Text

Response to Request for Additional Information on License Amendment Request to Adopt Emergency Action Level Schemes Pursuant to NEI 99-01, Revision 6, Development of Emergency Action Levels for Non-Passive Reactors
	ML15159B025
Person / Time
Site:	Dresden, Peach Bottom, Oyster Creek, Limerick, Clinton, Quad Cities, LaSalle, Crane
Issue date:	06/05/2015
From:	David Helker; Exelon Generation Co
To:	; Document Control Desk, Office of Nuclear Material Safety and Safeguards, Office of Nuclear Reactor Regulation
References
	RA-15-048, RS-15-167, TMl-15-070
	Download: ML15159B025 (68)
	v • d • e

200 E"xelon Way Exelon Generation Kennett Square. PA 19348 www. -ix.:1011corr>.c c 111 RS-15-167 RA-15-048 TMl-15-070 June 5, 2015 10 CFR 50.90 10 CFR 50, Appendix E 10 CFR 50.4 U.S. Nuclear Regulatory Commission A TIN: Document Control Desk Washington, DC 20555 Clinton Power Station, Unit 1 Facility Operating License No. NPF-62 NRC Docket No. 50-461 Dresden Nuclear Power Station, Units 1, 2 and 3 Facility Operating License No. DPR-2 Renewed Facility Operating License Nos. DPR-19 and DPR-25 NRC Docket Nos. 50-237. 50-249. and 72-37 LaSalle County Station, Units 1 and 2 Facility Operating License Nos. NPF-11 and NPF-18 NRC Docket Nos. 50-373. 50-374. 72-70 Limerick Generating Station, Units 1 and 2 Renewed Facility Operating License Nos. NPF-39 and NPF-85 NRC Docket Nos. 50-352. 50-353. and 72-65 Oyster Creek Nuclear Generating Station Renewed Facility Operating License No. DPR-16 NRC Docket Nos. 50-219 and 72-15 Peach Bottom Atomic Power Station, Units 1, 2 and 3 Facility Operating License No. DPR-12 Renewed Facility Operating License Nos. DPR-44 and DPR-56 NRC Docket Nos. 50-171. 50-277. 50-278 and 72-29

U.S. Nuclear Regulatory Commission Response to Request for Additional Information License Amendment Request Adoption of NEI 99-01, Revision 6 EAL Schemes June 5, 2015 Page 2

Subject:

Quad Cities Nuclear Power Station, Units 1 and 2 Renewed Facility Operating License Nos. DPR-29 and DPR-30 NRG Docket Nos. 50-254, 50-265, and 72-53 Three Mile Island Nuclear Station, Unit 2 Facility Possession-Only License No. DPR-73 NRG Docket No. 50-320 Response to Request for Additional Information License Amendment Request to Adopt Emergency Action Level Schemes Pursuant to NEI 99-01, Revision 6, "Development of Emergency Action Levels for Non-Passive Reactors"

References:

1)

Letter from James Barstow (Exelon Generation Company, LLC) to U.S. Nuclear Regulatory Commission - License Amendment Request to Adopt Emergency Action Level Schemes Pursuant to NEI 99-01, Revision 6, "Development of Emergency Action Levels for Non-Passive Reactors," dated May 30, 2014 (ML14164A054)

2)

Electronic Mail Request from Joel Wiebe (U.S. Nuclear Regulatory Commission) to Richard Gropp (Exelon Generation Company, LLC)

- Preliminary Request for Information Regarding Emergency Action Level Fleet Amendments, dated November 11, 2014

3)

Letter from James Barstow (Exelon Generation Company, LLC) to U.S. Nuclear Regulatory Commission - Response to Request for Additional Information - License Amendment Request to Adopt Emergency Action Level Schemes Pursuant to NEI 99-01, Revision 6, "Development of Emergency Action Levels for Non-Passive Reactors," dated March 2, 2015 (ML15071A118)

4)

Electronic Mail Request from Blake Purnell (U.S. Nuclear Regulatory Commission) to Richard Gropp (Exelon Generation Company, LLC) - Preliminary Request for Information Regarding Emergency Action Level Fleet Amendments, dated May 19, 2015 By letter dated May 30, 2014 (Reference 1), Exelon Generation Company, LLC (Exelon) submitted a License Amendment Request (LAR) to support changes to Emergency Plans.

Specifically, the proposed changes involve revising the Emergency Plans for the affected facilities to adopt the Nuclear Energy lnstitute's (NEl's) revised Emergency Action Level (EAL) schemes described in NEI 99-01, Revision 6, "Development of Emergency Action Levels for Non-Passive Reactors," which have been endorsed by the NRG as documented in an NRG letter dated March 28, 2013.

U.S. Nuclear Regulatory Commission Response to Request for Additional Information License Amendment Request Adoption of NEI 99-01, Revision 6 EAL Schemes June 5, 2015 Page 3 Appendix E,Section IV.B.2, of 1 O CFR 50 stipulates that a licensee desiring to change its entire EAL scheme shall submit an application for an amendment to its license and receive NRC approval before implementing the change. Exelon's currently approved Emergency Plan EAL schemes are based on the guidance established in NEI 99-01, Revision 5, "Methodology for Development of Emergency Action Levels. " Exelon is proposing to adopt the EAL schemes based on the latest NRG-endorsed guidance, which is described in NEI 99-01, Revision 6.

In a U.S. Nuclear Regulatory Commission (NRC) electronic mail message dated November 4, 2014 (Reference 2), the NRC indicated that it had reviewed the information submitted in the Reference 1 letter pertaining to the proposed LAR and requested additional clarifying information to support its continued review. The Reference 2 electronic mail message contained a number of draft NRC questions, which were further discussed during a December 4, 2014, teleconference between Exelon and NRC representatives.

By letter dated March 2, 2015 (Reference 3), Exelon provided a response to the NRC's request for additional information described in the Reference 2 electronic mail message.

Subsequently, in an electronic mail message dated May 19, 2015 (Reference 4), the NRC issued an additional request for information. The Reference 4 electronic mail message identified five additional questions pertaining to the proposed EALs for Clinton Power Station, Dresden Nuclear Power Station, and Quad Cities Nuclear Power Station, which were the subject of further discussions with the NRC during a teleconference on May 26, 2015. As a result of the discussions, the NRC is requesting that Exelon provide additional clarifying information in support of the continued review of the Reference 1 LAR.

Accordingly, Attachment 1 provides Exelon's response to the request for information contained in the Reference 4 electronic mail message. Attachment 1 also includes additional information supporting minor clarifying changes for other EALs as a result of the discussions with the NRC. Attachments 2 through 8 include revised clean pages for the affected EAL Basis documents and other supporting information for the plants cited.

Exelon has reviewed the information supporting a finding of No Significant Hazards Consideration and the Environmental Consideration provided to the NRC in the Reference 1 submittal. The additional information provided in this submittal does not affect the bases for concluding that the proposed license amendment does not involve a significant hazards consideration. Furthermore, the additional information provided in this submittal does not affect the bases for concluding that neither an environmental impact statement nor an environmental assessment needs to be prepared in connection with the proposed amendment.

U.S. Nuclear Regulatory Commission Response to Request for Additional Information License Amendment Request Adoption of NEI 99-01, Revision 6 EAL Schemes June 5, 2015 Page4 There are no regulatory commitments contained in this submittal.

If you have any questions concerning this submittal, please contact Richard Gropp at (610) 765-5557.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the 5th day of June 2015.

Respectfully, David P. Helker Manager, Licensing and Regulatory Affairs Exelon Generation Company, LLC Attachments: -

Response to the Request for Additional Information - License Amendment Request to Adopt Emergency Action Level Schemes Pursuant to NEI 99-01, Revision 6, "Development of Emergency Action Levels for Non-Passive Reactors" -

Revised Radiological Emergency Plan Annex Information for Clinton Power Station

A - Revised EAL Comparison Matrix Document

B - Revised EAL Basis Documents -

Revised Radiological Emergency Plan Annex Information for Dresden Station

A - Revised EAL Comparison Matrix Document

C - Revised EAL Basis Documents -

Revised Radiological Emergency Plan Annex Information for LaSalle County Station

A - Revised EAL Basis Document -

Revised Radiological Emergency Plan Annex Information for Limerick Generating Station Enclosure SA - Revised EAL Basis Document -

Revised Radiological Emergency Plan Annex Information for Oyster Creek Nuclear Generating Station

A - Revised EAL Basis Document

U.S. Nuclear Regulatory Commission Response to Request for Additional Information License Amendment Request Adoption of NEI 99-01, Revision 6 EAL Schemes June 5, 2015 Page 5 Attachments (continued): -

Revised Radiological Emergency Plan Annex Information for Peach Bottom Atomic Power Station

A - Revised EAL Basis Document -

Revised Radiological Emergency Plan Annex Information for Quad Cities Nuclear Station

A - Revised EAL Basis Documents cc:

w/ Attachment 1 only Regional Administrator - NRC Region I Regional Administrator-NRC Region Ill NRC Senior Resident Inspector-Clinton Power Station NRC Senior Resident Inspector - Dresden Nuclear Power Station NRC Senior Resident Inspector - LaSalle County Station NRC Senior Resident Inspector - Limerick Generating Station NRC Senior Resident Inspector - Oyster Creek Nuclear Generating Station NRC Senior Resident Inspector-Peach Bottom Atomic Power Station NRC Senior Resident Inspector-Quad Cities Nuclear Power Station NRC Senior Resident Inspector - Three Mile Island Nuclear Station NRC Project Manager, NRR-Exelon Fleet NRC Project Manager, NRR - Clinton Power Station NRC Project Manager, NRR - Dresden Nuclear Power Station NRC Project Manager, NRR - LaSalle County Station NRC Project Manager, NRR - Limerick Generating Station NRC Project Manager, NRR - Oyster Creek Nuclear Generating Station NRC Project Manager, NRR - Peach Bottom Atomic Power Station NRC Project Manager, NRR - Quad Cities Nuclear Power Station NRC Project Manager, NRR - Three Mile Island Nuclear Station R. R. Janati, Commonwealth of Pennsylvania S. Gray, State of Maryland Director, Bureau of Nuclear Engineering, New Jersey Department of Environmental Protection Illinois Emergency Management Agency Division of Nuclear Safety Chairman, Board of County Commissioners of Dauphin County, PA Chairman, Board of Supervisors of Londonderry Township, PA Mayor of Lacey Township, Forked River, NJ

ATTACHMENT 1 Response to the Request for Additional Information -

License Amendment Request to Adopt Emergency Action Level Schemes Pursuant to NEI 99-01, Revision 6, "Development of Emergency Action Levels for Non-Passive Reactors" Response to Request For Additional Information License Amendment Request Adoption of NEI 99-01, Revision 6 EAL Schemes Page 1of6 Response to Request for Additional Information License Amendment Request to Adopt Emergency Action Level Schemes Pursuant to NEI 99-01. Revision 6, "Development of Emergencv Action Levels for Non-Passive Reactors" By letter dated May 30, 2014 (Reference 1 ), Exelon Generation Company, LLC (Exelon) submitted a license amendment request to support changes to Emergency Plans. Specifically, the proposed changes involve revising the Emergency Plans for the affected facilities to adopt the Nuclear Energy lnstitute's (NEl's) revised Emergency Action Level (EAL) schemes described in NEI 99-01, Revision 6, "Development of Emergency Action Levels for Non-Passive Reactors," which have been endorsed by the NRC as documented in an NRC letter dated March 28, 2013 (Reference 5).

Appendix E,Section IV.B.2, of 1 O CFR 50 stipulates that a licensee desiring to change its entire EAL scheme shall submit an application for an amendment to its license and receive NRC approval before implementing the change. Exelon's currently approved Emergency Plan EAL schemes are based on the guidance established in NEI 99-01, Revision 5, "Methodology for Development of Emergency Action Levels. " Exelon is proposing to adopt the EAL schemes based on the latest NRG-endorsed guidance, which is described in NEI 99-01, Revision 6.

In a U.S. Nuclear Regulatory Commission (NRC) electronic mail message dated November 4, 2014 (Reference 2), the NRC indicated that it had reviewed the information submitted in the Reference 1 letter pertaining to the proposed license amendment and requested additional clarifying information to support its continued review. The Reference 2 electronic mail message contained a number of draft NRC questions, which were further discussed during a December 4, 2014, teleconference between Exelon and NRC representatives.

By letter dated March 2, 2015 (Reference 3), Exelon provided a response to the NRC's request for additional information described in the Reference 2 electronic mail message.

Subsequently, in an electronic mail message dated May 19, 2015 (Reference 4), the NRC issued an additional request for additional information. The Reference 4 electronic mail message identified five additional questions pertaining to the proposed EALs for Clinton Power Station, Dresden Nuclear Power Station, and Quad Cities Nuclear Power Station, which were the subject of further discussions with the NRC during a teleconference on May 26, 2015. Exelon is also providing additional information supporting minor clarifying changes for other EALs as a result of the discussions with the NRC. The questions in the Reference 4 electronic mail message are identified below followed by Exelon's response.

Response to Request For Additional Information License Amendment Request Adoption of NEI 99-01, Revision 6 EAL Schemes Page 2 of 6 I.

RESPONSE TO NRC QUESTIONS Clinton Power Station Question 1

1. For threshold CT6 (RC4), using surveys to determine Max Safe (Max Normal) radiation levels is inconsistent with our expectations for timely EAL criteria. Please develop criteria that would be indicative of CT6 (RC4) rather than using surveys.

Response

With regard to the threshold for EAL RC4 for Clinton Power Station, there are two areas that have radiation monitoring instrumentation capable of determining Max Norm values. As a result, the Basis for EAL RC4 has been revised to list the instrumentation separately in the threshold. Further, since all of the Max Safe Radiation values were determined by radiation survey, the Basis for EAL CT6 has been revised to only include Max Safe Temperature values as they are all instrumented monitoring points. In addition, the Justification column in the Comparison Matrix document has also been revised to reflect the changes. The revised pages have been included in Attachment 2 of this submittal.

Dresden Nuclear Power Station Question 1

1.

Their response to RA/ #16 states that the SBO DGs are "... not normally connected to any bus," and as such should not be used for EA Ls MG 1, MS 1, MA 1, MG2, CA 1, and CU1 as it is not timely enough to credit as a source unless justification is provided that documents that these can be operable within a timely period.

Response

Exelon agrees that the Station Blackout (SBO) Diesel Generator (DG) should not be included in the list of available power sources and has removed the reference to the SBO DG from the Basis documents for EALs MG 1, MS 1, MA 1, MG2, CA 1 and CU 1 for Dresden Nuclear Power Station. The revised pages have been included in Attachment 3 of this submittal.

Question 2

2. For threshold CT6, using surveys to determine Max Safe radiation levels is inconsistent with our expectations for timely EAL criteria. Please develop criteria that would be indicative of CT6 rather than using surveys.

Response to Request For Additional Information License Amendment Request Adoption of NEI 99-01, Revision 6 EAL Schemes Page 3 of 6

Response

Since all the Max Safe Radiation values were determined by radiation surveys, the Basis for EAL CT6 has been revised for Dresden Nuclear Power Station to only include Max Safe Temperature values as they are all instrumented monitoring points. In addition, the Justification column in the Comparison Matrix document has also been revised to reflect the changes. The revised pages have been included in Attachment 3 of this submittal.

Quad Cities Nuclear Power Station Question 1

1. Their response to RA/ #16 states that their SBO DG will be "... available within one hour of the onset of an SBO event, " and as such should not be used for EA Ls MG 1, MS 1, MA 1, MG2, CA 1, and CU1 as it is not timely enough to credit as a source. Please justify further, or revise accordingly.

Response

Exelon agrees that the SBO DG should not be included in the list of available power sources and has removed the reference to the SBO DG from the Basis documents for EALs MG1, MS1, MA1, MG2, CA1 and CU1 for Quad Cities Nuclear Power Station. The revised pages have been included in Attachment 8 of this submittal.

Question 2

2. For threshold CT6, using surveys to determine Max Safe radiation levels is inconsistent with our expectations for timely EAL criteria. Please develop criteria that would be indicative of CT6 rather than using surveys.

Response

Quad Cities Nuclear Power Station does not rely on radiological surveys for any Max Safe Radiation levels, and therefore, no change was made to threshold EAL CT6.

II. ADDITIONAL CHANGES AND CLARIFICATIONS Additional Proposed Minor Changes

1. The following wording has been deleted from the Basis section for EAL RC3 for the Exelon Boiling Water Reactor (BWR) sites, which includes Clinton Power Station, Dresden Nuclear Power Station, LaSalle County Station, Limerick Generating Station, Oyster Creek Nuclear Generating Station, Peach Bottom Atomic Power Station, and Quad Cities Nuclear Power Station. These minor changes were made to the Basis for this particular EAL, since the statements were not necessarily applicable to all the affected BWR sites and in order to maintain consistency with the NRG-endorsed NEI 99-01, Revision 6 guidance.

Response to Request For Additional Information License Amendment Request Adoption of NEI 99-01, Revision 6 EAL Schemes Page 4 of 6 "The release of mass from the RCS due to the as-designed/expected operation of any relief valve does not warrant an emergency classification.

A stuck-open Safety Relief Valve (SRV) or SRV leakage is not considered either identified or unidentified leakage by Technical Specification and, therefore, is not applicable to this EAL. "

The revised RC3 Basis document for each affected site has been included in Attachments 2 through 8 of this submittal.

2. A value in the Basis documents for EALs MG1 and MS3 for Quad Cities Nuclear Power Station was revised. Specifically, the value for Minimum Steam Cooling Water Level was changed from -166 inches to -190 inches. This change reflects the current (in use)

EAL thresholds, which were revised since the Reference 1 submittal to reflect changes in site fuel type use from GE14 to OPTIMA 2 fuel.

The revised MG1 and MS3 Basis documents for Quad Cities Nuclear Power Plant are included in Attachment 8 of this submittal.

3. An entry in the Communications Capability table in the Basis documents for EALs MU?

and CU4 for Clinton Power Station was modified as follows:

All Telephone Lines (Commercial and Microwave)

Modified to read:

All Telephone Lines This change reflects the present (in use) EAL thresholds, which were revised since the Reference 1 submittal to reflect changes in the site's communication capability related to eliminating the use of microwave communication under the EALs.

The revised MU? and CU4 Basis documents for Clinton Power Station are included in of this submittal.

Proposed Implementation Schedule In addition, the Reference 1 letter includes the following statement regarding implementation of the revised EAL schemes:

Exelon requests approval of the proposed changes by May 30, 2015, with a targeted implementation period for the plants within 90 days pending outage schedules and/or required training cycles.

In order for Exelon to effectively implement the EAL changes at the ten stations identified in the Reference 1 letter sufficient time will be needed. Training must be completed for the Operations' crews and other Emergency Response Organization (ERO) personnel at each of the sites. This training needs to be coordinated with the established training cycles for Response to Request For Additional Information License Amendment Request Adoption of NEI 99-01, Revision 6 EAL Schemes Page 5 of 6 the Operations' crews. In addition, training cycles for Operations and other ERO personnel are scheduled around plant outages in order to avoid potential impact on staffing activities and resources.

Therefore, Exelon proposes to implement the EAL changes on a plant-by-plant basis for those plants listed in the Reference 1 submittal within 90 days after completion of required training and in coordination with plant outage schedules.

Permanently Shutdown Plants Furthermore, the Reference 1 letter also includes references to the following facilities that are in a permanently shutdown condition:

o Dresden Nuclear Power Station, Unit 1 Facility Operating License No. DPR-2 o

Peach Bottom Atomic Power Station, Unit 1 Facility Operating License DPR-12 o Three Mile Island Nuclear Station, Unit 2 Facility Possession-Only License No. DPR-73 These facilities were included, since the respective Emergency Plans for the three sites incorporate these shutdown units within the scope of their Plans as the excerpts below indicate.

Dresden - Emergency Plan (EP-AA-1004)

1. 1 Facility Description Dresden Station, Units 1, 2 and 3, is located in the Goose Lake Township of Grundy County in northeastern Illinois. Unit 1 is in permanent shutdown (see Figure 1-1 ).

The plant consists of three Boiling Water Reactor (BWR) Nuclear Steam Supply Systems (NSSS) and turbine generators provided by General Electric Company.

Unit 1 is a dual cycle boiling water reactor designed for a power output of 700 MWt and has officially been retired as of August 31, 1984. Units 2 and 3 are equipped with nuclear steam supply systems (NSSS) designed for a power output of 2957 MWt.

Peach Bottom - Emergency Plan (EP-AA-1007)

1. 1 Facility Description The Peach Bottom Atomic Power Station (PBAPS) is a fixed nuclear facility operated by Exelon Nuclear. The station consists of one High Temperature Gas Cooled Reactor designated as Unit 1, which is in the SAFSTOR status of decommissioning, two operating Boiling Water Reactors designated as Units 2 and 3, and an Independent Spent Fuel Storage Installation (ISFSI).

Response to Request For Additional Information License Amendment Request Adoption of NEI 99-01, Revision 6 EAL Schemes Page 6 of 6 The PBAPS station is located partly in York County and partly in Lancaster County in southeastern Pennsylvania, on the west shore of Conowingo Pond, near the mouth of Rock Run Creek. The plant is about 38 miles NNE of Baltimore, MD; 65 miles WSW of Philadelphia, PA; 45 miles SE of Harrisburg, PA; and 20 miles SSE of Lancaster, PA.

Conowingo Pond is a reservoir formed by the backwater of Conowingo Dam on the Susquehanna River; the dam is located about 9 miles downstream from PBAPS. The nearest communities are Delta, PA, and Cardiff, MD, located approximately 4 and 6 miles WSW of the site, respectively.

Three Mile Island - Emergency Plan (EP-AA-1009)

1. 1 Facility Description TM/ Unit 1 is operated by Exelon Nuclear. The TM/ Unit #1 is an 870 Mwe, pressurized water-type, nuclear steam supply system supplied by Babcock & Wilcox Company.

TM/ Unit 2 is owned by First Energy Corporation. The TM/ Unit 2 reactor was damaged during an accident in 1979 and is currently defue/ed and the plant maintained in long-term monitored storage. Monitoring of this facility is performed by Exelon Nuclear through a service agreement with First Energy Corporation. The arrangement of the major TM/-1 and TM/-2 facilities is shown in Figures TM/ 1-1 and TM/ 1-2.

References

1)

Letter from James Barstow (Exelon Generation Company, LLC) to U.S. Nuclear Regulatory Commission - License Amendment Request to Adopt Emergency Action Level Schemes Pursuant to NEI 99-01, Revision 6, "Development of Emergency Action Levels for Non-Passive Reactors," dated May 30, 2014 (ML14164A054)

2)

Electronic Mail Request from Joel Wiebe (U.S. Nuclear Regulatory Commission) to Richard Gropp (Exelon Generation Company, LLC) - Preliminary Request for Information Regarding Emergency Action Level Fleet Amendments, dated November 11, 2014

3)

Letter from James Barstow (Exelon Generation Company, LLC) to U.S. Nuclear Regulatory Commission - Response to Request for Additional Information - License Amendment Request to Adopt Emergency Action Level Schemes Pursuant to NEI 99-01, Revision 6, "Development of Emergency Action Levels for Non-Passive Reactors," dated March 2, 2015 (ML15071A118)

4)

Electronic Mail Request from Blake Purnell (U.S. Nuclear Regulatory Commission) to Richard Gropp (Exelon Generation Company, LLC) - Preliminary Request for Information Regarding Emergency Action Level Fleet Amendments, dated May 19, 2015

5)

Letter from Mark Thaggard (U.S. Nuclear Regulatory Commission) to Susan Perkins-Grew (Nuclear Energy Institute) - U.S. Nuclear Regulatory Commission Review and Endorsement of NEI 99-01, Revision 6, November 2012, dated March 28, 2013

ATTACHMENT 2 REVISED RADIOLOGICAL EMERGENCY PLAN ANNEX INFORMATION FOR CLINTON POWER STATION EP-AA-1003 Enclosures

A - Revised EAL Comparison Matrix Document

8 - Revised EAL Basis Documents

Clinton - Revised Comparison Matrix NEI 99-01 Rev 6 Category: Reactor Coolant System Barrier RCS Leak Rate Operating Mode Applicability:

Power Operation, Startup, Hot Standby, Hot Shutdown Fission Product Barrier Threshold:

Loss RC3 A. UNISOLABLE break in any of the following: (site-specific systems with potential for high-energy line breaks)

OR B. Emergency RPV Depressurization Potential Loss A. UNISOLABLE primary system leakage that results in exceeding EITHER of the following:

1. Max Normal Operating Temperature.

OR

2. Max Normal Operating Area Radiation Level.

Proposed EAL Category: Reactor Coolant System Barrier RCS Leak Rate Operating Mode Applicability:

1, 2, 3 Fission Product Barrier (FPB) Threshold:

Loss

1. UNISOLABLE Main Steam Line (MSL), Feedwater,, RWCU or RCIC line break.

OR

2. Emergency RPV Depressurization is required.

Potential Loss RC4

3. UNISOLABLE primary system leakage that results in EITHER of the following:

a. Secondary Containment area temperature > EOP-8 Maximum Normal operating levels.

OR

b. Secondary Containment area radiation level > EOP-8 Maximum Normal operating levels in any of the following areas:

Spent Fuel Storage Area New Fuel Storage Vault Fuel Building Handling Platform Page 1of2 Justification D NoChange 0 Difference D Deviation

1) Listed site-specific systems and threshold values to ensure timely classification.

2) Due to the ability to declare in a timely manner Potential Loss 3.b only lists areas with installed radiation monitoring instruments, and not areas requiring a survey to determine max norm values. The areas requiring a rad survey are bound by potential loss 3.a max normal temp instrumentation, and CT6.1 UNISOLABLE direct downstream pathway to the environment exists after primary containment isolation signal.

Clinton - Revised Comparison Matrix NEI 99-01 Rev 6 Category: Containment Barrier Primary Containment Isolation Failure Operating Mode Applicability:

Power Operation, Startup, Hot Standby, Hot Shutdown Fission Product Barrier Threshold:

Loss CT3 A. UNISOLABLE direct downstream pathway to the environment exists after primary containment isolation signal OR B. Intentional primary containment venting per EOPs OR C. UNISOLABLE primary system leakage that results in exceeding EITHER of the following:

1. Max Safe Operating Temperature.

OR

2. Max Safe Operating Area Radiation Level.

Proposed EAL Category: Containment Barrier Primary Containment Isolation Failure Operating Mode Applicability:

1, 2, 3 Fission Product Barrier (FPB) Threshold:

Loss

1. UNISOLABLE direct downstream pathway to the environment exists after primary containment isolation signal.

OR CT6

2. Intentional Primary Containment venting/purging per EOPs or SAMGs due to accident conditions.

OR

3. UNISOLABLE primary system leakage that results in Secondary Containment area temperature > EOP-8 Maximum Safe operating levels.

Page 2 of 2 Justification D NoChange 0 Difference D Deviation

1) Listed site-specific threshold values to ensure timely classification.

2) Due to the ability to declare in a timely manner Loss 3 only lists areas with installed monitoring instruments, and not areas requiring a survey to determine max safe values. The areas requiring a radiation survey are bound by loss 3 max safe temp instrumentation, and loss 1 UNISOLABLE direct downstream pathway to the environment exists after primary containment isolation signal.

Clinton Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION Initiating Condition:

Primary Containment Conditions Operating Mode Applicability:

1, 2, 3 Fission Product Barrier (FPB} Threshold:

LOSS

1. Drywall pressure >1.68 psig.

AND

2. Drywall pressure rise is due to RCS leakage Basis: _

RC3 The> 1.68 psig pressure is the Drywall high pressure setpoint which indicates a LOCA by automatically initiating ECCS.

The second threshold condition focuses the fission product barrier loss threshold on a failure of the RCS instead of the non-LOCA malfunctions that may adversely affect Drywall pressure. Pressures of this magnitude can be caused by non-LOCA events such as a loss of Drywall cooling or inability to control Drywall vent/purge.

There is no RCS Potential Loss threshold associated with Primary Containment Pressure.

Basis Reference(s}:

1.

NEI 99-01 Rev 6, Table 9-F-2

2.

4401.01, EOP-1 RPV Control

3.

4402.01, EOP-6 Primary Containment Control

4.

Clinton Power Station EOP Technical Bases

5.

Technical Specifications Table 3.3.1.1-1

6.

Technical Specifications Table 3.3.5.1-1

7.

4001.01, Reactor Coolant Leakage

Clinton Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION Initiating Condition:

RCS Leak Rate Operating Mode Applicability:

1, 2, 3 Fission Product Barrier (FPB) Threshold:

LOSS RC4

1. UNISOLABLE Main Steam Line (MSL), Feedwater, RWCU, or RCIC line break.

OR

2. Emergency RPV Depressurization is required.

POTENTIAL LOSS

3. UNISOLABLE primary system leakage that results in EITHER of the following:

a. Secondary Containment area temperature > EOP-8 Maximum Normal operating levels.

OR

b. Secondary Containment area radiation level > EOP-8 Maximum Normal operating levels in any of the following areas:

Spent Fuel Storage Area-1 RIX-AR016 New Fuel Storage Vault -1 RIX-AR019/AR052 Basis:

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

Failure to isolate the leak, within 15 minutes or if known that the leak cannot be isolated within 15 minutes, from the start of the leak requires immediate classification.

Classification of a system break over system leakage is based on information available to the Control Room from the event. Indications that should be considered are:

Reports describing magnitude of steam or water release.

Use of system high flow alarms I indications, if available, Significant changes in makeup requirements, Abnormal reactor water level changes in response to the event.

The use of the above indications provides the Control Room the bases to determine that the on going event is more significant than the indications that would be expected from system leakage and therefore should be considered a system break.

Clinton Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION RC4 (cont)

Basis (cont):

Loss Threshold #1 Basis Large high-energy lines that rupture outside primary containment can discharge significant amounts of inventory and jeopardize the pressure-retaining capability of the RCS until they are isolated. If it is determined that the ruptured line cannot be promptly isolated, the RCS barrier Loss threshold is met.

Loss Threshold #2 Basis Emergency RPV Depressurization in accordance with the EOPs is indicative of a loss of the RCS barrier. If Emergency RPV Depressurization is performed, the plant operators are directed to open safety relief valves (SRVs) and keep them open. Even though the RCS is being vented into the suppression pool, a Loss of the RCS barrier exists due to the diminished effectiveness of the RCS to retain fission products within its boundary.

Potential Loss Threshold #3 Basis Potential loss of RCS based on primary system leakage outside the primary containment is determined from EOP temperature or radiation Max Normal Operating values in areas such as main steam line tunnel, RCIC, etc., which indicate a direct path from the RCS to areas outside primary containment.

A Max Normal Operating value is the highest value of the identified parameter expected to occur during normal plant operating conditions with all directly associated support and control systems functioning properly.

The indicators reaching the threshold barriers and confirmed to be caused by RCS leakage from a primary system warrant an Alert classification. A primary system is defined to be the pipes, valves, and other equipment which connect directly to the RPV such that a reduction in RPV pressure will effect a decrease in the steam or water being discharged through an unisolated break in the system.

In general, multiple indications should be used to determine if a primary system is discharging outside Primary Containment. For example, a high area radiation condition does not necessarily indicate that a primary system is discharging outside Primary Containment since this may be caused by radiation shine from nearby steam lines or the movement of radioactive materials. Conversely, a high area radiation condition in conjunction with other indications (e.g. room flooding, high area temperatures, reports of steam in Secondary Containment, an unexpected rise in Feedwater flowrate, or unexpected Main Turbine Control Valve closure) may indicate that a primary system is discharging outside Primary Containment. Max Norm Radiation values are obtained by survey through the implementation of the Emergency Operating Procedures for determining UNISOLABLE RCS leakage. These values should only be used for determination of the EAL if they are available in a timely manner.

An UNISOLABLE leak which is indicated by Max Normal Operating values escalates to a Site Area Emergency when combined with Containment Barrier CT6 Loss Threshold

1 (after a containment isolation) and a General Emergency when the Fuel Clad Barrier criteria is also exceeded.

Clinton Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION RC4 (cont)

Basis Reference(s):

1.

NEI 99-01 Rev 6, Table 9-F-2

2.

M05-1002, Main steam

3.

USAR 5.2.5

4.

USAR Tables 5.2-9a and 5.2-9b

5.

9043.06, Drywell Floor Drain Sump Flow Test OOPS404

6.

9443.01, Drywell Equipment Drain Sump Flow E31-N578 Channel Cal 01 PS274

7.

4406.01, EOP-8 Secondary Containment Control

8.

Clinton Power Station Emergency Operating Procedures Technical Bases, Section 1 O

9.

USAR Figure 6.2-132

Clinton Annex RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION Initiating Condition:

Primary Containment Isolation Failure Operating Mode Applicability:

1, 2, 3 Fission Product Barrier (FPB) Threshold:

LOSS Exelon Nuclear CT6

1. UNISOLABLE direct downstream pathway to the environment exists after primary containment isolation signal.

OR

2. Intentional Primary Containment venting/purging per EOPs or SAGs due to accident conditions.

OR

3. UNISOLABLE primary system leakage that results in Secondary Containment area temperature > EOP-8, Maximum Safe operating levels.

Basis:

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

Failure to isolate the leak, within 15 minutes or if known that the leak cannot be isolated within 15 minutes, from the start of the leak requires immediate classification.

These thresholds address incomplete containment isolation that allows an UNISOLABLE direct release to the environment.

Loss Threshold #1 Basis The use of the modifier "direct" in defining the release path discriminates against release paths through interfacing liquid systems or minor release pathways, such as instrument lines, not protected by the Primary Containment Isolation System (PCIS).

Leakage into a closed system is to be considered only if the closed system is breached and thereby creates a significant pathway to the environment. Examples include unisolable Main Steamline, RCIC steamline breaks, unisolable RWCU system breaks, and unisolable containment atmosphere vent paths.

Examples of "downstream pathway to the environment" could be through the Turbine/Condenser, or direct release to the Turbine or Reactor Building.

The existence of a filter is not considered in the threshold assessment. Filters do not remove fission product noble gases. In addition, a filter could become ineffective due to iodine and/or particulate loading beyond design limits (i.e., retention ability has been exceeded) or water saturation from steam/high humidity in the release stream.

Clinton Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION CT6 (cont)

Basis (cont):

Following the leakage of RCS mass into primary containment and a rise in primary containment pressure, there may be minor radiological releases associated with allowable primary containment leakage through various penetrations or system components. Minor releases may also occur if a primary containment isolation valve(s) fails to close but the primary containment atmosphere escapes to an enclosed system.

These releases do not constitute a loss or potential loss of primary containment but should be evaluated using the Recognition Category R ICs.

Loss Threshold #2 Basis EOPs may direct primary containment isolation valve logic(s) to be intentionally bypassed, even if offsite radioactivity release rate limits will be exceeded. Under these conditions with a valid primary containment isolation signal, the containment should also be considered lost if primary containment venting is actually performed.

Intentional venting of primary containment for primary containment pressure or combustible gas control to the secondary containment and/or the environment is a Loss of the Containment. Venting for primary containment pressure control when not in an accident situation (e.g., to control pressure below the Drywell high pressure scram setpoint or Continuous Containment Purge operating as designed, below isolation setpoints) does not meet the threshold condition.

Loss Threshold #3 Basis The Max Safe Operating Temperature is the highest value of this parameter at which neither: (1) equipment necessary for the safe shutdown of the plant will fail, nor (2) personnel access necessary for the safe shutdown of the plant will be precluded. EOPs utilize these temperatures levels to establish conditions under which RPV depressurization is required.

The temperatures should be confirmed to be caused by RCS leakage from a primary system. A primary system is defined to be the pipes, valves, and other equipment which connect directly to the RPV such that a reduction in RPV pressure will effect a decrease in the steam or water being discharged through an unisolated break in the system.

In general, multiple indications should be used to determine if a primary system is discharging outside Primary Containment. For example a high area temperature condition in conjunction with other indications (e.g. room flooding, reports of steam in Secondary Containment, an unexpected rise in Feedwater flowrate, or unexpected Main Turbine Control Valve closure) may indicate that a primary system is discharging outside Primary Containment. In combination with RCS Barrier RC4 Potential Loss Threshold #3 this threshold would result in a Site Area Emergency.

There is no Potential Loss threshold associated with Primary Containment Isolation Failure.

Clinton Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION CT6 {cont)

Basis Reference(s):

1.

NEI 99-01 Rev 6, Table 9-F-2

2.

CPS 4402.01, EOP-6 Primary Containment Control

3.

SAG-2

4.

Clinton Power Station Emergency Operating Procedures Technical Bases, Sections 8 and 9

5.

4406.01, EOP-8 Secondary Containment Control

6.

Clinton Power Station Emergency Operating Procedures Technical Bases, Section 10

7.

USAR Figure 6.2-132

Clinton Annex Initiating Condition:

RECOGNITION CATEGORY SYSTEM MALFUNCTIONS Loss of all On-site or Off-site communications capabilities.

Operating Mode Applicability:

1, 2, 3 Emergency Action Level (EAL):

Exelon Nuclear MU7

1. Loss of ALL Table M3 Onsite communications capability affecting the ability to perform routine operations.

OR

2. Loss of ALL Table M3 Offsite communication capability affecting the ability to perform offsite notifications.

OR

3. Loss of ALL Table M3 NRC communication capability affecting the ability to perform NRG notifications.

Table M3 Communications Capability System On site Off site NRC Plant Radio x

Plant Page x

PCS Phones x

x x

All telephone Lines x

x x

ENS x

x HPN x

x Satellite Phones x

x Basis:

This IC addresses a significant loss of on-site, offsite, or NRG communications capabilities.

While not a direct challenge to plant or personnel safety, this event warrants prompt notifications to Offsite Response Organizations (OROs) and the NRG.

This IC should be assessed only when extraordinary means are being utilized to make communications possible (e.g., use of non-plant, privately owned equipment, relaying of on-site information via individuals or multiple radio transmission points, individuals being sent to offsite locations, etc.).

Clinton Annex Basis (cont):

EAL #1Basis RECOGNITION CATEGORY SYSTEM MALFUNCTIONS Exelon Nuclear MU7 {cont)

Addresses a total loss of the communications methods used in support of routine plant operations.

EAL#2 Basis Addresses a total loss of the communications methods used to notify all OROs of an emergency declaration. The OROs referred to here are listed in procedure EP-MW-114-1 OO-F-01, Nuclear Accident Reporting System (NAAS) Form.

EAL#3 Basis Addresses a total loss of the communications methods used to notify the NRG of an emergency declaration.

Basis Reference(s):

1.

NEI 99-01 Rev 6, SU6

2.

EP-MW-124-1001, Facilities Inventories and Equipment Tests

3.

UFSAR Section 9.5.2

Clinton Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN I REFUELING SYSTEM MALFUNCTIONS Initiating Condition:

Loss of all onsite or offsite communications capabilities.

Operating Mode Applicability:

4,5,D Emergency Action Level (EAL):

CU4

1. Loss of ALL Table C1 Onsite communications capability affecting the ability to perform routine operations.

OR

2. Loss of ALL Table C1 Offsite communication capability affecting the ability to perform offsite notifications.

OR

3. Loss of ALL Table C1 NRC communication capability affecting the ability to perform NRC notifications.

Table C1 Communications Capability System On site Off site NRC Plant Radio x

Plant Page x

PCS Phones x

x x

All telephone Lines x

x x

ENS x

x HPN x

x Satellite Phones x

x Basis:

This IC addresses a significant loss of on-site, offsite, or NRC communications capabilities.

While not a direct challenge to plant or personnel safety, this event warrants prompt notifications to Offsite Response Organizations (OROs) and the NRC.

This IC should be assessed only when extraordinary means are being utilized to make communications possible (e.g., use of non-plant, privately owned equipment, relaying of on-site information via individuals or multiple radio transmission points, individuals being sent to offsite locations, etc.).

EAL #1 Basis Addresses a total loss of the communications methods used in support of routine plant operations.

Clinton Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN I REFUELING SYSTEM MALFUNCTIONS CU4 {cont)

Basis (cont):

EAL #2 Basis Addresses a total loss of the communications methods used to notify all OROs of an emergency declaration. The OROs referred to here are listed in procedure EP-MW-114-100-F-O 1, Nuclear Accident Reporting System (NAAS) Form.

EAL #3 Basis Addresses a total loss of the communications methods used to notify the NRG of an emergency declaration.

Basis Reference(s):

1.

NEI 99-01 Rev 6, CU5

2.

EP-MW-124-1001, Facilities Inventories and Equipment Tests

3.

UFSAR Section 9.5.2

ATIACHMENT3 REVISED RADIOLOGICAL EMERGENCY PLAN ANNEX INFORMATION FOR DRESDEN NUCLEAR POWER STATION EP-AA-1004 Enclosures

A - Revised EAL Comparison Matrix Document

8 - Revised EAL Basis Documents

Dresden - Revised Comparison Matrix NEI 99-01 Rev 6 Category: Containment Barrier Primary Containment Isolation Failure Operating Mode Applicability:

Power Operation, Startup, Hot Standby, Hot Shutdown Fission Product Barrier Threshold:

Loss CT3 A. UNISOLABLE direct downstream pathway to the environment exists after primary containment isolation signal OR B. Intentional primary containment venting per EOPs OR C. UNISOLABLE primary system leakage that results in exceeding EITHER of the following:

1. Max Safe Operating Temperature.

OR

2. Max Safe Operating Area Radiation Level.

Proposed EAL Category: Containment Barrier Primary Containment Isolation Failure Operating Mode Applicability:

1, 2, 3 Fission Product Barrier (FPB) Threshold:

Loss

1. UNISOLABLE direct downstream pathway to the environment exists after primary containment isolation signal.

OR CT6

2. Intentional Primary Containment venting/purging per EOPs or SAMGs due to accident conditions.

OR

3. UNISOLABLE primary system leakage that results in Secondary Containment area temperature > DEOP 300-1, Maximum Safe operating levels.

Page 1of1 Justification D NoChange 0 Difference D Deviation

1) Listed site-specific threshold values to ensure timely classification.

2) Due to the ability to declare in a timely manner Loss 3 only lists areas with installed monitoring instruments, and not areas requiring a survey to determine max safe values. The areas requiring a radiation survey are bound by loss 3 max safe temp instrumentation, and loss 1 UNISOLABLE direct downstream pathway to the environment exists after primary containment isolation signal.

Dresden Annex RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION Initiating Condition:

Primary Containment Pressure Operating Mode Applicability:

1, 2, 3 Fission Product Barrier (FPB) Threshold:

LOSS 1. Drywell pressure >2.0 psig.

AND

2. Drywell pressure rise is due to RCS leakage.

Basis:

Exelon Nuclear RC3 The > 2.0 psig primary containment pressure is the Drywell high pressure setpoint which indicates a LOCA by automatically initiating ECCS.

The second threshold condition focuses the fission product barrier loss threshold on a failure of the RCS instead of the non-LOCA malfunctions that may adversely affect primary containment pressure. Pressures of this magnitude can be caused by non-LOCA events such as a loss of Drywell cooling or inability to control primary containment vent/purge.

There is no Potential Loss threshold associated with Primary Containment Pressure.

Basis Reference(s):

1.

NEI 99-01 Rev 6, Table 9-F-2

2.

Technical Specifications Table 3.3.5.1-1

3.

DAN 902(3)-5 D-11

4.

DEOP 100 RPV Control

5.

DEOP 200-1 Primary Containment Control

Dresden Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION Initiating Condition:

Primary Containment Isolation Failure Operating Mode Applicability:

1, 2, 3 Fission Product Barrier (FPB) Threshold:

LOSS CT6

1. UNISOLABLE direct downstream pathway to the environment exists after primary containment isolation signal.

OR

2. Intentional Primary Containment venting/purging per EOPs or SAMGs due to accident conditions.

OR

3. UNISOLABLE primary system leakage that results in Secondary Containment area temperature> DEOP 300-1, Maximum Safe operating levels.

Basis:

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

Failure to isolate the leak, within 15 minutes or if known that the leak cannot be isolated within 15 minutes, from the start of the leak requires immediate classification.

These thresholds address incomplete containment isolation that allows an UNISOLABLE direct release to the environment.

Loss Threshold #1 Basis The use of the modifier "direct" in defining the release path discriminates against release paths through interfacing liquid systems or minor release pathways, such as instrument lines, not protected by the Primary Containment Isolation System (PCIS).

Leakage into a closed system is to be considered only if the closed system is breached and thereby creates a significant pathway to the environment. Examples include unisolable Main Steamline, HPCI steamline breaks, unisolable RWCU system breaks, and unisolable containment atmosphere vent paths.

Examples of "downstream pathway to the environment" could be through the Turbine/Condenser, or direct release to the Turbine or Reactor Building.

The existence of a filter is not considered in the threshold assessment. Filters do not remove fission product noble gases. In addition, a filter could become ineffective due to iodine and/or particulate loading beyond design limits (i.e., retention ability has been exceeded) or water saturation from steam/high humidity in the release stream.

Dresden Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION CT6 (cont)

Basis (cont):

Following the leakage of RCS mass into primary containment and a rise in primary containment pressure, there may be minor radiological releases associated with allowable primary containment leakage through various penetrations or system components. Minor releases may also occur if a primary containment isolation valve(s) fails to close but the primary containment atmosphere escapes to an enclosed system.

These releases do not constitute a loss or potential loss of primary containment but should be evaluated using the Recognition Category R ICs.

Loss Threshold #2 Basis EOPs may direct primary containment isolation valve logic(s) to be intentionally bypassed, even if offsite radioactivity release rate limits will be exceeded. Under these conditions with a valid primary containment isolation signal, the containment should also be considered lost if primary containment venting is actually performed.

Intentional venting of primary containment for primary containment pressure or combustible gas control to the secondary containment and/or the environment is a Loss of the Containment. Venting for primary containment pressure control when not in an accident situation (e.g., to control pressure below the Drywell high pressure scram setpoint) does not meet the threshold condition.

Loss Threshold #3 Basis The Max Safe Operating Temperature is the highest value of this parameter at which neither: (1) equipment necessary for the safe shutdown of the plant will fail, nor (2) personnel access necessary for the safe shutdown of the plant will be precluded. EOPs utilize these temperatures to establish conditions under which RPV depressurization is required.

The temperatures should be confirmed to be caused by RCS leakage from a primary system. A primary system is defined to be the pipes, valves, and other equipment which connect directly to the RPV such that a reduction in RPV pressure will effect a decrease in the steam or water being discharged through an unisolated break in the system.

In general, multiple indications should be used to determine if a primary system is discharging outside Primary Containment. For example, a high area temperature condition in conjunction with other indications (e.g. room flooding, reports of steam in the Reactor Building, an unexpected rise in Feedwater flowrate, or unexpected Main Turbine Control Valve closure) may indicate that a primary system is discharging into the Reactor Building.

In combination with RCS Barrier RC4 Potential Loss Threshold #3 this threshold would result in a Site Area Emergency.

There is no Potential Loss threshold associated with Primary Containment Isolation Failure.

Dresden Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION CT6 (cont)

Basis Reference(s):

1.

NEI 99-01 Rev 6, Table 9-F-2

2.

DEOP 200-1 Primary Containment Control

3.

DEOP 200-2 Hydrogen Control

4.

DEOP 500-4 Containment Venting

5.

DEOP 300-1 Secondary Containment Control

Dresden Annex Exelon Nuclear Initiating Condition:

RECOGNITION CATEGORY SYSTEM MALFUNCTIONS Prolonged loss of all Off-site and all On-Site AC power to emergency buses.

Operating Mode Applicability:

1, 2, 3 Emergency Action Level (EAL):

Note:

MG1 The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.

1 Loss of ALL offsite AC power to unit ECCS buses.

AND

2. Failure of DG 2(3), and shared DG 2/3 emergency diesel generators to supply power to unit ECCS buses.

AND

3. EITHER of the following:

a. Restoration of at least one unit ECCS bus in < 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is not likely.

OR

b. RPV water level cannot be restored and maintained > -164 inches.

Basis:

SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS.

These are typically systems classified as safety-related.

This IC addresses a prolonged loss of all power sources to AC emergency buses. A loss of all AC power compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink.

A prolonged loss of these buses will lead to a loss of any fission product barriers.

In addition, fission product barrier monitoring capabilities may be degraded under these conditions.

RPV values are actual levels, not indicated levels. Therefore, they may need level compensation depending on conditions.

The EAL should require declaration of a General Emergency prior to meeting the thresholds for IC FG1.

This will allow additional time for implementation of offsite protective actions.

Dresden Annex Basis (cont):

RECOGNITION CATEGORY SYSTEM MALFUNCTIONS Exelon Nuclear MG1 {cont)

Escalation of the emergency classification from Site Area Emergency will occur if it is projected that power cannot be restored to at least one AC emergency bus by the end of the analyzed station blackout coping period. Beyond this time, plant responses and event trajectory are subject to greater uncertainty, and there is an increased likelihood of challenges to multiple fission product barriers.

The estimate for restoring at least one emergency bus should be based on a realistic appraisal of the situation. Mitigation actions with a low probability of success should not be used as a basis for delaying a classification upgrade. The goal is to maximize the time available to prepare for, and implement, protective actions for the public.

The EAL will also require a General Emergency declaration if the loss of AC power results in parameters that indicate an inability to adequately remove decay heat from the core.

Basis Reference(s):

1.

NEI 99-01 Rev 6, SG1

2.

UFSAR 8.3

3.

12E-2302A, Station Key Diagram 4160V and 480V Switchgears Part 1

4.

DOA-6400-01, 138-kV System and 345-kV Alternate Supply Failure

5.

DOA 6500-01 4-KV Bus Failure

6.

UFSAR Fig. 9.5-14 Single-Line Electrical Diagram of Station Blackout Generator Ties to Plant Auxiliary Electric System

7.

UFSAR 9.5.9

8.

DOP 6620-05, Powering Unit 2(3) 4-KV Susses via the SBO D/G 2(3)

9.

DGA-12 Partial or Complete Loss of AC Power

10.

DEOP100 RPV Control

11.

DEOP 0010-00 Guidelines for Use of Dresden Emergency Operating Procedures and Severe Accident Management Guidelines

Dresden Annex Initiating Condition:

RECOGNITION CATEGORY SYSTEM MALFUNCTIONS Exelon Nuclear MS1 Loss of all offsite and all onsite AC power to emergency buses for 15 minutes or longer.

Operating Mode Applicability:

1, 2, 3 Emergency Action Level (EAL):

Note:

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

1. Loss of ALL off site AC Power to unit ECCS buses.

AND

2. Failure of DG 2(3), and shared DG 2/3 emergency diesel generators to supply power to unit ECCS buses.

AND

3. Failure to restore power to at least one ECCS bus in < 15 minutes from the time of loss of both offsite and onsite AC power.

Basis:

SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS.

These are typically systems classified as safety-related.

This IC addresses a total loss of AC power that compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink. In addition, fission product barrier monitoring capabilities may be degraded under these conditions. This IC represents a condition that involves actual or likely major failures of plant functions needed for the protection of the public.

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

Escalation of the emergency classification level would be via ICs RG1, FG1, MG1, or MG2.

Dresden Annex Basis Reference(s):

1.

NEI 99-01 Rev 6, SS1

2.

UFSAR 8.3 RECOGNITION CATEGORY SYSTEM MALFUNCTIONS Exelon Nuclear MS1 {cont)

3.

12E-2302A, Station Key Diagram 4160V and 480V Switchgears Part 1

4.

DOA-6400-01, 138-kV System and 345-kV Alternate Supply Failure

5.

DOA 6500-01 4KV Bus Failure

6.

UFSAR Fig. 9.5-14 Single-Line Electrical Diagram of Station Blackout Generator Ties to Plant Auxiliary Electric System

7.

UFSAR 9.5.9

8.

DOP 6620-05, Powering Unit 2(3) 4KV Susses via the SBO DIG 2(3)

9.

DGA-12 Partial or Complete Loss of AC Power

Dresden Annex Initiating Condition:

RECOGNITION CATEGORY SYSTEM MALFUNCTIONS Exelon Nuclear MA1 Loss of all but one AC power source to emergency buses for 15 minutes or longer.

Operating Mode Applicability:

1, 2, 3 Emergency Action Level (EAL):

Note:

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

1. AC power capability to unit ECCS buses reduced to only one of the following power sources for~ 15 minutes.

Reserve auxiliary Transformer TR-22 (TR-32)

Unit auxiliary transformer TR-21 (TR-31)

Unit Emergency Diesel Generator DG 2(3)

Shared Emergency Diesel Generator DG 2/3 Unit crosstie breakers AND

2. ANY additional single power source failure will result in a loss of ALL AC power to SAFETY SYSTEMS.

Basis:

SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS.

These are typically systems classified as safety-related.

This IC describes a significant degradation of offsite and onsite AC power sources such that any additional single failure would result in a loss of all AC power to SAFETY SYSTEMS. In this condition, the sole AC power source may be powering one, or more than one, train of safety-related equipment. This IC provides an escalation path from IC MU1.

An "AC power source" is a source recognized in AOPs and EOPs, and capable of supplying required power to an emergency bus. Some examples of this condition are presented below.

A loss of all offsite power with a concurrent failure of all but one emergency power source (e.g., an onsite diesel generator).

Dresden Annex Basis (cont):

RECOGNITION CATEGORY SYSTEM MALFUNCTIONS Exelon Nuclear MA1 {cont)

A loss of all offsite power and loss of all emergency power sources (e.g., onsite diesel generators) with a single train of emergency buses being fed from the unit main generator.

A loss of emergency power sources (e.g., onsite diesel generators) with a single train of emergency buses being fed from an offsite power source.

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

Escalation of the emergency classification level would be via IC MS 1.

Basis Reference(s):

1.

NEI 99-01 Rev 6, SA1

2.

UFSAR 8.3

3.

12E-2302A, Station Key Diagram 4160V and 480V Switchgears Part 1

4.

DOA-6400-01, 138 KV System and 345 KV Alternate Supply Failure

5.

DOA 6500-01 4KV Bus Failure

6.

UFSAR Fig. 9.5-14 Single-Line Electrical Diagram of Station Blackout Generator Ties to Plant Auxiliary Electric System

7.

UFSAR 9.5.9 Station Blackout System

8.

DOP 6620-05, Powering Unit 2(3) 4KV Susses via the SBO DIG 2(3)

9.

DGA-12 Partial or Complete Loss of AC Power

Dresden Annex Exelon Nuclear Initiating Condition:

RECOGNITION CATEGORY SYSTEM MALFUNCTIONS Loss of all AC and Vital DC power sources for 15 minutes or longer.

Operating Mode Applicability:

1, 2, 3 Emergency Action Level (EAL):

Note:

MG2 The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.

1.

Loss of ALL offsite AC power to unit ECCS buses.

AND

2.

Failure of DG 2(3) and shared DG 2/3 emergency diesel generators to supply power to vital buses.

AND

3.

Voltage is < 105 VDC on 125 VDC battery buses #2 and #3.

AND

4.

ALL AC and Vital DC power sources have been lost for~ 15 minutes.

Basis:

SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS.

These are typically systems classified as safety-related.

This IC addresses a concurrent and prolonged loss of both AC and Vital DC power. A loss of all AC power compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink. A loss of Vital DC power compromises the ability to monitor and control SAFETY SYSTEMS. A sustained loss of both AC and DC power will lead to multiple challenges to fission product barriers.

Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. The 15-minute emergency declaration clock begins at the point when all EAL conditions are met.

Dresden Annex Basis Reference(s):

1.

NEI 99-01 Rev 6, SG8

2.

UFSAR 8.3 RECOGNITION CATEGORY SYSTEM MALFUNCTIONS Exelon Nuclear MG2 (cont)

3.

12E-2302A, Station Key Diagram 4160V and 480V Switchgears Part 1

4.

DOA-6400-01, 138-kV System and 345-kV Alternate Supply Failure

5.

DOA 6500-01 4KV Bus Failure

6.

UFSAR Fig. 9.5-14 Single-Line Electrical Diagram of Station Blackout Generator Ties to Plant Auxiliary Electric System

7.

UFSAR 9.5.9

8.

DOP 6620-05, Powering Unit 2(3) 4KV Susses via the SBO D/G 2(3)

9.

DGA-12 Partial or Complete Loss of AC Power

10.

UFSAR 8.3.2

11.

DOA 6900-02(3) Failure of Unit 2(3) 125 VDC Power Supply

12.

Technical Specification B.3.8.4, DC Power Sources - Operating

Dresden Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN I REFUELING SYSTEM MALFUNCTIONS CA1 Initiating Condition:

Loss of all offsite and all onsite AC power to emergency buses for 15 minutes or longer.

Operating Mode Applicability:

4,5,D Emergency Action Level (EAL):

Note:

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

Loss of ALL offsite AC power to unit ECCS buses.

AND Failure of DG 2(3), and shared DG 2/3 emergency diesel generators to supply power to unit ECCS buses.

AND Failure to restore power to at least one unit ECCS bus in < 15 minutes from the time of loss of both offsite and onsite AC power.

Basis:

SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS.

These are typically systems classified as safety-related This IC addresses a total loss of AC power that compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink.

When in the cold shutdown, refueling, or defueled mode, this condition is not classified as a Site Area Emergency because of the increased time available to restore an emergency bus to service. Additional time is available due to the reduced core decay heat load, and the lower temperatures and pressures in various plant systems. Thus, when in these modes, this condition represents an actual or potential substantial degradation of the level of safety of the plant.

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

Escalation of the emergency classification level would be via IC CS6 or RS1.

Dresden Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN I REFUELING SYSTEM MALFUNCTIONS CA1 (cont)

Basis Reference(s):

1.

NEI 99-01 Rev 6, CA2

2.

UFSAR 8.3

3.

12E-2302A, Station Key Diagram 4160V and 480V Switchgears Part 1

4.

DOA-6400-01, 138 KV System and 345 KV Alternate Supply Failure

5.

DOA 6500-01 4KV Bus Failure

6.

UFSAR Fig. 9.5-14

7.

UFSAR 9.5.9

8.

DOP 6620-05, Powering Unit 2(3) 4KV Susses via the SBO DIG 2(3)

9.

DGA-12 Partial or Complete Loss of AC Power

Dresden Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN I REFUELING SYSTEM MALFUNCTIONS CU1 Initiating Condition:

Loss of all but one AC power source to emergency buses for 15 minutes or longer.

Operating Mode Applicability:

4,5, D Emergency Action Level (EAL):

Note:

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

1.

AC power capability to unit ECCS buses reduced to only one of the following power sources for~ 15 minutes.

Reserve auxiliary Transformer TR-22 (TR-32)

Unit auxiliary transformer TR-21 (TR-31)

Unit Emergency Diesel Generator DG 2(3)

Shared Emergency Diesel Generator DG 2/3 Unit crosstie breakers AND

2.

ANY additional single power source failure will result in a loss of ALL AC power to SAFETY SYSTEMS.

Basis:

SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS.

These are typically systems classified as safety-related.

This IC describes a significant degradation of offsite and onsite AC power sources such that any additional single failure would result in a loss of all AC power to SAFETY SYSTEMS. In this condition, the sole AC power source may be powering one, or more than one, train of safety-related equipment.

Dresden Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN I REFUELING SYSTEM MALFUNCTIONS CU1 (cont)

Basis (cont):

When in the cold shutdown, refueling, or defueled mode, this condition is not classified as an Alert because of the increased time available to restore another power source to service. Additional time is available due to the reduced core decay heat load, and the lower temperatures and pressures in various plant systems.

Thus, when in these modes, this condition is considered to be a potential degradation of the level of safety of the plant.

An "AC power source" is a source recognized in AOPs and EOPs, and capable of supplying required power to an emergency bus. Some examples of this condition are presented below.

A loss of all offsite power with a concurrent failure of all but one emergency power source (e.g., an onsite diesel generator).

A loss of all offsite power and loss of all emergency power sources (e.g., onsite diesel generators) with a single train of emergency buses being back-fed from the unit main generator.

A loss of emergency power sources (e.g., onsite diesel generators) with a single train of emergency buses being back-fed from an offsite power source.

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

The subsequent loss of the remaining single power source would escalate the event to an Alert in accordance with IC CA 1.

Basis Reference(s):

1.

NEI 99-01 Rev 6 CU2

2.

UFSAR 8.3

3.

12E-2302A, Station Key Diagram 4160V and 480V Switchgears Part 1

4.

DOA-6400-01, 138 KV System and 345 KV Alternate Supply Failure

5.

DOA 6500-01 4KV Bus Failure

6.

UFSAR Fig. 9.5-14 Single-Line Electrical Diagram of Station Blackout Generator Ties to Plant Auxiliary Electric System

7.

UFSAR 9.5.9 Station Blackout System

8.

DOP 6620-05, Powering Unit 2(3) 4KV Susses via the SBO DIG 2(3)

9.

DGA-12 Partial or Complete Loss of AC Power

ATTACHMENT 4 REVISED RADIOLOGICAL EMERGENCY PLAN ANNEX INFORMATION FOR LASALLE COUNTY STATION EP-AA-1005 Enclosure

A - Revised EAL Basis Document

LaSalle Annex RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION Initiating Condition:

Primary Containment Pressure Operating Mode Applicability:

1, 2, 3 Fission Product Barrier (FPB) Threshold:

LOSS

1. Drywell pressure >1.77 psig.

AND

2. Drywell pressure rise is due to RCS leakage Basis:

Exelon Nuclear RC3 The > 1.77 psig primary containment pressure is the Drywell high pressure setpoint which indicates a LOCA by automatically initiating ECCS.

The second threshold condition focuses the fission product barrier loss threshold on a failure of the RCS instead of the non-LOCA malfunctions that may adversely affect primary containment pressure. Pressures of this magnitude can be caused by non-LOCA events such as a loss of Drywell cooling or inability to control primary containment vent/purge.

There is no Potential Loss threshold associated with Primary Containment Pressure.

Basis Reference(s):

1.

NEI 99-01 Rev 6, Table 9-F-2

2.

UFSAR Table 3.3.5.1-1

3.

Technical Specifications Table 3.3.5.1-1

4.

LGA-001, RPV Control

5.

LGA-003, Primary Containment Control

ATTACHMENT 5 REVISED RADIOLOGICAL EMERGENCY PLAN ANNEX INFORMATION FOR LIMERICK GENERATING STATION EP-AA-1008 Enclosure Enclosure SA - Revised EAL Basis Document

Limerick Generating Station Annex RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION Initiating Condition:

Primary Containment Pressure Operating Mode Applicability:

1, 2, 3 Fission Product Barrier (FPB) Threshold:

LOSS

1. Drywall pressure > 1.68 psig.

AND

2. Drywall pressure rise is due to RCS leakage Basis:

Exelon Nuclear RC3 The > 1.68 psig primary containment pressure is the Drywall high pressure setpoint which indicates a LOCA by automatically initiating ECCS.

The second threshold condition focuses the fission product barrier loss threshold on a failure of the RCS instead of the non-LOCA malfunctions that may adversely affect primary containment pressure. Pressures of this magnitude can be caused by non-LOCA events such as a loss of Drywall cooling or inability to control primary containment vent/purge.

There is no Potential Loss threshold associated with Primary Containment Pressure.

Basis Reference(s):

1.

NEI 99-01 Rev 6, Table 9-F-2

2.

T-101 RPV Control

3.

T-102 Primary Containment Control - Bases

AITACHMENT6 REVISED RADIOLOGICAL EMERGENCY PLAN ANNEX INFORMATION FOR OYSTER CREEK NUCLEAR GENERATING STATION EP-AA-1010 Enclosure

A - Revised EAL Basis Document

Oyster Creek Generating Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION Initiating Condition:

Primary Containment Pressure Operating Mode Applicability:

1, 2 Fission Product Barrier (FPB) Threshold:

LOSS

1. Drywall pressure > 3.0 psig.

AND

2. Drywall pressure rise is due to RCS leakage Basis:

RC3 The> 3.0 psig primary containment pressure is the Drywall high pressure setpoint which indicates a LOCA by automatically initiating ECCS.

The second threshold condition focuses the fission product barrier loss threshold on a failure of the RCS instead of the non-LOCA malfunctions that may adversely affect primary containment pressure. Pressures of this magnitude can be caused by non-LOCA events such as a loss of Drywall cooling or inability to control primary containment vent/purge.

There is no Potential Loss threshold associated with Primary Containment Pressure.

Basis Reference(s):

1.

NEI 99-01 Rev 6, Table 9-F-2

2.

EMG-3200.01 A, RPV Control - No A TWS

3.

EMG-3200.02, Primary Containment Control

4.

2000-BAS-3200.02, EOP User's Guide

ATTACHMENT 7 DISCUSSION OF REVISION TO THE REVISED RADIOLOGICAL EMERGENCY PLAN ANNEX INFORMATION FOR PEACH BOTTOM ATOMIC POWER STATION EP-AA-1007 Enclosure

A - Revised EAL Basis Document

Peach Bottom Atomic Power Station Annex RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION Initiating Condition:

Primary Containment Pressure Operating Mode Applicability:

1, 2, 3 Fission Product Barrier (FPB) Threshold:

LOSS

1. Drywall pressure >2.0 psig.

AND

2. Drywall pressure rise is due to RCS leakage Basis:

Exelon Nuclear RC3 The> 2.0 psig primary containment pressure is the Drywall high pressure setpoint which indicates a LOCA by automatically initiating ECCS.

The second threshold condition focuses the fission product barrier loss threshold on a failure of the RCS instead of the non-LOCA malfunctions that may adversely affect primary containment pressure. Pressures of this magnitude can be caused by non-LOCA events such as a loss of Drywall cooling or inability to control primary containment vent/purge.

There is no Potential Loss threshold associated with Primary Containment Pressure.

Basis Reference(s):

1.

NEI 99-01 Rev 6, Table 9-F-2

2.

T-102, Primary Containment Control-Bases

3.

T-101, RPV Control

ATTACHMENT 8 REVISED RADIOLOGICAL EMERGENCY PLAN ANNEX INFORMATION FOR QUAD CITIES NUCLEAR POWER STATION EP-AA-1006 Enclosure Enclosure BA - Revised EAL Basis Documents

Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION Initiating Condition:

Primary Containment Pressure Operating Mode Applicability:

1, 2, 3 Fission Product Barrier (FPB) Threshold:

LOSS

1. Drywall pressure >2.5 psig.

AND

2. Drywall pressure rise is due to RCS leakage.

Basis:

RC3 The > 2.5 psig primary containment pressure is the Drywall high pressure setpoint which indicates a LOCA by automatically initiating ECCS.

The second threshold focuses the fission product barrier loss threshold on a failure of the RCS instead of the non-LOCA malfunctions that may adversely affect primary containment pressure. Pressures of this magnitude can be caused by non-LOCA events such as a loss of Drywall cooling or inability to control primary containment vent/purge.

There is no Potential Loss threshold associated with Primary Containment Pressure.

Basis Reference(s):

1.

NEI 99-01 Rev 6, Table 9-F-2

2.

QGA 100 RPV Control

3.

QGA 200 Primary Containment Control

Quad Cities Annex Exelon Nuclear Initiating Condition:

RECOGNITION CATEGORY SYSTEM MALFUNCTIONS Prolonged loss of all Off-site and all On-Site AC power to emergency busses.

Operating Mode Applicability:

1, 2, 3 Emergency Action Level (EAL):

Note:

MG1 The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.

1 Loss of ALL offsite AC power to unit ECCS busses.

AND

2. Failure of Unit EOG 1 (2), and shared EOG 1/2 emergency diesel generators to supply power to unit ECCS busses.

AND

3. EITHER of the following:

a. Restoration of at least one unit ECCS bus in < 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months is not likely.

OR

b. RPV water level cannot be restored and maintained> -190 inches.

Basis:

SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS.

These are typically systems classified as safety-related.

RPV values are actual levels, not indicated levels. Therefore, they may need level compensation depending on conditions. Compensated values may be used in accordance with the SAMG program.

This IC addresses a prolonged loss of all power sources to AC emergency buses. A loss of all AC power compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink.

A prolonged loss of these buses will lead to a loss of any fission product barriers.

In addition, fission product barrier monitoring capabilities may be degraded under these conditions.

The EAL should require declaration of a General Emergency prior to meeting the thresholds for IC FG1.

This will allow additional time for implementation of offsite protective actions.

Quad Cities Annex Basis (cont):

RECOGNITION CATEGORY SYSTEM MALFUNCTIONS Exelon Nuclear MG1 (cont)

Escalation of the emergency classification from Site Area Emergency will occur if it is projected that power cannot be restored to at least one AC emergency bus by the end of the analyzed station blackout coping period. Beyond this time, plant responses and event trajectory are subject to greater uncertainty, and there is an increased likelihood of challenges to multiple fission product barriers.

The estimate for restoring at least one emergency bus should be based on a realistic appraisal of the situation. Mitigation actions with a low probability of success should not be used as a basis for delaying a classification upgrade. The goal is to maximize the time available to prepare for, and implement, protective actions for the public.

The EAL will also require a General Emergency declaration if the loss of AC power results in parameters that indicate an inability to adequately remove decay heat from the core.

Basis Reference(s):

1.

NEI 99-01 Rev 6, SG1

2.

UFSAR Figure 8.3-1

3.

UFSAR Section 8.3

4.

QCOA 6100-03 Loss of Offsite Power

5.

QOP 6100-02 Restoring Reserve Auxiliary Transformer 12 To Service

6.

QOP 6100-04 Restoring Reserve Auxiliary Transformer 22 To Service

7.

QCOA 6100-04 Station Blackout

8.

GE letter No. 92-38 from LG. Knutson to Pat Donahue, dated April 7, 1992, "AC TURBINE LOADS SMALL TASK NO. QC107" (Station Blackout analysis)

9.

QGA 100 RPV Control

Quad Cities Annex Exelon Nuclear Initiating Condition:

RECOGNITION CATEGORY SYSTEM MALFUNCTIONS Loss of all offsite and all onsite AC power to emergency busses for 15 minutes or longer.

Operating Mode Applicability:

1, 2, 3 Emergency Action Level (EAL):

Note:

MS1 The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.

1. Loss of ALL offsite AC Power to unit ECCS busses.

AND

2. Failure of Unit EOG 1 (2), and shared EOG 1 /2 emergency diesel generators to supply power to unit ECCS busses.

AND

3. Failure to restore power to at least one ECCS bus in < 15 minutes from the time of loss of both offsite and onsite AC power.

Basis:

SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS.

These are typically systems classified as safety-related.

This IC addresses a total loss of AC power that compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink. In addition, fission product barrier monitoring capabilities may be degraded under these conditions. This IC represents a condition that involves actual or likely major failures of plant functions needed for the protection of the public.

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

Escalation of the emergency classification level would be via ICs RG1, FG1 or MG1.

Quad Cities Annex Basis Reference(s):

1.

NEI 99-01 Rev 6, SS1

2.

UFSAR Figure 8.3-1

3.

UFSAR Section 8.3 RECOGNITION CATEGORY SYSTEM MALFUNCTIONS

4.

QCOA 6100-03 Loss of Offsite Power Exelon Nuclear MS1 {cont)

5.

QOP 6100-02 Restoring Reserve Auxiliary Transformer 12 To Service

6.

QOP 6100-04 Restoring Reserve Auxiliary Transformer 22 To Service

7.

QCOA 6100-04 Station Blackout

8.

GE letter No. 92-38 from LG. Knutson to Pat Donahue, dated April 7, 1992, "AC TURBINE LOADS SMALL TASK NO. QC107" (Station Blackout analysis)

Quad Cities Annex Initiating Condition:

RECOGNITION CATEGORY SYSTEM MALFUNCTIONS Exelon Nuclear MA1 Loss of all but one AC power source to emergency buses for 15 minutes or longer.

Operating Mode Applicability:

1, 2, 3 Emergency Action Level (EAL):

Note:

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

1. AC power capability to unit ECCS busses reduced to only one of the following power sources for~ 15 minutes.

Reserve auxiliary Transformer TR-12 (TR-22)

Unit Auxiliary Transformer TR-11 (TR-21)

Unit Emergency Diesel Generator Shared Emergency Diesel Generator Unit crosstie breakers AND

2. ANY additional single power source failure will result in a loss of ALL AC power to SAFETY SYSTEMS.

Basis:

SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS.

These are typically systems classified as safety-related.

This IC describes a significant degradation of offsite and onsite AC power sources such that any additional single failure would result in a loss of all AC power to SAFETY SYSTEMS. In this condition, the sole AC power source may be powering one, or more than one, train of safety-related equipment. This IC provides an escalation path from IC MU1.

An "AC power source" is a source recognized in AOPs and EOPs, and capable of supplying required power to an emergency bus. Some examples of this condition are presented below.

A loss of all offsite power with a concurrent failure of all but one emergency power source (e.g., an onsite diesel generator).

Quad Cities Annex Basis (cont):

RECOGNITION CATEGORY SYSTEM MALFUNCTIONS Exelon Nuclear MA1 (cont)

A loss of emergency power sources (e.g., onsite diesel generators) with a single train of emergency buses being fed from an offsite power source.

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

Escalation of the emergency classification level would be via IC MS 1.

Basis Reference(s):

1.

NEI 99-01 Rev 6, SA 1

2.

UFSAR Figure 8.3-1

3.

UFSAR Section 8.3

4.

QCOA 6100-03 Loss of Offsite Power

5.

QOP 6100-02 Restoring Reserve Auxiliary Transformer 12 To Service

6.

QOP 6100-04 Restoring Reserve Auxiliary Transformer 22 To Service

7.

QCOA 6100-04 Station Blackout

8.

GE letter No. 92-38 from LG. Knutson to Pat Donahue, dated April 7, 1992, "AC TURBINE LOADS SMALL TASK NO. QC107" (Station Blackout analysis)

Quad Cities Annex Exelon Nuclear Initiating Condition:

RECOGNITION CATEGORY SYSTEM MALFUNCTIONS Loss of all AC and Vital DC power sources for 15 minutes or longer.

Operating Mode Applicability:

1, 2, 3 Emergency Action Level (EAL):

Note:

MG2 The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.

1.

Loss of ALL offsite AC power to unit EGGS busses.

AND

2.

Failure of Unit EDG 1 (2), and shared EDG 1/2 emergency diesel generators to supply power to vital busses.

AND

3.

Voltage is < 105 VDC on 125 VDC battery busses #1 and #2.

AND

4.

ALL AC and Vital DC power sources have been lost for~ 15 minutes.

Basis:

SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the EGGS.

These are typically systems classified as safety-related.

This IC addresses a concurrent and prolonged loss of both AC and Vital DC power. A loss of all AC power compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink. A loss of Vital DC power compromises the ability to monitor and control SAFETY SYSTEMS. A sustained loss of both AC and DC power will lead to multiple challenges to fission product barriers.

Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. The 15-minute emergency declaration clock begins at the point when all EALs are met.

Quad Cities Annex Basis Reference(s):

1.

NEI 99-01 Rev 6, SGS

2.

UFSAR Figure 8.3-1

3.

UFSAR Section 8.3 RECOGNITION CATEGORY SYSTEM MALFUNCTIONS

4.

QCOA 6100-03 Loss of Offsite Power Exelon Nuclear MG2 (cont}

5.

QOP 6100-02 Restoring Reserve Auxiliary Transformer 12 To Service

6.

QOP 6100-04 Restoring Reserve Auxiliary Transformer 22 To Service

7.

QCOA 6100-04 Station Blackout

8.

GE letter No. 92-38 from LG. Knutson to Pat Donahue, dated April 7, 1992, "AC TURBINE LOADS SMALL TASK NO. QC107" (Station Blackout analysis)

9.

Technical Specifications 3.8.4 and B3.8.4

10.

UFSAR Section 8.3.2 11.

QOP 6900-02 125 VDC Electrical System

12.

QCTS 0230-01 Unit One (Two) 125 VDC Service Test Normal or Alternate Battery

Quad Cities Annex Exelon Nuclear Initiating Condition:

RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MS3 Inability to shutdown the reactor causing a challenge to RPV water level or RCS heat removal.

Operating Mode Applicability:

1, 2 Emergency Action Level (EAL):

1. Automatic scram did not shutdown the reactor as indicated by Reactor Power> 5%.

AND

2. ALL manual I ARI actions to shutdown the reactor have been unsuccessful as indicated by Reactor Power > 5%.

AND

3. EITHER of the following conditions exist:

RPV water level cannot be restored and maintained > -190 inches.

OR Heat Capacity Limit (QGA 200, Figure M) exceeded.

Basis:

This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor scram that results in a reactor shutdown, all subsequent operator manual actions, both inside and outside the Control Room including driving in control rods and boron injection, are unsuccessful, and continued power generation is challenging the capability to adequately remove heat from the core and/or the RCS. This condition will lead to fuel damage if additional mitigation actions are unsuccessful and thus warrants the declaration of a Site Area Emergency.

In some instances, the emergency classification resulting from this IC/EAL may be higher than that resulting from an assessment of the plant responses and symptoms against the Recognition Category F ICs/EALs.

This is appropriate in that the Recognition Category F ICs/EALs do not address the additional threat posed by a failure to shutdown the reactor. The inclusion of this IC and EAL ensures the timely declaration of a Site Area Emergency in response to prolonged failure to shutdown the reactor.

A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.

RPV values are actual levels, not indicated levels. Therefore, they may need level compensation depending on conditions.

Escalation of the emergency classification level would be via IC RG1 or FG1.

Quad Cities Annex Basis Reference(s):

1.

NEI 99-01 Rev 6, SS5

2.

QGA 100 RPV Control RECOGNITION CATEGORY SYSTEM MALFUNCTIONS

3.

QGA 101 RPV Control (ATWS)

4.

QGA 200 Primary Containment Control Exelon Nuclear MS3 {cont)

Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN I REFUELING SYSTEM MALFUNCTIONS Initiating Condition:

Loss of all offsite and all onsite AC power to emergency busses for 15 minutes or longer.

Operating Mode Applicability:

4,5,D Emergency Action Level (EAL):

Note:

CA1 The Emergency Director should declare the event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.

1. Loss of ALL offsite AC power to unit ECCS busses.

AND

2. Failure of Unit EOG 1 (2), and shared EOG 1/2 emergency diesel generators to supply power to unit ECCS busses.

AND

3. Failure to restore power to at least one unit ECCS bus in < 15 minutes from the time of loss of both offsite and onsite AC power.

Basis:

SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS.

These are typically systems classified as safety-related This IC addresses a total loss of AC power that compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink.

When in the cold shutdown, refueling, or defueled mode, this condition is not classified as a Site Area Emergency because of the increased time available to restore an emergency bus to service. Additional time is available due to the reduced core decay heat load, and the lower temperatures and pressures in various plant systems. Thus, when in these modes, this condition represents an actual or potential substantial degradation of the level of safety of the plant.

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

Escalation of the emergency classification level would be via IC CS6 or RS1.

Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN I REFUELING SYSTEM MALFUNCTIONS CA1 {cont)

Basis Reference(s}:

1.

NEI 99-01 Rev 6, CA2

2.

UFSAR Figure 8.3-1

3.

UFSAR Section 8.3

4.

QCOA 6100-03 Loss of Offsite Power

5.

QOP 6100-02 Restoring Reserve Auxiliary Transformer 12 To Service

6.

QOP 6100-04 Restoring Reserve Auxiliary Transformer 22 To Service

7.

QCOA 6100-04 Station Blackout

8.

GE letter No. 92-38 from LG. Knutson to Pat Donahue, dated April 7, 1992, "ACTURBINE LOADS SMALL TASK NO. QC107" (Station Blackout analysis)

Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN I REFUELING SYSTEM MALFUNCTIONS CU1 Initiating Condition:

Loss of all but one AC power source to emergency buses for 15 minutes or longer.

Operating Mode Applicability:

4,5, D Emergency Action Level (EAL):

Note:

The Emergency Director should declare the event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.

1.

AC power capability to unit ECCS busses reduced to only one of the following power sources for~ 15 minutes.

Reserve auxiliary Transformer TR-12 (TR-22)

Unit auxiliary transformer TR-11 (TR-21)

Unit Emergency Diesel Generator Shared Emergency Diesel Generator Unit crosstie breakers AND

2.

ANY additional single power source failure will result in a loss of ALL AC power to SAFETY SYSTEMS.

Basis:

SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS.

These are typically systems classified as safety-related.

This IC describes a significant degradation of offsite and onsite AC power sources such that any additional single failure would result in a loss of all AC power to SAFETY SYSTEMS. In this condition, the sole AC power source may be powering one, or more than one, train of safety-related equipment.

When in the cold shutdown, refueling, or defueled mode, this condition is not classified as an Alert because of the increased time available to restore another power source to service. Additional time is available due to the reduced core decay heat load, and the lower temperatures and pressures in various plant systems.

Thus, when in these modes, this condition is considered to be a potential degradation of the level of safety of the plant.

Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN I REFUELING SYSTEM MALFUNCTIONS CU1 {cont)

Basis (cont):

An "AC power source" is a source recognized in AOPs and EOPs, and capable of supplying required power to an emergency bus. Some examples of this condition are presented below.

A loss of all offsite power with a concurrent failure of all but one emergency power source (e.g., an onsite diesel generator).

A loss of emergency power sources (e.g., onsite diesel generators) with a single train of emergency buses being fed from an offsite power source.

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

The subsequent loss of the remaining single power source would escalate the event to an Alert in accordance with IC CA 1.

Basis Reference(s):

1.

NEI 99-01 Rev 6 CU2

2.

UFSAR Figure 8.3-1

3.

UFSAR Section 8.3

4.

QCOA 6100-03 Loss of Offsite Power

5.

QOP 6100-02 Restoring Reserve Auxiliary Transformer 12 To Service

6.

QOP 6100-04 Restoring Reserve Auxiliary Transformer 22 To Service

7.

QCOA 6100-04 Station Blackout

8.

GE letter No. 92-38 from LG. Knutson to Pat Donahue, dated April 7, 1992, "AC TURBINE LOADS SMALL TASK NO. QC107" (Station Blackout analysis)

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RA-15-048, Response to Request for Additional Information on License Amendment Request to Adopt Emergency Action Level Schemes Pursuant to NEI 99-01, Revision 6, Development of Emergency Action Levels for Non-Passive Reactors

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RA-15-048, Response to Request for Additional Information on License Amendment Request to Adopt Emergency Action Level Schemes Pursuant to NEI 99-01, Revision 6, Development of Emergency Action Levels for Non-Passive Reactors

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