNPLANNED Loss of decay heat removal capability.

EALs:

EALs:

Note:

1. RCS temperature > 2000 F due to an UNPLANNED loss Full inventory is pressurizer level s 22% actual with loop of decay heat removal capability.

stops either isolated or unisolated.

OR 1.

RCS temperature > 200 F due loan UNPLANNED loss

2.

Loss of ALL RCS temperature and RCS level indication of decay heat removal capability for greater than the for 15 minutes or longer.

specified duration on Table C-2.

OR

2.

a.

RCS temperature cannot be monitored.

AND

b.

RCS pressure rise > 10 psi due to an UNPLANNED loss of decay heat removal capability (this EAL does not apply in RCS solid plant conditions).

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

4-187 Rev. Proposed

Section 4 Emergency Preparedness Plan EMERGENCY ACTION LEVEL Bases RECOGNITION CATEGORY SYSTEM MALFUNCTIONS - COLD Containment Radiation Monitoring FC2 Loss:

1.

Containment Radiation Monitor (2RMR-RQ206 or 207) > FC2 Line on Graph F-1.

Graph F-I: U2 FC2 Loss (CRM Reading for 1% Clad Damage)

_1E+3 0

~1E+2 0

1E+1 0

2 4

6 8

10 12 14 16 18 20 22 24 26 28 30 32 34 36 38 40 42 44 46 48 Post LOCA Time (Hours after shutdown)

Potential Loss:

None Basis:

Generic The site specific reading is a value which indicates the release of reactor coolant, with elevated activity indicative of fuel damage, into the containment.

The reading should be calculated assuming the instantaneous release and dispersal of the reactor coolant noble gas and iodine inventory associated with a concentration of 300 pjCi/gm dose equivalent 1-131 into the containment atmosphere.

Reactor coolant concentrations of this magnitude are several times larger than the maximum concentrations (including iodine spiking) allowed within technical specifications and are therefore indicative of fuel damage.

This value is higher than that specified for RC2(L)1. Thus, this threshold indicates a loss of both the Fuel Clad barrier and RCS barrier that appropriately escalates the EMERGENCY CLASSIFICATION LEVEL to a SITE AREA EMERGENCY.

There is no potential loss threshold associated with this item.

4-193 Rev. Proposed

Section 4 Emergency Preparedness Plan EMERGENCY ACTION LEVEL Bases RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION RCS Leak Rate RC5 Loss:

1.

RCS leak rate greater than available makeup capacity as indicated by RCS subcooling < 190 F normal containment or < 460 F adverse containment.

Potential Loss:

1.

UNISOLABLE RCS leak exceeding the capacity of one charging pump (130 gpm) in the normal charging mode.

Basis:

Generic Loss Threshold #1 This threshold addresses conditions where leakage from the RCS is greater than available inventory control capacity such that a loss of subcooling has occurred. The loss of subcooling is the fundamental indication that the inventory control systems are inadequate in maintaining RCS pressure and inventory against the mass loss through the leak.

Potential Loss Threshold #1 This threshold is based on the apparent inability to maintain normal liquid inventory within the Reactor Coolant System (RCS) by normal operation of the Chemical and Volume Control System which is considered to be the flow rate equivalent to one charging/makeup pump discharging to the charging header. Isolating letdown is a standard abnormal operating procedure action and may prevent unnecessary classifications when a non-RCS leakage path such as a CVCS leak exists. The intent of this condition is met if attempts to isolate Letdown are NOT successful. Additional charging/makeup pumps being required is indicative of a substantial RCS leak.

Site Specific Loss Threshold #1 RCS subcooling is determined by evaluation of the saturation temperature that corresponds to the indicated reactor coolant system pressure minus the average reactor coolant loop hot leg temperature or average in-core thermocouple.

Potential Loss Threshold #1 This threshold is based on the capacity of a single charging pump flow of 130 GPM per UFSAR 9.3.4.3.8.

4-201 Rev. Proposed

Section 4 EMERGENCY ACTION LEVEL Bases Emergency Preparedness Plan RECOGNITION CATEGORY SYSTEM MALFUNCTIONS - COLD Containment Radiation Monitoring Loss:

None Potential Loss:

CT2

1.

Containment Radiation Monitor (2RMR-RQ206 or 207) > CT2 Line on Graph F-I.

Graph F-I: U2 CT2 Potential Loss (CRM Reading for 20% Clad Damage) 1E+5 1 E+4 0

to0 (02 N

w 1E+3 0

U 1 E+2 0

2 4

6 8

10 12 14 16 18 20 22 24 26 28 30 32 34 36 38 40 42 44 46 48 Post LOCA Time (Hours after shutdown)

Basis:

Generic There is no loss threshold associated with this item.

The site specific reading is a value which indicates significant fuel damage well in excess of the thresholds associated with both loss of fuel clad and loss of RCS barriers.

As stated in Section 3.8 of NEI 99-01 Rev 5, a major release of radioactivity requiring OFFSITE PROTECTIVE ACTIONS from core damage is not possible unless a major failure of fuel cladding allows radioactive material to be released from the core into the reactor coolant.

Regardless of whether containment is challenged, this amount of activity in containment, if released, could have such severe consequences that it is prudent to treat this as a potential loss of containment, such that a GENERAL EMERGENCY declaration is warranted.

NUREG-1228, "Source Term Estimation During Incident Response to Severe Nuclear Power Plant Accidents," indicates that such conditions do not exist when the amount of clad damage is less than 20%.

4-206 Rev. Proposed

Section 4 Emergency Preparedness Plan EMERGENCY ACTION LEVEL Bases RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HA3 (continued)

Basis:

Generic These EALs escalate from HU3 in that the occurrence of the event has resulted in VISIBLE DAMAGE to plant structures or areas containing equipment necessary for a safe shutdown, or has caused damage to the safety systems in those structures evidenced by CONTROL ROOM indications of degraded system response or performance. The occurrence of VISIBLE DAMAGE and/or degraded system response is intended to discriminate against lesser events. The initial report should not be interpreted as mandating a lengthy damage assessment prior to classification. No attempt is made in this EAL to assess the actual magnitude of the damage. The significance here is not that a particular system or structure was damaged, but rather, that the event was of sufficient magnitude to cause this degradation.

EALs #2 - #6 These EALs should specify site specific structures or areas that contain safety systems or components and functions required for safe shutdown of the plant. Site specific Safe Shutdown Analysis should be consulted for equipment and plant areas required to establish or maintain safe shutdown.

EAL #1 Seismic events of this magnitude can result in a VITAL AREA being subjected to forces beyond design limits, and thus damage may be assumed to have occurred to plant safety systems.

This threshold should be based on site specific FSAR design basis. See EPRI-sponsored "Guidelines for Nuclear Plant Response to an Earthquake", dated October 1989, for information on seismic event categories.

The National Earthquake Center can confirm if an earthquake has occurred in the area of the plant.

EAL #2 This EAL is based on a tornado striking (touching down) or high winds that have caused VISIBLE DAMAGE to structures containing functions or systems required for safe shutdown of the plant.

The high wind value should be based on site specific FSAR design basis as long as it is within the range of the instrumentation available for wind speed.

4-248 Rev. Proposed

Section 4 Emergency Preparedness Plan EMERGENCY ACTION LEVEL Bases RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HA4 (continued)

The use of VISIBLE DAMAGE should not be interpreted as mandating a lengthy damage assessment prior to classification. The declaration of an ALERT and the activation of the TECHNICAL SUPPORT CENTER will provide the EMERGENCY DIRECTOR with the resources needed to perform detailed damage assessments.

The EMERGENCY DIRECTOR also needs to consider any security aspects of the EXPLOSION.

This EAL should specify site specific structures or areas that contain safety systems or components and functions required for safe shutdown of the plant. Site specific Safe Shutdown Analysis should be consulted for equipment and plant areas required to establish or maintain safe shutdown.

Site Specific Table H-1 lists areas that house equipment that is needed to ensure safe shutdown of the plant. Personnel access to those areas may be an important factor in monitoring and controlling equipment operability. Table H-1 includes structures that are in contact with or immediately adjacent to (directly impacts or obstructs) the areas that actually contain the equipment of concern.

A steam line break or steam EXPLOSION that damages permanent structures or equipment in one of these areas would be classified under this EAL.

Basis Reference(s):

1.

NEI 99-01 Rev 5, HA2

2.

U2 UFSAR Table 3.2-1, Quality Assurance Category I and Seismic Category I Systems and Components, Rev 19

3.

U2 UFSAR, Table 3.2-2, Classification of Structures, Rev 19 4-258 Rev. Proposed

Section 4 Emergency Preparedness Plan EMERGENCY ACTION LEVEL Bases RECOGNITION CATEGORY SYSTEM MALFUNCTIONS - HOT SA1 INITIATING CONDITION:

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

Operating Mode Applicability:

1,2,3,4 EALs:

Note: The EMERGENCY DIRECTOR should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the condition has exceeded, or will likely exceed, the applicable time.

1.

a.

AC power to AE and DF 4KV emergency busses is reduced to a single power source for 15 minutes or longer.

AND

b.

Any additional single power source failure will result in loss of ALL AC power to BOTH AE and DF 4KV emergency busses.

Basis:

Generic The condition indicated by this IC is the degradation of the OFFSITE and ONSITE AC power systems such that any additional single failure would result in a loss of all AC power to emergency buses. This condition could occur due to a loss of OFFSITE power with a concurrent failure of all but one emergency generator to supply power to its emergency busses. Another related condition could be the loss of all OFFSITE power and loss of ONSITE emergency generators with only one train of emergency busses being backfed from the unit main generator, or the loss of ONSITE emergency generators with only one train of emergency busses being backfed from OFFSITE power.

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

Site Specific None Basis Reference(s):

1.

NEI 99-01 Rev 5, SA5

2.

NEI 99-01 Rev 5, FAQ# 36 4-275 Rev. Proposed

Section 4 Emergency Preparedness Plan EMERGENCY ACTION LEVEL Bases RECOGNITION CATEGORY SYSTEM MALFUNCTIONS - HOT SS4 INITIATING CONDITION:

Inability to monitor a significant transient in progress.

Operating Mode Applicability:

1,2,3,4 EALs:

Note: The EMERGENCY DIRECTOR should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the condition has exceeded, or will likely exceed, the applicable time.

1.

a.

Loss of > 75% of EITHER of the following for 15 minutes or longer:

CONTROL ROOM Annunciator Panels (Al, A2, A4 - Al 1).

OR CONTROL ROOM critical safety function indications (Table S-1).

Table S-1: Critical Safety Functions Reactivity Control (ability to shut down the reactor and keep shut down)

RCS inventory (ability to cool the core)

" Secondary heat removal (ability to maintain a heat sink)

AND

b.

A Table S-2 significant transient is in progress.

Table S-2: Significant Transients Automatic turbine runback > 25% thermal reactor power Electrical load rejection > 25% full electrical load 0

Reactor trip Safety Injection actuation AND

c.

COMPENSATORY INDICATIONS are unavailable.

it 4-285 Rev. Proposed

Section 4 Emergency Preparedness Plan EMERGENCY ACTION LEVEL Bases RECOGNITION CATEGORY SYSTEM MALFUNCTIONS - HOT SA4 INITIATING CONDITION:

Loss of safety system annunciation or indication in the CONTROL ROOM with either:

(1) a significant transient in progress, or (2) COMPENSATORY INDICATIONS are unavailable.

Operating Mode Applicability:

1,2,3,4 EALs:

Note: The EMERGENCY DIRECTOR should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the condition has exceeded, or will likely exceed, the applicable time.

1.

a.

Loss of > 75% of EITHER of the following for 15 minutes or longer:

CONTROL ROOM Annunciator Panels (Al, A2, A4 - Al 1).

OR

" CONTROL ROOM critical safety function indications (Table S-1).

Table S-1: Critical Safety Functions Reactivity Control (ability to shut down the reactor and keep it shut down)

RCS inventory (ability to cool the core)

• Secondary heat removal (ability to maintain a heat sink)

AND

b.

EITHER of the following:

0 A Table S-2 significant transient is in progress.

Table S-2: Significant Transients Automatic turbine runback > 25% thermal reactor power Electrical load rejection > 25% full electrical load Reactor trip Safety Injection actuation OR COMPENSATORY INDICATIONS are unavailable.

4-288 Rev. Proposed

Appendix 4 Beaver Valley Power Station Unit No. 1 EAL Evaluation

Beaver Valley Power Station Unit No. 1 EAL Evaluation DIFFERENCES - DEVIATIONS The items considered to be differences or deviations are based on the definitions provided in RIS 2003-18, Supplement 2. Any plant EAL [or Initiating Condition (IC) or Fission Product Barrier (FPB) threshold value] that does not meet the intent of the NEI 99-01, Revision 5 guidance or may result in an event being classified differently from the guidance is identified as a deviation and will be listed as such in this evaluation.

The basis section for each of the deviations documents the rationale for not adopting the NEI 99-01, Revision 5 guidance. Items identified as deviations will not be implemented without prior NRC review and approval.

ADMINISTRATIVE CHANGES The following changes apply throughout the set of EALs and are not specifically identified in the comparison tables:

1.

The NEI phrase "NOTIFICATION OF UNUSUAL EVENT" has been changed to "UNUSUAL EVENT" to sustain common terminology.

2.

The IC identification numbering has been modified to allow consistent grouping by event category.

3.

Numerical values, signs and key words of threshold values are bolded for emphasis.

4.

The NEI note "[T]he Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the

[condition will likely exceed the applicable time.] or [condition (duration) has exceeded, or will likely exceed, the applicable time.]" is included at the bottom of several pages of the EAL matrix located in the proposed Beaver Valley Power Station Emergency Preparedness Plan, Section 4, "Emergency Conditions." An asterisk is used as a marker in the EALs which require use of the note. In this evaluation, for simplicity, the note is included as part of the EAL.

5.

Terms that are defined in Beaver Valley Power Station Emergency Preparedness Plan, Section 1, "Definitions," are indicated in the proposed BVPS-1 EALs as all capitals.

6.

Used the term none instead of not applicable in the FPB matrix.

These changes are considered administrative in nature and are neither a difference nor a deviation in accordance with RIS 2003-18, Supplement 2.

Page 4 of 48

RCS Barrier R

5 De

1. Critical Safety Function Status Loss Not Applicable Potential Loss A. RCS Integrity - Red Entry Conditions Met.

OR B.

Heat Sink - Red Entry Conditions Met.

RCI: Critical Safety Function Status Loss None Potential Loss

1. RCS Integrity-Orange entry conditions met.

OR

2.

a. Heat Sink-Red entry conditions met.

AND

b. Heat Sink is required.

Rev 5 Differences None Rev 5 Deviations The conditional statement "Heat Sink is required" was added as a condition to potential loss threshold #2.

6. Containment Radiation Monitoring RC2: Containment Radiation Monitoring Rev 5 Differences Loss Loss Removed the word "reading" for human factors considerations (minimize A. Containment radiation monitor reading greater than (site specific

1. Containment Radiation Monitor (RM-1RM-219A or B) > 8 R/hr (RC2 Line extraneous words).

value),

on Graph F-i).

Rev 5 Deviations Potential Loss None Not Applicable Potential Loss None

2. RCS Leak Rate RC5: RCS Leak Rate Rev 5 Differences Loss Loss (Potential Loss): Removed "with Letdown isolated" to simplify recognition A. RCS leak rate greater than available makeup capacity as indicated by

1. RCS leak rate greater than available makeup capacity as indicated by conditions.

a loss of RCS subcooling.

RCS subcooling < 18° F normal containment or < 33* F adverse (Potential Loss): Added "UNISOLABLE" to clarify that the intent is not to Potential Loss containment.

declare an emergency for a momentary leak that can be operationally isolated.

Potential Loss RevS5 Deviations A. RCS leak rate indicated greater than (site specific capacity of one charging pump in the normal charging mode) with Letdown isolated.

1. UNISOLABLE RCS leak exceeding the capacity of one charging pump None (129 gpm) in the normal charging mode.

4.

SG Tube Rupture RC6: SG Tube Leakage I Rupture Rev 5 Differences Loss Loss Added leakage to FPB category title to allow for consistent language with CT6 A. RUPTURED SG results in an ECCS (SI) actuation

1. RUPTURED SG results in an St actuation.

(NEI CT4) and fleet standardization.

Potential Loss Potential Loss Rev 5 Deviations Not Applicable None None

7. Other Site Specific Indications Note: BVPS-1 does not have any additional FPB thresholds in this category.

Loss Rev 5 Differences A. (site specific) as applicable.

N/A None Potential Loss Rev 5 Deviations A. (site specific) as applicable.

None Page 19 of 48

INI9

-0Re. 5 BP-.. EI Diffrec oDeiato AA.

Initiating Condition - ALERT Any release of gaseous or liquid radioactivity to the environment greater than 200 times the Radiological Effluent Technical Specifications/ODCM for 15 minutes or longer.

Operating Mode Applicability: All Example Emergency Action Level: (1 or 2 or 3 or 4 or 5)

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

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

(site specific monitor list and threshold values)

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

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

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

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

Normal can be considered as the highest reading in the past twenty-four RAI INITIATING CONDITION:

Any release of gaseous or liquid radioactivity to the environment greater than 200 times the ODCM limit for 15 minutes or longer.

Operating Mode Applicability: 1, 2, 3. 4, 5. 6, D EALs:

Note:

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

1. ANY of the following gaseous effluent monitors greater than the reading shown for 15 minutes or longer:

SLCRS Vent (RM-tVS-1i10 Ch 5)...................................

6.76E+05 cpm Ventilation Vent (RM-1VS-109 Ch 5)..............................

2.94E+05 cpm OR

2. ANY of the following liquid effluent monitors > 200 times the High-High alarm setpoint, not to exceed 8.5E+05 cpm, established by a current radioactivity discharge permit for 15 minutes or longer Liquid Waste Effluent Monitor (RM-1LW-104)

Laundry and Contaminated Shower Drains Monitor (RM-1LW-116)

OR

3. Confirmed sample analysis for gaseous or liquid releases > 200 times the ODCM limit for 15 minutes or longer.

Rev 5 Differences Removed "VALID" in accordance with NEI 99-01 Rev 5, FAQ# 4.

Established "NORMAL LEVELS" as a defined term in accordance with NEI 99-01 Rev 5, FAQ# 5.

Removed the words "reading' for human factors considerations (minimize extraneous words).

NEI AA1.4 is N/A for BVPS-1 because the plant is not equipped with a perimeter radiation monitoring system.

NEI AA1.5 is N/A for BVPS-1 because the plant is not equipped with a automatic real-time dose assessment system.

Rev 5 Deviations None or ex ungacui eavau.

I ___________________________I Page 24 of 48

I NEI 990 Rev 5 S...-1 E sDfeeceo eito CU2 Initiating Condition - NOTIFICATION OF UNUSUAL EVENT UNPLANNED loss of RCS/RPV inventory.

Operating Mode Applicability: Refueling Example Emergency Action Level: (1 or 2)

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

1.

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

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

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

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

cue INITIATING CONDITION:

UNPLANNED loss of RCS inventory.

Operating Mode Applicability: 6 EALs:

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

1. UNPLANNED RCS level drop as indicated by EITHER of the following:

Refueling Outage Temporary Level Instruments (LI-1RC-481C) < 97 inches (vessel flange) for 15 minutes or longer when the RCS level band is established above the vessel flange.

Note: Changes to the nesting format of EAL CU8.2 is considered administrative.

Rev 5 Differences Replaced 'unexplained" with "UNPLANNED" in accordance with NEI 99-01 Rev 5, FAQ# 10.

Rev 5 Deviations None 2.

OR

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

OR

a.

RCS level cannot be monitored.

AND

b.

Loss of RCS inventory as indicated by UNPLANNED level rise in Containment sumns or incore instrument sumo.

Page 47 of 48

I N

.EI.

9 Rev. 5 I*EA I

o r

Dev i

CA4 lnitiating Condition - ALERT Inability to maintain plant in cold shutdown.

Operating Mode Applicability: Cold Shutdown, Refueling Example Emergency Action Level: (1 or 2)

1.

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

CA10 INITATING CONDITION:

Inability to maintain plant in Cold Shutdown.

Operating Mode Applicability: 5, 6 EALs:

Note: Full inventory is pressurizer level 2> 22% actual with loop stops either isolated or unisolated.

1. RCS temperature > 200" F due to an UNPLANNED loss of decay heat removal capability for greater than the specified duration on Table C-2.

Rev 5 Differences Specified the UNPLANNED event in BVPS-1 CA10.1 as being due to an UNPLANNED loss of decay heat removal capability to be consistent with NEI CU4 IC in accordance with NEI 99-01 Rev 5 FAQ#13.

Added the site specific condition for full inventory to the site specific basis section and as a note in the EAL section.

Reworded CA10.2 in accordance with NEI 99-01 Rev 5, FAQ# 13.

Reworded the first RCS entry condition in Table C-2 to be a positive statement. Reworded the second condition for consistency with the wording in the first condition. These changes are considered administrative.

Rev 5 Deviations The conditional statement "RCS temperature cannot be monitored" was added as a condition to CA1 0.2. This is consistent with the basis of CA4 in the proposed NEI 99-01 Rev 6.

Table: RCS Reheat Duration Thresholds RCS Containment Closure Duration Intact (but not RCS N/A 60 minutes*

Reduced Inventory

[PWR])

Not Intact or RCS Established 20 minutes*

Reduced Inventory Not Established 0 minutes (PWR)

If an RCS heat removal system is in operation within this time frame and RCS temperature is being reduced, the EAL is not

__applicable.

Table C-2: RCS Reheat Duration Thresholds RCS CONTAINMENT Duration CLOSURE Intact with Full RCS N/A 60 minutes" Inventory Not Intact Established 20 minutes" OR Not Established 0 minutes Not Full RCS Inventory

• If an RCS heat removal system is in operation within this time frame and RCS temperature is being reduced, the EAL is not aoolicable.

2.

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

OR

2. a.

RCS temperature cannot be monitored.

AND

b.

RCS pressure rise > 10 psi due to an UNPLANNED loss of decay heat removal capability (this EAL does not apply in RCS solid plant conditions).

CU4 CUl0 Rev 5 Differences Initiating Condition - NOTIFICATION OF UNUSUAL EVENT INITIATING CONDITION:

Removed "with irradiated fuel in reactor vessel" from IC in accordance with UNPLANNED loss of decay heat removal capability with irradiated fuel in the UNPLANNED loss of decay heat removal capability.

NEI 99-01 Rev 5, FAC# 11.

RPV.

Operating Mode Applicability: 5, 6 Specified the UNPLANNED event in CU10.1 as being due to an UNPLANNED loss of decay heat removal capability to be consistent with NEI CU4 IC and the Operating Mode Applicability: Cold Shutdown, Refueling EALs:

change to BVPS-1 CA10.2 in accordance with NEI 99-01 Rev 5 FAQ#13.

Example Emergency Action Level: (1 or 2)

Note: The EMERGENCY DIRECTOR should not wait until the applicable time Rev 5 Deviations Note: The Emergency Director should not wait until the applicable time has has elapsed, but should declare the event as soon as it is determined that the elapsed, but should declare the event as soon as it is determined that the condition will likely exceed the applicable time.

None condition will likely exceed the applicable time.

1.

RCS temperature > 200" F due to an UNPLANNED loss of decay heat 1

UNPLANNED event results in RCS temperature exceeding the Technical removal capability.

Specification cold shutdown temperature limit.

OR

2.

Loss of all RCS temperatureand RCS/RPV level indication for 15 minutes

2.

Loss of ALL RCS temperature and RCS level indication for 15 minutes or or longer.

longer.

Page 48 of 48

Appendix 5 Beaver Valley Power Station Unit No. 2 EAL Evaluation

Beaver Valley Power Station Unit No. 2 EAL Evaluation DIFFERENCES - DEVIATIONS The items considered to be differences or deviations are based on the definitions provided in RIS 2003-18, Supplement 2. Any plant EAL [or Initiating Condition (IC) or Fission Product Barrier (FPB) threshold value] that does not meet the intent of the NEI 99-01, Revision 5 guidance or may result in an event being classified differently from the guidance is identified as a deviation and will be listed as such in this evaluation.

The basis section for each of the deviations documents the rationale for not adopting the NEI 99-01, Revision 5 guidance. Items identified as deviations will not be implemented without prior NRC review and approval.

ADMINISTRATIVE CHANGES The following changes apply throughout the set of EALs and are not specifically identified in the comparison tables:

1.

The NEI phrase "NOTIFICATION OF UNUSUAL EVENT" has been changed to "UNUSUAL EVENT" to sustain common terminology.

2.

The IC identification numbering has been modified to allow consistent grouping by event category.

3.

Numerical values, signs and key words of threshold values are bolded for emphasis.

4.

The NEI note "[T]he Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the

[condition will likely exceed the applicable time.] or [condition (duration) has exceeded, or will likely exceed, the applicable time.]" is included at the bottom of several pages of the EAL matrix located in the proposed Beaver Valley Power Station Emergency Preparedness Plan, Section 4, "Emergency Conditions." An asterisk is used as a marker in the EALs which require use of the note. In this evaluation, for simplicity, the note is included as part of the EAL.

5.

Terms that are defined in Beaver Valley Power Station Emergency Preparedness Plan, Section 1, "Definitions," are indicated in the proposed BVPS-2 EALs as all capitals.

6.

Used the term none instead of not applicable in the FPB matrix.

These changes are considered administrative in nature and are neither a difference nor a deviation in accordance with RIS 2003-18, Supplement 2.

Page 4 of 47

NEII91*PI E"

Di*ferencn,]*l Devi-ti RCS Barrier

1. Critical Safety Funct ion Status RCI:

Critical Safety Function Status Rev 5 Differences Loss Loss None Not Applicable None Rev 5 Deviations Potential Loss Potential Loss The conditional statement "Heat Sink is required" was added as a condition to A.

RCS Integrity - Red Entry Conditions Met.

1. RCS Integrity - Orange entry conditions met.

potential loss threshold #2.

OR OR B. Heat Sink - Red Entry Conditions Met.

2.

a.

Heat Sink - Red entry conditions met.

AND

b.

Heat Sink is required.

6. Containment Radiation Monitoring RC2:

Containment Radiation Monitoring Rev 5 Differences Loss Loss Removed the word "reading" for human factors considerations (minimize A.

Containment radiation monitor reading greater than (site specific

1. Containment Radiation Monitor (2RMR-RQ206 or 207) > 1.1E+01 R/hr extraneous words).

value).

(RC2 Line on Graph F-i).

Rev 5 Deviations Potential Loss Potential Loss None Not Applicable None

2. RCS Leak Rate RC5:

RCS Leak Rate Rev 5 Differences Loss Loss (Potential Loss): Removed "with Letdown isolated" to simplify recognition A. RCS leak rate greater than available makeup capacity as indicated by

1. RCS leak rate greater than available makeup capacity as indicated by conditions.

a loss of RCS subcooling.

RCS subcooling < 19° F normal containment or < 46° F adverse (Potential Loss): Added "UNISOLABLE" to clarify that the intent is not to Potential Loss containment.

declare an emergency for a momentary leak that can be operationally isolated.

A.

RCS leak rate indicated greater than (site specific capacity of one Potential Loss Rev 5 Deviations charging pump in the normal charging mode) with Letdown isolated.

1.

UNISOLABLE RCS leak exceeding the capacity of one charging pump None (130 gpm) in the normal charging mode.

4. SG Tube Rupture RC6:

SG Tube Leakage I Rupture Rev 5 Differences Loss Loss Added leakage to FPB category title to allow for consistent language with CT6 A.

RUPTURED SG results in an ECCS (SI) actuation.

1 RUPTURED SG results in an SI actuation.

(NEI CT4) and fleet standardization.

Potential Loss Potential Loss Rev 5 Deviations Not Applicable None None

7.

Other Site Specific Indications Note BVPS-2 does not have any additional FPB thresholds in this category.

Loss Rev 5 Differences A.

(site specific) as applicable.

N/A None Potential Loss Rev 5 Deviations A. (site specific) as applicable.

None Page 19 of 47

I NEI 990 Rev 5

.12EAsDfeeneo evain CA4 Initiating Condition - ALERT Inability to maintain plant in cold shutdown.

Operating Mode Applicability: Cold Shutdown, Refueling Example Emergency Action Level: (1 or 2)

1.

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

CAl0 INITIATING CONDITION:

Inability to maintain plant in Cold Shutdown.

Operating Mode Applicability: 5, 6 EALs:

Note: Full inventory is pressurizer level -> 22% actual with loop stops either isolated or unisolated.

1. RCS temperature> 200° F due to an UNPLANNED loss of decay heat removal capability for greater than the specified duration on Table C-2.

Rev 5 Differences Specified the UNPLANNED event in CAI0.1 as being due to an UNPLANNED loss of decay heat removal capability to be consistent with NEI CU4 IC and the change to CA10.2 in accordance with NEI 99-01 Rev 5 FAQ#13.

Added the site specific condition for full inventory to the site specific basis section and as a note in the EAL section.

Reworded CA10.2 in accordance with NEI 99-01 Rev 5, FAQ# 13.

Reworded the first RCS entry condition in Table C-2 to be a positive statement. Reworded the second condition for consistency with the wording in the first condition. These changes are considered administrative.

Rev 5 Deviations The conditional statement "RCS temperature cannot be monitored" was added as a condition to CA1 0.2. This is consistent with the basis of CA4 in the proposed NEI 99-01, Rev 6.

Table: RCS Reheat Duration Thresholds RCS Containment Closure Duration Intact (but not RCS N/A 60 minutes*

Reduced Inventory

[PWR])

Not Intact or RCS Established 20 minutes*

Reduced Inventory Not Established 0 minutes (PWR)

If an RCS heat removal system is in operation within this time frame and RCS temperature is being reduced, the EAL is not annlicable.

Table C-2: RCS Reheat Duration Thresholds RCS CONTAINMENT Duration CLOSURE Intact with Full RCS N/A 60 minutes*

Inventory Not Intact Established 20 minutes*

OR Not Established 0 minutes Not Full RCS Inventory L

If an RCS heat removal system is in operation within this time frame and RCS temperature is being reduced, the EAL is not applicable.

• rr

2.

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

OR

2.

a.

RCS temperature cannot be monitored.

AND

b.

RCS pressure rise > 10 psi due to an UNPLANNED loss of decay heat removal capability (this EAL does not apply in RCS solid plant I

conitionsi.b I

CU4 Initiating Condition - NOTIFICATION OF UNUSUAL EVENT UNPLANNED loss of decay heat removal capability with irradiated fuel in the RPV.

Operating Mode Applicability: Cold Shutdown, Refueling Example Emergency Action Level: (1 or 2)

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

1.

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

2.

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

CU10 INITIATING CONDITION:

UNPLANNED Loss of decay heat removal capability.

Operating Mode Applicability: 5, 6 EALs:

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

1. RCS temperature > 200° F due to an UNPLANNED loss of decay heat removal capability.

OR

2.

Loss of ALL RCS temperature and RCS level indication for 15 minutes*

or Ionaer.

Rev 5 Differences Removed "with irradiated fuel in reactor vessel" from IC in accordance with NEI 99-01 Rev 5, FAQ#1 1.

Specified the UNPLANNED event in CU10.1 as being due to an UNPLANNED loss of decay heat removal capability to be consistent with NEI CU4 IC and the change to BVPS-2 CA10.2 in accordance with NEI 99-01 Rev 5 FAQ#13.

Rev 5 Deviations None Page 47 of 47

Appendix 6 Proposed Beaver Valley Power Station, Unit Nos. 1 and 2 EAL Wallboards

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