ML040980153

Request for License Amendments Related to Application of Alternative Source Term

ML040980153
Person / Time
Site:	Limerick
Issue date:	02/27/2004
From:	Gallagher M Exelon Generation Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
-RFPFR
Download: ML040980153 (266)

Text

Exelknm.

Exelon Nuclear www.exeloncorp.com Nuclear 200 Exelon WayNula Kennett Square, PA 19348 10 CFR 50.90 February 27, 2004 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk

Washington, DC 20555-0001 Limerick Generating Station, Units 1 & 2 Facility Operating License Nos. NPF-39 and NPF-85 NRC Docket Nos. 50-352 and 50-353

Subject:

Request for License Amendments Related to Application of Alternative Source Term

References:

(1) U. S. Nuclear Regulatory Commission, Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," July 2000 (2) U. S. Nuclear Regulatory Commission Standard Review Plan 15.0.1, "Radiological Consequence Analyses Using Alternative Source Terms,'

Revision 0, July 2000 (3) Technical Specification Task Force (TSTF) Traveler, TSTF-51, "Revise Containment Requirements During Handling of Irradiated Fuel and Core Alterations," Revision 2 (4) Exelon/AmerGen 180-Day Response to Generic Letter 2003-01, Control Room Habitability, December 9, 2003 Pursuant to 10 CFR 50.67, "Accident Source Term," and 10 CFR 50.90, "Application for amendment of license or construction permit," Exelon Generation Company, LLC (Exelon) he;,eby requests an amendment to the Facility Operating Licenses listed above. The proposed change is requested to support application of an alternative source term (AST) methodology, with the exception that Technical Information Document (TID) 14844, "Calculation of Distance Factors for Power and Test Reactor Sites," will continue to be used as the radiation dose basis for equipment qualification. This submittal has used the methods described in Regulatory Guide 1.183 (Reference 1) except for the few instances where alternative methods of compliance have been proposed as allowed by the guidance in this reference. 1hese alternative methods have been fully discussed in Tables A through E in Attachment 1 of this LAR.

On December 23,1999, the NRC published regulation 10 CUR 50.67 in the Federal Register.

This regulation provides a mechanism for operating license holders to revise the current accident source term used in design-basis radiological analyses with an AST. Regulatory A-ol

Request for License Amendments Related to Application of Alternative Source Term February 27, 2004 Page 2 guidance for the implementation of AST is provided in Reference 1. This regulatory guide provides guidance on acceptable applications of ASTs. The use of AST changes only the regulatory assumptions regarding the analytical treatment of the design basis accidents (DBAs).

Exelon has performed radiological consequence analyses of the four DBAs that result in offsite exposure to support a full-scope implementation of AST as described in Reference 1. The AST analyses for Limerick Generating Station (LGS), Units 1 & 2, were performed following the guidance in References 1 and 2.

The proposed changes to the current licensing basis for LGS that are justified by the AST analyses include:

• TS and associated Bases revisions to increase Main Steam Isolation Valve (MSIV) maximum allowable closure time;

• TS and associated Bases revisions to reflect the changes in the laboratory testing acceptance criteria associated with the charcoal adsorbers in the Reactor Enclosure Recirculation System (RERS), Standby Gas Treatment System (SGTS), and Control Room Emergency Fresh Air System (CREFAS).

• TS change reflecting replacement of automatic initiation of the CREFAS radiation mode with a 30-minute manual isolation.

• TS and associated Bases revisions to reflect lower RERS flows associated with the dose calculation requirements.

• TS and associated Bases revisions to change the applicability requirements for the following systems during movement of recently irradiated fuel assemblies in secondary containment and to reflect that these systems are no longer required to be operable during core alterations:

• Standby Gas Treatment System,

+ Secondary Containment, and

• Control Room Emergency Fresh Air System

• TS and associated Bases revisions to reflect use of the Standby Liquid Control (SLC)

System to buffer suppression pool pH to prevent iodine re-evolution during a postulated loss of coolant accident.

The proposed changes related to the applicability requirements during movement of irradiated fuel assemblies are consistent with Technical Specification Task Force Traveler (TSTF)-51, Revision 2.

TSTF-51, Revision 2, was approved by the NRC on November 1, 1999. TSTF-51 changes the TS operability requirements for certain engineered safety features such that they are not required after sufficient radioactive decay has occurred to ensure that offsite doses remain within limits. Since a portion of this license amendment request is based on TSTF-51, Exelon is committing to the applicable provisions of Nuclear Utilities Management and Resources Council (NUMARC) 93-01, "Industry Guideline for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants,"

Revision 3, as described in TSTF-51. NUMARC 93-01 provides recommendations on the need to

Request for License Amendments Related to Application of Alternative Source Term February 27, 2004 Page 3 initiate actions to verify and/or re-establish secondary containment, and if needed, primary containment, in the event of a fuel handling accident.

Exelon has been an active participant on the NEI Control Room Habitability (CRH) Task Force and understands the NRC position regarding CRH and acknowledges the fact that Generic Letter 2003-01 has been issued. This submittal does not directly address the CRH issue other than to provide an increase in the assumed unfiltered inleakage value. However, Exelon has provided a formal response to the Generic Letter (reference 4). Although an ASTM E741 tracer gas test has not been performed to date, the assumed unfiltered inleakage value in the AST dose analyses is equal to 100% of the full Control Room pressurization airflow in the emergency modes of operation. With the assumed inleakage value this high, it is Exelon's judgment that the measured value is not reasonably expected to exceed this assumed value. Other CRH actions have been addressed via Generic Letter response.

This request is subdivided as follows.

1. Attachment 1 provides a Description of Proposed Changes, Technical Analysis, and Regulatory Analysis.

2. Attachment 2 provides the Markup of Technical Specification pages.

3. Attachment 3 provides the Markup of Technical Specification Bases pages (for Information only).

4. Attachment 4 provides the Retyped Technical Specification pages.

5. Attachment 5 provides the Retyped Technical Specification Bases pages (for Information only).

6. Attachment 6 provides the List of Commitments resulting from the proposed changes.

7. Attachment 7 provides a compact disk (CD) containing LGS meteorological data for the calculation of the atmospheric dispersion factors (X/Qs). The CD also provides the PAVAN and ARCON96 input parameters.

8. Attachment 8 provides a discussion of the technical parameters and methodologies used in the AST calculations.

The proposed changes have been reviewed by the Plant Operations Review Committee and approved by the Nuclear Safety Review Board in accordance with the requirements of the Exelon Quality Assurance Program.

Exelon requests approval of the proposed amendments by February 27, 2005. Once approved, the amendments shall be implemented within 60 days. This implementation period will provide adequate time for the affected station documents to be revised using the appropriate change control mechanisms. In accordance with 10 CFR 50.91 (b), Exelon is notifying the State of Pennsylvania of this application for changes to the TS by transmitting a copy of this letter and its attachments to the designated State Official.

Request for License Amendments Related to Application of Alternative Source Term February 27, 2004 Page 4 If you have any questions or require additional information, please contact Doug Walker at (610) 765- 5726.

I declare under penalty of perjury that the foregoing is true and correct.

Respectfully, Executed on 02 a 7-o5(

Michael P. Gallagher Director, Licensing and Regulatory Affairs Mid-Atlantic Regional Operating Group Attachments: 1. Description of Proposed Changes, Technical Analysis, and Regulatory Analysis

2. Markup of Technical Specification pages

3. Markup of Technical Specification Bases pages (Information only)

4. Retyped Technical Specification pages

5. Retyped Technical Specification Bases pages (Information only)

6. List of Commitments

7. LGS Meteorological data (Information only)

8. Technical Parameters for AST Calculations cc: H. J. Miller, Administrator, Region I, USNRC S. Hansel, USNRC Senior Resident Inspector, LGS G. Wunder, Senior Project Manager Limerick (acting), USNRC (by FedEx)

R. R. Janati - Commonwealth of Pennsylvania

ATTACHMENT 1 Limerick Generating Station Units I & 2 License Amendment Request "LGS Alternative Source Term Implementation"

2.0 PROPOSED CHANGE

S

5.0 REGULATORY ANALYSIS

5.1 No Significant Hazards Consideration 5.2 Applicable Regulatory Requirements/Criteria

7.0 REFERENCES

Request for License Amendments Page 1 of 76 Alternative Source Term February 27, 2004

1.0 DESCRIPTION

In accordance with 10 CFR 50.67, "Accident Source Term," and 10 CFR 50.90, Application for amendment of license or construction permit," Exelon Generation Company, LLC (Exelon) requests a change to Appendix A, Technical Specifications (TS), of Facility Operating License Nos. NPF-39 and NPF-85 for the Limerick Generating Station (LGS), Units 1 & 2. The proposed changes are requested to support application of an alternative source term (AST) methodology, with the exception that Technical Information Document (TID) 14844, 'Calculation of Distance Factors for Power and Test Reactor Sites," will continue to be used as the radiation dose basis for equipment qualification.

Radiological consequence analyses have been performed for the four bounding Design Basis Accidents (DBAs) that result in offsite exposure (i.e., Loss of Coolant Accident (LOCA), Main Steam Line Break (MSLB), Fuel Handling Accident (FHA),

and Control Rod Drop Accident (CRDA)) to support a full-scope implementation of AST. The AST analyses for LGS were performed following the guidance in Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors and Standard Review Plan 15.0.1, "Radiological Consequence Analyses Using Alternative Source Terms". These analyses have been performed using NRC approved computer codes by qualified consultants and have had extensive cross-functional reviews and challenges by Exelon personnel.

The proposed changes to the TS will allow LGS to apply the results of the plant-specific AST analyses using the guidance in Regulatory Guide 1.183 and meet the requirements of 10 CFR 50.67. Approval of this change will provide a source term for LGS that will result in a more accurate assessment of the DBA radiological doses.

The improved dose assessment allows relaxation of some current licensing basis requirements as described below.

This proposed change will increase allowable Main Steam Isolation Valve (MSIV) closure time from 5 to 10 seconds. Unplanned MSIV repairs are potential contributors to increased outage duration and unplanned personnel exposure.

The radiological analysis includes a leakage rate of 100 scf per hour for any main steam isolation valve and a combined maximum main steam line pathway leakage of 200 scf per hour, therefore, the current requirement to satisfy a maximum of 11.5 scf per hour for any main steam isolation valve after restoration is being removed.

To satisfy the condition of application of AST to control the suppression pool pH following a LOCA, LGS is proposing to use the Standby Liquid Control (SLC)

System. This requires revising the Technical Specifications applicability requirements for the SLC system to include Operational Condition 3. Clarification is also being made to the Surveillance Requirements section for the SLC system to verify the value for the required weight of Boron-1 0, instead of using the currently specified weight of sodium pentaborate. This is an equivalent change.

In addition, implementation of AST will no longer require secondary containment to be established except during Operations with the Potential for Draining the Reactor Request for License Amendments Page 2 of 76 Alternative Source Term February 27, 2004 Vessel (OPDVs) and movement of recently irradiated fuel. This proposed change provides the flexibility of performing fuel floor activities (such as control rod blade exchanges and fuel movements) as well as movement of large equipment through the secondary containment boundary in support of outage activities while remaining within all safety limits.

Other benefits of AST are the cost savings that will be achieved by reducing credited charcoal efficiencies in accident analyses. This additional margin will extend available charcoal life by changing methyl iodine penetration acceptance criteria and will result in less frequent charcoal filter regeneration. HEPA efficiency credit reductions provide additional operating margin, but no reduction in test acceptance criteria are proposed. These benefits apply to the Reactor Enclosure Recirculation System (RERS), Standby Gas Treatment System (SGTS), and the Control Room Emergency Fresh Air System (CREFAS).

Adopting the AST methodology may also support future evaluations and license amendments.

2.0 PROPOSED CHANGE

S The proposed changes related to the applicability requirements during movement of irradiated fuel assemblies are consistent with Technical Specification Task Force Traveler (TSTF)-51, NRevise Containment Requirements During Handling of Irradiated Fuel and Core Alterations," Revision 2. The NRC approved TSTF-51 on October 15, 1999. TSTF-51 changes the TS operability requirements for engineered safety features such that they are not required after sufficient radioactive decay has occurred to ensure that offsite doses remain within limits.

Since a portion of this license amendment request is based on TSTF-51, Exelon is committing to the applicable provisions of Nuclear Utilities Management and Resources Council (NUMARC) 93-01, "Industry Guideline for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants," Revision 3, as described in TSTF-51.

NUMARC 93-01 provides recommendations on the need to initiate actions to verify and/or re-establish secondary containment, and if needed, primary containment, in the event of a fuel handling accident.

Proposed changes to the Technical Specifications resulting from this submittal are summarized below:

2.1 TS Section 1.0, "Definitions' The proposed change revises the definition of DOSE EQUIVALENT 1-131 in TS Definition 1.9 to replace the word "thyroid" with "inhalation committed effective dose equivalent (CEDE)" and to add a reference to "Table 2.1 of Federal Guidance Report 11, Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion, ORNL, 1989, as described in Regulatory Guide 1.183".

2.2 TS Section 1.0, "Definitions" Request for License Amendments Page 3 of 76 Alternative Source Term February 27, 2004 The proposed change adds the definition of RECENTLY IRRADIATED FUEL as TS Definition 1.35. RECENTLY IRRADIATED FUEL is fuel that has occupied part of a critical reactor core within the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Subsequent definitions in this section are renumbered to reflect this addition.

2.3 TS Section 3/4 1.5, "Standby Liquid Control (SLC) System" The proposed change revises the Applicability of TS Section 3.1.5 to include Operational Condition 3 for the SLC system. This change implements AST assumptions regarding the use of the SLC System to control the suppression pool pH following a LOCA involving significant fission product release. Action 3.1.5 has been revised to include action statements for inoperable SLC equipment in Operational Condition 3, which can include going to COLD SHUTDOWN. SR 4.1.5.b.2 is revised to reflect the Boron-10 weight requirement that is equivalent to the current requirements for Sodium Pentaborate at 29% enrichment.

2.4 TS Section Tables 3.3.2-1, "Isolation Actuation Instrumentation Action Statements" The proposed change revises Table Notation (*) for TS Table 3.3.2-1 by: 1) replacing the term "irradiated fuel" with "RECENTLY IRRADIATED FUEL;" 2) removing "refueling area," since secondary containment can consist of the common refueling area and the Reactor Enclosure zones; and 3) deleting the "during CORE ALTERATIONS" criteria. The table notation applies to the applicable operation conditions for the Refueling Area Unit 1 and Unit 2 Ventilation Exhaust Duct Radiation - High and the Refueling Area Manual isolation instrumentation. These changes are consistent with TSTF-51.

2.5 TS Section Table 4.3.2.1-1, "Isolation Actuation Instrumentation Surveillance Requirements" The proposed change revises Table Notation (*) for TS Table 4.3.2.1-1 by:

1) replacing the term "irradiated fuel" with 'RECENTLY IRRADIATED FUEL;"

2) removing "refueling area," since secondary containment can consist of the common refueling area and the Reactor Enclosure zones;" and 3) deleting the "during CORE ALTERATIONS" criteria. The table notation applies to the operation conditions for which surveillance is required for the Refueling Area Unit 1 and Unit 2 Ventilation Exhaust Duct Radiation - High and the Refueling Area Manual isolation instrumentation. These changes are consistent with TSTF-51.

2.6 TS Section Table 3.3.7.1-1, "Radiation Monitoring Instrumentation" The proposed change revises Table Notation (*) for TS Table 3.3.7.1-1 by replacing the term "irradiated fuel" with "RECENTLY IRRADIATED FUEL" and adding the criteria "or during operations with a potential for draining the reactor vessel with the vessel head removed and fuel in the vessel". The table notation applies to the applicable operation conditions for the Main Control Room Normal Fresh Air Supply Radiation Monitor. In addition, the Main Control Room Normal Fresh Air Supply Radiation Monitor is no longer applicable to Operational Condition 5 and is only required as an Alarm function only. The trip function is being removed.

Request for License Amendments Page 4 of 76 Alternative Source Term February 27, 2004 2.7 TS Section Table 4.3.7.1-1, "Radiation Monitoring Instrumentation Surveillance Requirements' The proposed change revises Table Notation (*) for TS Table 4.3.7.1-1 by replacing the term "irradiated fuel" with "RECENTLY IRRADIATED FUEL" and adding the criteria "or during operations with a potential for draining the reactor vessel with the vessel head removed and fuel in the vessel". The table notation applies to the operation conditions for which surveillance is required for the Main Control Room Normal Fresh Air Supply Radiation Monitor. In addition, the Main Control Room Normal Fresh Air Supply Radiation Monitor is no longer applicable to Operational Condition 5.

2.8 TS Section 3.4.7 and 4.4.7, "Main Steam Isolation Valves (MSIV)'

The proposed change revises Limiting Condition for Operation 3.4.7 to increase the MSIV maximum closing time from "less than or equal to 5 seconds" to "less than or equal to 10 seconds". Additionally, the proposed change also revises the Surveillance Requirement 4.4.7 to increase the MSIV full closure from "between 3 and 5 seconds" to "between 3 and 10 seconds".

2.9 TS 3.6.1.2, Restore Action c., "Primary Containment Leakage" The proposed change revises the action statement to restore "the leakage rate to

<100 scf per hour for any MSIV that exceeds 100 scf per hour." The current restore value is < 11.5 scf per hour, for any MSIV that exceeds 100 scfh, based on the existing radiological analysis.

2.10 TS Section 3.6.5.1.2, "Refueling Area Secondary Containment Integrity The proposed change deletes "OPERATIONAL CONDITION *" in the Applicability section of TS 3.6.5.1.2 and the corresponding explanation is relocated from the bottom of the page. Additionally, the Applicability and Action Statements are revised by replacing the term "irradiated fuel" with "RECENTLY IRRADIATED FUEL" and deleting reference to 'CORE ALTERATIONS".

2.11 TS Section 3.6.5.2.2, "Refueling Area Secondary Containment Automatic Isolation Valves" The proposed change deletes the "OPERATIONAL CONDITION *" in the Applicability section of TS 3.6.5.2.2 and the corresponding explanation is relocated from the bottom of the page. Additionally, the Applicability and Action Statements are revised by replacing the term "irradiated fuel" with "RECENTLY IRRADIATED FUEL" and deleting reference to 'CORE ALTERATIONS".

2.12 TS Section 3.6.5.3, "Standby Gas Treatment System - Common System' The proposed change deletes the (*) in the Applicability and Action section of TS 3.6.5.3 and the corresponding explanation is relocated from the bottom of the page.

Additionally, the Applicability section and Action Statements a.2 and b. are revised by replacing the term "irradiated fuel" with "RECENTLY IRRADIATED FUEL" and deleting references to "CORE ALTERATIONS". The action statement b. is being Request for License Amendments Page 5 of 76 Alternative Source Term February 27, 2004 revised to only be applicable to handling of recently irradiated fuel in the secondary containment, or during operations with a potential for draining the reactor vessel.

2.13 TS Section 4.6.5.3, "Standby Gas Treatment System - Common System" The charcoal adsorber sample acceptance criteria for the methyl iodide penetration tests in Surveillance Requirements 4.6.5.3.b.2 and 4.6.5.3.c has been increased from less than 0.5% to less than 1.25%.

2.14 TS Section 4.6.5.4, "Reactor Enclosure Recirculation System" The proposed change relaxes the following Surveillance Requirements (SR) related to the RERS charcoal adsorbers as shown:

• SR 4.6.5.4.a to annotate a flow range through the HEPA filters of a minimum of 30,000 cfm through the HEPA filters

• SR 4.6.5.4.b.1 to clarify that the in-place penetration test is performed at the rated flow rate (60,000 cfm + 10%) instead of annotating a specific flow of 60,000 cfm + 10%.

• SR 4.6.5.4.b.2 to verify at least once per 24 months, or (1)after structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire, or chemical release in any communicating ventilation zone, that a laboratory analysis of a representative carbon sample obtained shows methyl iodide penetration of less than 15% rather than 2.5%.

• SR 4.6.5.4.b.3 to verify a subsystem flow rate within a range of 30,000 to 66,000 cfm.

• SR 4.6.5.4.c to verify after 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of operation, that a laboratory analysis of a representative carbon sample shows methyl iodide penetration of less than 15%

rather than 2.5%.

• SR 4.6.5.4.d.1 to clarify that the in-place penetration test is performed at the rated flow rate (60,000 cfm + 10%) instead of annotating a specific flow of 60,000 cfm + 10%.

• SR 4.6.5.4.e to clarify that the in-place penetration test is performed at the rated flow rate (60,000 cfm + 10%) instead of annotating a specific flow of 60,000 cfm

+ 10%.

• SR 4.6.5.4.f to clarify that the in-place penetration test is performed at the rated flow rate (60,000 cfm + 10%) instead of annotating a specific flow of 60,000 cfm

+ 10%.

2.15 TS Section 3.7.1.2, "Emergency Service Water System - Common System" The proposed change expands the definition of the (*) to include "handling RECENTLY IRRADIATED FUEL in the secondary containment and during operations with a potential for draining the reactor vessel" (consistent with TSTF-51).

Additionally, the (*) in the LCO, the Applicability section and Action c. of TS 3.7.1.2 is Request for License Amendments Page 6 of 76 Alternative Source Term February 27, 2004 deleted and the corresponding explanation is relocated from the bottom of the applicable page.

2.16 TS Section 3.7.1.3, "Ultimate Heat Sink" The proposed change expands the definition of the (*) to include "handling RECENTLY IRRADIATED FUEL in the secondary containment and during operations with a potential for draining the reactor vessel" (consistent with TSTF-51).

Additionally, the reference to OPERATIONAL CONDITION (*) in the Applicability section and Action c. of TS 3.7.1.3 is deleted and the corresponding explanation is relocated from the bottom of the page.

2.17 TS Section 3.7.2, "Control Room Emergency Fresh Air Supply System - Common System "

- Applicability Section - the proposed change expands the definition of the (*) to include "when handling RECENTLY IRRADIATED FUEL in the secondary containment, or during operations with a potential for draining the reactor vessel" (consistent with TSTF-51). Additionally, the (*) in the Applicability section is deleted and the corresponding explanation is relocated from the bottom of the page.

- Action b. - the operational condition is revised to expand the definition of the (*)

to include "when handling RECENTLY IRRADIATED FUEL in the secondary containment, or during operations with a potential for draining the reactor vessel" (consistent with TSTF-51). Additionally, the (*) in the Applicability section is deleted and the corresponding explanation is relocated from the bottom of the page.

- Action b.2 - the action statement is revised by replacing the term "irradiated fuel" with "RECENTLY IRRADIATED FUEL" and deleting reference to "CORE ALTERATIONS".

- Action c - the reference to Operational Condition (*) is deleted and the action has been incorporated into Action b.2.

- Notation (*) at the bottom of the page is deleted and included in the applicable sections.

2.18 TS Section 4.7.2, "Control Room Emergency Fresh Air Supply System - Common System" The proposed change relaxes the following Surveillance Requirements (SR) related to the charcoal adsorbers as shown:

• The charcoal adsorber sample acceptance criteria for the methyl iodide penetration tests in Surveillance Requirements 4.7.2.c.2 and 4.2.7.d has been increased from less than 2.5% to less than 10%.

• The proposed change revises SR 4.7.2.e.3 to only require verification of the manual initiation of the radiation mode of CREFAS and removes reference to the outside air intake high radiation mode.

Request for License Amendments Page 7 of 76 Alternative Source Term February 27, 2004 2.19 TS Section 3.8.1.2, "AC Sources - Shutdown" The proposed change expands the definition of the (*) to include "when handling RECENTLY IRRADIATED FUEL in the secondary containment or during operations with a potential for draining the reactor vessel" (consistent with TSTF-51).

Additionally, the (*) in the Applicability section of TS 3.8.1.2 is deleted and the corresponding explanation is relocated from the bottom of the page.

2.20 TS Section 3.8.2.2, "DC Sources - Shutdown" The proposed change expands the definition of the (*) to include "when handling RECENTLY IRRADIATED FUEL in the secondary containment or during operations with a potential for draining the reactor vessel" (consistent with TSTF-51).

Additionally, the (*) in the Applicability section of TS 3.8.2.2 is deleted and the corresponding explanation is relocated from the bottom of the page. The statement in TS 3.8.2.2 Action c. is revised to change 'irradiated fuel" to "RECENTLY IRRADIATED FUEL."

2.21 TS Section 3.8.3.2 "Electrical Power Systems, Distribution - Shutdown" The proposed change expands the definition of the (*) to include "when handling RECENTLY IRRADIATED FUEL in the secondary containment or during operations with a potential for draining the reactor vessel" (consistent with TSTF-51).

Additionally, the (*) in the Applicability section of TS 3.8.3.2 is deleted and the corresponding explanation is relocated from the bottom of the page. The statement in TS 3.8.3.2 Actions a. and b. are revised to change 'irradiated fuel" to "RECENTLY IRRADIATED FUEL."

3.0 BACKGROUND

On December 23, 1999, the NRC published regulation 10 CFR 50.67 in the Federal Register. This regulation provides a mechanism for operating license holders to revise the current accident source term used in design-basis radiological analyses with an Alternate Source Term (AST). Regulatory guidance for the implementation of AST is provided in Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors", July 2000 (Reference 7.2). This regulatory guide provides guidance on acceptable applications of ASTs. The use of AST changes only the regulatory assumptions regarding the analytical treatment of the design basis accidents (DBAs).

The fission product release from the reactor core into containment is referred to as the "source term," and it is characterized by the composition and magnitude of the radioactive material, the chemical and physical properties of the material, and the timing of the release from the reactor core. Since the publication of U.S. Atomic Energy Commission Technical Information Document, TID-14844, Calculation of Distance Factors for Power and Test Reactor Sites (Reference 7.1), significant advances have been made in understanding the composition and magnitude, chemical form, and timing of fission product releases from severe nuclear power plant accidents. Many of these insights developed out of the major research efforts Request for License Amendments Page 8 of 76 Alternative Source Term February 27, 2004 started by the NRC and the nuclear industry after the accident at Three Mile Island.

NUREG-1465 (Reference 7.4) was published in 1995 with revised ASTs for use in the licensing of future Light Water Reactors (LWRs). The NRC, in 10 CFR 50.67, later allowed the use of the ASTs described in NUREG-1465 at operating plants.

This NUREG represents the result of decades of research on fission product release and transport in LWRs under accident conditions. One of the major insights summarized in NUREG-1465 involves the timing and duration of fission product releases.

The five release phases representing the progress of a severe accident in a LWR are described in NUREG-1465 as:

1. Coolant Activity Release

2. Gap Activity Release

3. Early In-Vessel Release

4. Ex-Vessel Release

5. Late In-Vessel Release Phases 1, 2, and 3 are considered in current DBA evaluations; however, they are all assumed to occur instantaneously. Phases 4 and 5 are related to severe accident evaluations. Under the AST, the coolant activity release is assumed to occur instantaneously and end with the onset of the gap activity release.

The requested license amendment involves a full-scope application of the AST, addressing the composition and magnitude of the radioactive material, its chemical and physical form, and the timing of its release as described in Regulatory Guide 1.183.

Exelon has performed radiological consequence analyses of the four DBAs that result in the most significant offsite exposures (i.e., LOCA, MSLB, FHA, and CRDA).

These analyses were performed to support full scope implementation of AST. The AST analyses have been performed in accordance with the guidance in Regulatory Guide 1.183 and NRC Standard Review Plan 15.0.1, "Radiological Consequence Analyses Using Alternative Source Terms" (Reference 7.3). The implementation consisted of the following steps:

• Identification of the AST based on plant-specific analysis of core fission product inventory,

• Calculation of the release fractions for the four DBAs that result in the most significant control room and offsite doses (i.e., LOCA, MSLB, FHA, and CRDA),

• Analysis of the atmospheric dispersion for the radiological propagation pathways,

• Calculation of fission product deposition rates and transport and removal mechanisms,

• Calculation of offsite and control room personnel Total Effective Dose Equivalent (TEDE) doses, and

Attachment 1 Request for License Amendments Page 9 of 76 Alternative Source Term February 27, 2004

• Evaluation of suppression pool pH to ensure that the iodine deposited into the suppression pool during a DBA LOCA does not re-evolve and become airborne as elemental iodine.

The analysis assumptions for the transport, reduction, and release of the radioactive material from the fuel and the reactor coolant are consistent with the guidance provided in applicable appendices of Regulatory Guide 1.183 for the four analyzed DBAs.

Accordingly, Exelon, as a holder of an operating license issued prior to January 10 1997, is requesting the use of AST for several areas of operational relief for systems used in the event of a Design Basis Accident (DBA), and without crediting the use of certain previously assumed safety systems/functions.

4.0 TECHNICAL ANALYSIS

4.1 Evaluation 4.1.1 Scope 4.1.1.1 Accident Radiological Consequence Analyses The DBA accident analyses documented in the LGS UFSAR that could potentially result in control room and offsite doses were addressed using methods and input assumptions consistent with AST. The following DBAs were addressed:

• CRDA, UFSAR Section 15.4.9;

• MSLB, UFSAR Section 15.6.4;

• LOCA, UFSAR Section 15.6.5; and

• FHA, UFSAR Section 15.7.4.

The analyses were performed in accordance with Regulatory Guide 1.183 to confirm compliance with the acceptance criteria presented in 10 CFR 50.67.

4.1.2 NUREG-0737, Item 1l.B.2 Exelon has determined that continued compliance will be maintained with NUREG-0737, Item ll.B.2, "Design Review of Plant Shielding and Environmental Qualification of Equipment for Spaces/Systems Which May be Used in Post-Accident Operations."

The source term associated with environmental qualification of equipment will remain consistent with previous commitments under 10 CFR 50.49.

Request for License Amendments Page 10 of 76 Alternative Source Term February 27, 2004 4.2 Method of Evaluation 4.2.1 Fission Product Inventory Pre-AST core source terms were determined based on TID-14844 methodology.

That is, inventory was based on the fission product equilibrium based on U-235 fission product yields and isotopic decay constants. In accordance with Regulatory Guide 1.183, this simplified approach is replaced with ORIGEN 2.1 (Reference 7.5) methodology used to determine core inventory. This program provides a more complete and accurate simulation of isotopic buildup and depletion, including consideration of fission product yields from all isotopes, and activation as well as decay.

The current licensed thermal power level at Limerick is 3458 MWt. These source terms were evaluated at end-of-cycle and at beginning of cycle (100 effective full power days (EFPD) to achieve equilibrium) conditions and worst-case inventory used for the selected isotopes. These values were then divided by power level to obtain activity in units of Ci/MWt. Accident analyses are based on a 3527 MWt power level that includes the current accident analysis design basis allowance for instrument uncertainty.

Source terms were based on a 2-year fuel cycle with a nominal 711 EFPD per cycle.

These source terms were developed using ORIGEN 2.1. The values extracted from the ORIGEN 2.1 runs generated for Peach Bottom, which are also applicable to Limerick, and are for the standard 60-isotope RADTRAD (Reference 7.6) library except that the activation products Co-58 and Co-60 used RADTRAD default library values.

The reactor coolant fission product inventory for MSLB analysis was based on the Technical Specification limits in terms of Dose Equivalent 1-131 (the concentration of 1-131 that alone would produce the same dose as the quantity and isotopic mixture of 1-131,1-132,1-133,1-134, and 1-135 actually assumed), using inhalation Committed Effective Dose Equivalent (CEDE) dose conversion factors from Federal Guidance Report 11 (Reference 7.15) as described in Regulatory Guide 1.183.

4.2.2 Radiological Consequence New Design Analyses were prepared for the simulation of the radionuclide release, transport, removal, and dose estimates associated with the postulated accidents listed in Section 4.1.1.1.

The RADTRAD computer code was used for these analyses. The RADTRAD program is a radiological consequence analysis code used to estimate post-accident doses at plant offsite locations and in the control room. The RADTRAD code is publicly available and is used by the NRC in safety reviews.

Offsite exclusion area boundary (EAB) and low population zone (LPZ) atmospheric dispersion factors (X/Qs) were calculated using the guidance of Regulatory Guide Request for License Amendments Page 11 of 76 Alternative Source Term February 27, 2004 1.145 (Reference 7.7) and the PAVAN computer code (Reference 7.8). This code has been used by the NRC in safety reviews.

The X/Q values resulting at the Control Room Intake were calculated using the NRC-sponsored computer codes ARCON96 (Reference 7.9), consistent with the procedures in Regulatory Guide 1.194 (Reference 7.19).

Figure 1 shows the "Layout of Release Points for LGS."

Airborne radioactivity drawn into the control room envelope results in both internal and external dose components that are used in the TEDE dose calculation. The noble gas inventory within the control room is the main contributor to the gamma ray whole body (i.e., external) dose component of the TEDE; the non-noble gas radionuclides, principally iodines, contribute to the internal organ dose component via the inhalation pathway.

Regulatory Guide 1.183, Section 4.1.1 states, the dose calculations should determine the TEDE and that TEDE is the sum of the committed effective dose equivalent (CEDE) from inhalation and the deep dose equivalent (DDE) from external exposure. Section 4.1.2 of the Regulatory Guide further explains that the exposure-to-CEDE factors for inhalation of radioactive material should be derived from the data provided in ICRP Publication 30, "Limits for Intakes of Radionuclides by Workers" and that Table 2.1 of Federal Guidance Report 11, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion" provides tables of conversion factors acceptable to the NRC staff. The factors in the column headed "effective" yield doses corresponding to the CEDE.

In a similar fashion, Section 4.1.4 of the Regulatory Guide emphasizes that the DDE should be calculated assuming submergence in semi-infinite cloud assumptions with appropriate credit for attenuation by body tissue. The DDE is nominally equivalent to the effective dose equivalent (EDE) from external exposure if the whole body is irradiated uniformly. Since this is a reasonable assumption for submergence exposure situations, EDE may be used in lieu of DDE in determining the contribution of external dose to the TEDE. Table 111.1 of Federal Guidance Report 12, "External Exposure to Radionuclides in Air, Water, and Soil", provides external EDE conversion factors acceptable to the NRC staff. The factors in the column headed "effective" yield doses corresponding to the EDE.

LGS post-LOCA direct shine dose from sources outside of the control room is dominated by a Unit 1 14" diameter core spray pipe, located 18 inches from the 36-inch thick shield wall between the control room and the Reactor Enclosure.

• The dose from the pipe has been re-evaluated based on AST-based ECCS fluid radionuclide concentrations integrated over the accident duration with standard control room occupancy credit for the 1 to 4 day and 4 to 30 day periods. Based on a review of functions required and occupancy demand near this source, it is expected that locations within the 0.22 rem isodose line can be managed by administrative controls to within this dose criterion. The 0.22 rem value is used Request for License Amendments Page 12 of 76 Alternative Source Term February 27, 2004 to characterize direct dose for the remainder of the control room. Generally, the cabinets within or near this isodose line are only needed at most for periodic monitoring, and do not require continuous operator attention.

Other sources were examined and the only external source of significance was from an RHR line located in the reactor enclosure approximately 50 feet from the control room shield wall. This source contributes an additional 0.12 rem. Other sources such as reactor enclosure airborne and external cloud and RERS, SGTS, and CREFAS filters are negligible because of shielding, distance or both.

4.3 Inputs and Assumptions 4.3.1 Accident Radiological Consequence Analyses Release Mode Releases were evaluated for full power conditions. The power level used is as described in Section 4.2.1 above for each event evaluated.

Onsite Meteorological Measurements Program The LGS meteorological measurement program meets the guidelines of Regulatory Guide 1.23 (Safety Guide 23), "Onsite Meteorological Programs" (Reference 7.11).

The tower base areas are on natural surfaces (e.g., short natural vegetation) with towers free from obstructions and micro-scale influences. This ensures that data is representative of the overall site area. The program consists of monitoring wind direction, wind speed, temperature, and precipitation. The method used for determining atmospheric stability is delta temperature (T), which measures the vertical temperature difference. These data, referenced in ANSI/ANS-2.5-1984 (Reference 7.10), are used to determine the meteorological conditions prevailing at the plant site.

The main meteorological weather tower (Tower 1) located at Weather Station No. 1 is a 280-foot tower situated approximately 3000 feet NW of the LGS structure vents. Tower 1 is also located approximately 2000 feet NNW of the center of the Unit 1 cooling tower location and approximately 2400 feet NW of the center of the Unit 2 cooling tower location. Grade elevation at Weather Station No. 1 is el 250' mean sea level (MSL).

Meteorological weather tower (Tower 2) located at Weather Station No. 2 is a 310-foot tower situated approximately 2100 feet west of the LGS structure vents. Tower 2 is also located approximately 1950 feet WSW of the center of the Unit 1 cooling tower location, and approximately 2600 feet WSW of the center of the Unit 2 cooling tower location.

The wind instruments on both towers are mounted on retractable booms extending upwind 10-0'" west of Tower 1 (WNW of Tower 2). Each face of the triangular towers is 3'-6n inches wide. The temperature sensors are located in aspirators and are 2'-0" from the tower.

Request for License Amendments Page 13 of 76 Alternative Source Term February 27, 2004 All sensors and related equipment are calibrated according to written procedures designed to ensure adherence to Regulatory Guide 1.23 guidelines for accuracy.

Calibrations occur at least every six (6) months, with component checks and adjustments performed as required.

Inspections and maintenance of all equipment is accomplished in accordance with procedures based on the instrument manufacturer's manuals. This inspection occurs at least once per week by qualified technicians capable of performing maintenance if required. In the event that the required maintenance could affect the instrument's calibration, another calibration is performed prior to returning the instrument to service.

Data from the towers are digitized and transmitted to the control room and to an on-site computer for archive storage. Periodically, all digital and analog data are sent to the approved meteorological monitoring consultant for data processing and analysis.

The digital data acquisition systems are remotely interrogated by the consultant to perform a daily quality check on system performance with the objective of identifying potential problems and to notify plant personnel as soon as possible in order to minimize down-time. This is performed each working day. All analog chart data are subject to a quality check by the consultant. This quality check consists of time continuity, instrument malfunction, inking problems, directional switching problems, negative speeds, missing data, and digitaVanalog correlation.

Data are compared with other site or regional data for consistency. If deviations occur, they are evaluated and dispositioned as appropriate. Site instrument technicians perform additional checks weekly on the instruments and collect charts for storage.

Meteorological data utilized for the calculation of new atmospheric dispersion coefficients (XIQs) were selected from the historical record of the Station meteorological monitoring tower network. Monitoring records dating back to 1972 and extending through 2002 for Tower 1 (primary tower) and Tower 2 (backup tower) were evaluated. It was desired that this calculation be based upon the continuous 5-year period that constitutes the highest Tower 1 data recovery. The period 1996-2000 was selected because it satisfies the requirements above and it represents the most complete and accurate data set that would be representative of the site meteorological data. The data was reviewed to ensure instrumentation problems and missing or anomalous observations do not affect the validity of the data. This is consistent with the NRC staff's guidance in Regulatory Guide 1.194 that considers five years of hourly observations to be representative of long-term trends.

Request for License Amendments Page 14 of 76 Alternative Source Term February 27, 2004 Tower 2 data were used only for substitution of any missing Tower 1 data as follows:

Limerick Meteorological Tower Instrument Levels (Elevation above subject tower grade)

Tower 1 (primarv) Tower 2 (backup)

Wind Speed:

Elevation 1 30 ft 159 ft Elevation 2 175 ft 304 ft Wind Direction:

Elevation 1: 30 ft 159 ft Elevation 2: 175 ft 304 ft The meteorological vendor has illustrated that the Tower 2 delta temperature data are sufficiently representative to be substituted for the Tower 1 delta temperature data; however, since the Tower 1 and Tower 2 delta temperature height intervals differ from each other somewhat, and also since for all years shown, the primary Tower 1 has data recovery rates well above the NRC's 90 percent requirements, it was deemed unnecessary to make such substitutions.

Hereinafter, the Tower 1 ARCON96 meteorological input database with applicable Tower 2 values substituted for missing Tower 1 values as indicated above will be identified as the "Tower 1 Modeling Database".

The designation of 'calm' is made to all wind speed observations 0.5 mph or less.

The higher of the starting speeds of the wind vane and anemometer equipment on each of the towers (i.e. 0.5 mph) was used as the threshold for calm winds per Regulatory Guide 1.145, Section 1.1.

Recorded meteorological data are used to generate joint frequency distributions of wind direction, wind speed, and atmospheric stability class used to provide estimates of airborne concentrations of gaseous effluents and projected offsite radiation dose.

Better than 90% data recovery is attained from each measuring and recording system.

A computerized spreadsheet was used to convert hour-by-hour delta-T data values recorded in "OF", as measured over a height range specified in "feet", into "delta-T/height" values in units of "0C/100 meters", which were then assigned the appropriate hourly stability class values as prescribed by Safety Guide 23. The narrow interval delta-T was used in the determination of the ground-level releases associated with LGS. Also, in order to provide wind speed data compatible with the ARCON96 input requirement for "wind speed times 10", raw wind speed values were reformatted within the spreadsheet by appropriately adjusting the decimal in the wind speed data, as applicable.

Wind roses and joint frequency distributions were reviewed for meteorological and climatological reasonableness and found to be acceptable prior to use. A review Request for License Amendments Page 15 of 76 Alternative Source Term February 27, 2004 was also conducted on specific hourly data prior to the execution of the atmospheric dispersion calculations in PAVAN and ARCON96. This consisted of manual spot checks of the spreadsheet reformatted data in comparison with the raw data provided by the vendor.

The LGS North and South Stacks are executed by ARCON96 as a vent release. As depicted in Attachment B, both stacks have a height of 416 ft MSL (199 ft above grade floor elevation). The stacks are located between Reactor Enclosures 1 and 2 with the North Stack situated on the north face of the buildings and the South Stack on the south face of the buildings. These stacks are less than 2.5 times the 194.75 ft high Reactor Enclosures (i.e., the highest adjacent building), and therefore, per Regulatory Guide 1.145, they are modeled as a 'vent' release.

Both the North and South Stacks are conservatively assumed to have a zero (0) flow, for which ARCON96 requires that the exit velocity and stack diameter each be assigned an input value of zero (0). Per Regulatory Guide 1.194, Table A-2, the actual building vertical cross-sectional area perpendicular to the wind direction must be utilized; therefore, the Reactor Enclosures' combined vertical cross-sectional area of 5851 m2 (calculated as height = 59.4 m, and w = 98.5 m), was input into ARCON96 to account for wake effects.

Transport Mode Atmospheric dispersion coefficients were calculated, for the identified release paths, based on site-specific meteorological data collected between 1996 through 2000 based on the current Regulatory Guide 1.145. The new dispersion coefficients developed represent a change to those used in the current UFSAR analyses. The offsite location values currently in the UFSAR are based on an early draft of Regulatory Guide 1.145.

The inleakage of unfiltered air into the control room occurs through the control room boundary, system components, and backflow at the control room doors as a result of ingress to or egress from the control room.

During the radiation mode of operation, the control room ventilation system supplies 525 cfm of filtered, outdoor air to maintain the control room at 0.1-inch water column positive pressure with respect to the adjacent areas. Intentionally admitting outdoor air into the control room facilitates reduction of infiltration through the control room boundary by assuring that air is exfiltrating from the zone at an adequate velocity (i.e., a velocity through the control room boundary to develop and maintain a pressure of 0.1-inch water column).

Automatic initiation of the radiation isolation mode is no longer credited, and manual action within 30 minutes is assumed to initiate CREFAS. During the initial 30 minutes the normal 2100 cfm of unfiltered intake is used along with an assumed additional 525 cfm of unfiltered inleakage. Since the Chlorine isolation mode could be in operation due to testing, manual transfer to the radiation isolation mode would be required. Therefore, for added conservatism, indefinite operation in the chlorine isolation mode is evaluated. This condition was demonstrated to be bounded by the results with the radiation isolation mode initiated at 30 minutes.

Request for License Amendments Page 16 of 76 Alternative Source Term February 27, 2004 During the radiation isolation mode, infiltration through the control room boundary is initially negligible because the control room will be at a positive pressure at the time of system isolation. Infiltration following isolation is assumed to be 525 cfm of unfiltered inleakage, which includes impacts of ingress and egress. The opening and closing of boundary doors can induce infiltration to the control room. However, air in-flow due to ingress and egress at LGS is minimized by the use of the door seals applied by the operators in the event of a LOCA. Installation of these seals is controlled by procedure.

The infiltration through the system components located outside the control room occurs through joints and seams in the ductwork, around damper shafts, through joints and penetrations in the air-handling units, and through the dampers that isolate the control room from non-habitable areas. The inleakage has not yet been measured via tracer gas testing. However, a tracer gas test is being scheduled as indicated in the response to Generic Letter 2003-01. This AST analysis assumes a value of 525 cfm and is a conservative estimate that should easily pass a tracer gas test since no driving force greater than the supplied intake would be expected. In the event that the tracer gas test indicates inleakage greater than 525 cfm, repairs would be implemented or recalculation performed in a timely manner to ensure control room operator doses remain within the acceptable levels described in 10CFR50.67 and General Design Criteria (GDC) 19.

Potential adverse interactions between the control room and adjacent zones that could allow the transfer of radioactive gases into the control room are minimized by maintaining the control room at a positive pressure of 0.1 inch water column with respect to adjacent areas during emergency pressurized modes. In the chlorine isolation mode (toxic gas), the control room is maintained at a neutral pressure.

During this mode, 525 cfm (normal emergency mode flow rate) is assumed to enter the control room unfiltered.

The standard breathing rates and occupancy factors used for control room personnel dose assessments and for the offsite personnel are shown in Table 1, Personnel Dose Inputs.

Removal Mode Removal mechanisms are included in the applicable event-specific discussions.

4.3.1.1 LOCA Inputs and Assumptions The key inputs and assumptions used in this analysis are included in Tables 2a through 5. These inputs and assumptions are grouped into three main categories:

release, transport, and removal.

LOCA Release Inputs Design basis Primary Containment leakage is assumed to be controlled to a La rate of 0.5% per day. For RADTRAD radioactivity transport analysis this leak rate will be Request for License Amendments Page 17 of 76 Alternative Source Term February 27, 2004 used for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and then reduced by 50% thereafter as allowed by Regulatory Guide 1.183, based on containment pressure reductions.

The entire leakage is treated as being into the secondary containment as discussed in Section 6.2 of the LGS UFSAR. For analytical purposes, the secondary containment used in the LOCA analysis refers to the reactor enclosure volume.

However, secondary containment could also encompass the refueling area. Using only the reactor enclosure volume provides conservatism. Due to the RERS recirculation fans operating after 3 minutes, credit is taken for 50% mixing in Secondary Containment. Conservatively, no credit is taken for RERS filtration during the drawdown period.

The exhaust from Secondary Containment is filtered through the SGTS filter trains, following a 15.5-minute drawdown period. After drawdown, SGTS HEPA and charcoal filters are available to reduce the release activity. The North Stack release point is treated as a zero velocity vent release (ground level equivalent) for Control Room X/Q determination, and as ground level release for offsite dose assessment.

Therefore, effectively, no elevated release is credited.

Based on the design and operation of the Containment Atmospheric Control System, and the Primary Containment Instrument Gas (PCIG) System, LGS does not routinely purge primary containment during power operations. Therefore, releases from containment purging prior to isolation during a DBA-LOCA are not considered.

High volume purging is generally only used for inerting and de-Inerting for outages.

Low volume purge lines are available for pressure or oxygen concentration control, and the PCIG System draws gas from the drywell for instrument gas to minimize pressure buildup.

Secondary containment bypass leakage potential has previously been evaluated.

These conclusions continue to apply with application of AST. The only bypass leakage paths are Containment Penetrations 7 (Primary Steam) and 8 (Primary Steam Line Drain). Because of the use of the MSIV Leakage Alternate Drain Pathway, MSIV leakage bypasses secondary containment and is released through the seismically rugged Turbine Condenser System.

The radioactivity associated with all MSIV leakage is assumed to be released directly from the Primary Containment and into the Main Steam Lines. MSIV leakage has separate limits and a separately analyzed dose assessment, therefore it is not included in the La fraction limit, and is instead separately controlled. There are no changes to La as a result of implementing AST.

MSIV leakage assumed in this accident analysis is 200 scfh total for all steam lines and 100 scfh for any one line, consistent with the current Technical Specifications.

At upstream conditions, this results in a flow rate of:

100 scfh/line

• 14.7 psia / (14.7 psia + 22 psig) / 60 min/hr = 0.668 cfm.

MSIV leakage testing is performed at 22 psig. Containment pressures above the MSIV test pressure persist for only about the first 6.5 minutes of the DBA-LOCA.

Request for License Amendments Page 18 of 76 Alternative Source Term February 27, 2004 During this limited time period very little containment air is transported into the inboard piping and even less to outboard components. Informal test runs suggest that leakage during this period results in negligible dose contributions, even if an adjustment were made to extrapolate leakage to what might be expected if MSIVs were tested at the LGS Pa of 44 psig.

However, to provide design margin, the above leak rate is increased by 25% for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to a value of 0.834 cfm. This margin also allows MSIV leakage to be reduced by 50% at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Outboard flow rates are based on expansion of this fluid from the MSIV test pressure to atmospheric pressure, and by further expansion based on worst case heating the fluid to steam line temperatures from standard temperatures. Steam line temperatures are derived based on a generic BWR evaluation crediting only conduction through pipe walls and insulation. Credit is taken for temperature reductions only at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and at 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br />.

The proposed change deletes the action statement requiring restoration of any MSIV that exceeds the 100 scfh limit to be restored to < 11.5 scfh.

This requirement is unnecessary because:

• The action is not necessary to assure that MSIV leakage remains within design basis limits.

• Maintenance goals such as this are not typically controlled by technical specifications.

• MSIV restoration leakage goals have not been required in other AST submittals.

• MSIV repairs will typically involve consistent best effort practices in order to avoid the need for rework or earlier future repairs and resulting cost and dose implications.

Flow rates out of the condenser are similarly calculated with the assumption of a condenser air space temperature of 120 'F for the accident duration.

Determination of inboard steam line, outboard steam line and condenser effective filter efficiencies is calculated, using AEB-98-03, "Assessment of Radiological Consequences for the Perry Pilot Plant Application using the Revised (NUREG-1465) Source Term", formulations and settling and deposition velocities.

For this AST evaluation an ECCS liquid leak rate of 5 gpm is used. Per Regulatory Guide 1.183, Appendix A, Section 5.2, this value is 2 times any administrative limits used as part of the Program for control of "Primary Coolant Sources Outside Containment".

Suppression Pool pH was evaluated over the 30-day duration of the DBA LOCA. It was demonstrated that pH would remain above 7. Therefore, no iodine conversion to elemental with re-evolution is expected or considered in this calculation. This control of pH also significantly limits the potential for airborne release from (always subcooled) ECCS leakage inside and outside of Secondary Containment.

Request for License Amendments Page 19 of 76 Alternative Source Term February 27, 2004 Completion of the SLC system injection of its sodium pentaborate solution is required for pH control within 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> of the start of the LOCA.

Figure 2 illustrates the "LOCA Release Pathways", with an associated Table of "LOCA Leakage Rates and Secondary Containment Mixing Parameters".

LOCA Transport Inputs The Limerick Control Room is designed with one filtered air intake. The CREFAS filtration system associated with this intake is assumed to be manually initiated within 30 minutes.

The Control Room HVAC ventilated volume is 126,000 ft3. The total flow through the CREFAS filters is 3000 cfm. In the radiation isolation mode, 525 cfm is filtered outside air, and 2475 cfm is filtered recirculation flow from the control room. In the chlorine isolation mode the entire 3000 cfm is filtered recirculation flow. The assumed intake filter efficiencies are 99% for HEPA filtration of aerosols, and 80%

for the charcoal adsorber for elemental and organic iodines. This analysis assumes, and therefore provides margin for, up to 525 cfm of unfiltered intake into the control room from the control room intake vicinity.

In the radiation isolation mode, the Control Room exfiltration is 1050 cfm. In the chlorine isolation mode, the exfiltration is 525 cfm.

During the 30 minute period before manual initiation of the CREFAS radiation isolation mode, the normal intake of 2100 cfm plus an assumed 525 cfm of unfiltered inleakage is assumed, so the exfiltration rate is 2625 cfm during this period.

LOCA Removal Inputs For LGS, the RADTRAD computer program, including the Powers Natural Deposition algorithm based on NUREG/CR-6189, is used for modeling aerosol deposition in primary containment. No natural deposition is assumed for elemental or organic iodine. The lower bound (10%) level of deposition credit is used.

Modeling of aerosol settling and elemental iodine deposition is based on methodology used by NRC in AEB-98-03. For the two steam lines modeled, two nodes are used. The first node is from the reactor pressure vessel to the inboard MSIV. The second node is from the inboard MSIV to the Turbine Stop Valve that provides the seismically designed boundary of the MSIV alternate drain pathway.

For aerosol settling, only horizontal piping runs are credited, and only the bottom surface area is considered available. Per AEB-98-03, a median settling velocity is used, given the conservatism in using a well-mixed treatment. For elemental iodine deposition both horizontal and vertical piping is credited, as well as all surfaces. This is because this deposition is not gravity dependent.

For conservatism, no credit is taken for deposition in the drain lines that provide the previously licensed MSIV alternate drain pathway to the condenser.

Request for License Amendments Page 20 of 76 Alternative Source Term February 27, 2004 All MS drain lines are routed to a single penetration in the HP condenser at a point below the condenser tubing. Iodine resuspension from settled or deposited iodines is not calculated. Historically, this phenomena increased organic iodine release by about a factor of two based on resuspension of TID-1 4844 based elemental iodine fractions. The presence of this phenomenon is questionable with aerosols with significant cesium loadings. Furthermore, while deposition on condenser tubing is not formally credited, test cases have shown that substantial removal of elemental and even organic iodine would be predicted that would more than offset any resuspension. Flow rates out of the condenser are assumed to be at 120OF and atmospheric pressure. A factor of 1.25 is applied, as is done with leakage and flow through steam lines. This leak rate is also reduced by 50% after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, consistent with the change in Containment conditions.

4.3.1.2 MSLB Accident and Assumptions The key inputs and assumptions used in the AST MSLB analysis are included in Table 6. The postulated MSLB accident assumes a double-ended break of one main steam line outside the primary containment with displacement of the pipe ends that permits maximum blowdown rates. Two activity release cases corresponding to the pre-accident spike and maximum equilibrium concentration allowed by Technical Specifications of 4.0 pCi/gm and 0.21Ci/gm dose equivalent 1-131, respectively were assumed, with inhalation CEDE dose conversion factors from Federal Guidance Report 11 and external EDE dose conversion factors from Federal Guidance Report

12. The released activity assumptions are consistent with the guidance provided in Appendix D of Regulatory Guide 1.183, as indicated in Table 6.

The analysis assumes an instantaneous ground level release. The released reactor coolant and steam are assumed to expand to a hemispheric volume at atmospheric pressure and temperature (consistent with an assumption of no Turbine Enclosure holdup credit). This hemisphere is then assumed to move at a speed of 1 meter per second downwind past the control room intake. No credit is taken for buoyant rise of the steam cloud or for decay, and dispersion of the activity of the plume was conservatively ignored. For offsite locations, the buoyant rise of the steam cloud is similarly ignored, and the ground level dispersion is based on the conservative and simplified Regulatory Guide 1.5 methodology.

The radiological consequences following an MSLB accident were determined utilizing Regulatory Guide 1.183 guidance. The following conservative assumptions were used:

• There is instantaneous release from the break to the environment. No holdup in the Turbine Enclosure or dilution by mixing with Turbine Enclosure air volume is credited.

• The activity in the steam cloud is based on the total mass of water released from the break, not just the portion that flashes to steam. This assumption is conservative because it considers the maximum release of fission products.

• The fraction of liquid water contained in steam, which carries activity into the cloud is conservatively assumed to be 2.0%.

Request for License Amendments Page 21 of 76 Alternative Source Term February 27, 2004

• The flashing fraction of liquid water released is 40%. However, all activity in the water is conservatively assumed to be released.

• No credit for control room operator action or filtration of the control room intake for the duration of the event is taken.

4.3.1.3 FHA Analysis Inputs and Assumptions The key inputs and assumptions used in the AST FHA analysis are included in Table

7. The design basis FHA involves the drop of an assembly over the reactor core to maximize the fuel damage potential due to fall height.

The postulated FHA involves the drop of a fuel assembly on top of the reactor core during refueling operations. The bounding analysis assumes that 172 GE-14 fuel rods in the full core are damaged. A radial peaking factor of 1.7 was assumed in the analysis in addition to the source term corrections discussed in Section 4.2.1. A post-shutdown 24-hour decay period was used to determine the release activity inventory. This assumption is conservative when compared to actual plant refueling outage history. The analysis assumes that gap activity in the affected rods was released instantaneously into the water in the reactor well. The bounding fuel handling accident is one in which a fuel assembly is dropped from the highest position onto the core. This produces the maximum kinetic energy, which results in the maximum damage. The drop of a fuel assembly into the spent fuel pool will not generate as much fuel damage as that due to a drop into the vessel. The analysis assumes the fuel bundle is dropped into the vessel, but only assumes a water depth of 23 feet above the assemblies seated in the reactor pressure vessel. The slightly reduced decontamination factor due to spent fuel pool water coverage of less than 23 feet is offset by less fuel damage incurred in the spent fuel pool.

In accordance with Regulatory Guide 1.183, the analysis assumes that the activity in the Reactor Enclosure environment is released within two hours, through the vent (South Stack), as a zero velocity vent release with no further credit for Reactor Enclosure holdup or dilution, or SGTS operation.

The analysis does not credit CREFAS or control room isolation.

4.3.1.4 CRDA Analysis Inputs and Assumptions The key inputs and assumptions used in the AST CRDA analysis are included in Table 8. The design basis CRDA involves the rapid removal of the highest worth control rod resulting in a reactivity excursion that encompasses the consequences of any other postulated CRDA. For the dose consequence analysis, it was assumed that 1,200 of the fuel rods in the core were damaged, with melting occurring in 0.77 percent of the damaged rods as specified in GE Report NEDE-31152P. A conservative core average radial peaking factor of 1.7 as recommended by the fuel manufacturer was used in the analysis. For releases from the breached fuel, 10% of the core inventory of noble gases and iodines are assumed to be in the fuel gap. For releases attributed to fuel melting, 100% of the noble gases and 50% of the iodines are assumed to be released to the reactor coolant.

Request for License Amendments Page 22 of 76 Alternative Source Term February 27, 2004 Instantaneous mixing of the activity released from the fuel in the reactor coolant is assumed, with 100% of the noble gases, 10% of the iodines and 1% of the remaining radionuclides that are released into the reactor coolant reach the turbine and condenser. Of this activity,100% of the noble gases,10% of the iodines and 1% of the particulate radionuclides are available for release to the environment.

The Main Condenser is assumed to leak activity into the Turbine Enclosure at a rate of 1% per day. This activity is then released, unfiltered, to the environment by way of the North Stack, taking no credit for holdup in the Turbine Enclosure.

The forced flow path through the Steam Jet Air Ejectors (SJAE) discharges to the off-gas system. This pathway is assessed crediting elimination of iodine releases and a delay of noble gas releases by the off-gas system charcoal delay beds. This credit is as currently used and licensed in conformance with NEDO-31400A.

Forced flow from the Mechanical Vacuum Pump (MVP) is prevented by trips initiated upon detection of high radiation levels by the Main Steam Line Radiation Monitors (MSLRM). Therefore, any activity in this system is held up in the Condenser, and this forced release path need not be considered.

Unfiltered release from the Turbine Enclosure is via the North Stack at the rate of 1.0% of the condenser activity per day for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

For this analysis, as performed using the RADTRAD code, the LGS Unit 1 & 2 Control Room is modeled as a closed volume of 126,000 ft3. Normal maximum flow into the Control Room of 2100 cfm, plus a conservative assumption of 1050 cfm for unknown unfiltered inleakage is used. Flow into the Control Room is therefore assumed to be 3150 cfm, and to balance the system for analytical purposes, an equal flow of air is considered to leave. No credit is taken for any filtration of flows into the Control Room.

The air that enters the Control Room originates from a source that is characterized by a dispersion factor, calculated using ARCON96. Following a CRDA, the MVPs are immediately de-energized, thereby isolating this forced flow path. LGS uses a clean steam system for sealing steam, and therefore steam seal leakage is not a forced flow path. The remaining activity, all of which is assumed to have accumulated in the condenser, leaks into the Turbine Enclosure at a rate of 1% per day. The subsequent release into the environment from the Turbine Enclosure is postulated to escape through the North Stack. The total dose in the Control Room over the 24-hour period is the result of the released activities that enter through the air intake. The methodologies significant to this analysis are the dose consequence analysis in NUREG/CR-6604 Section 2.3 and the Radioactive Decay Calculations, Section 2.4.3 of this reference.

The Exclusion Area Boundary (EAB) and Low Population Zone (LPZ) x/4's have been determined and are located 731 meters and 2043 meters respectively, from the postulated release points. Having determined these dispersion factors, the total dose is modeled in RADTRAD 3.03 using the same nodal breakdown as used for determining the Control Room total dose.

Request for License Amendments Page 23 of 76 Alternative Source Term February 27, 2004 The analysis assumptions for the transport, reduction, and release of the radioactive material from the fuel and the reactor coolant are consistent with the guidance provided in Appendix C of Regulatory Guide 1.183, as indicated in Table 8.

4.4 RESULTS 4.4.1 Evaluation Results 4.4.1.1. Accident Radiological Consequence Analyses The postulated accident radiological consequence analyses were reviewed and updated for AST implementation impact and determined not to exceed regulatory limits.

4.4.1.2 LOCA Radiological Consequence Analyses The radiological consequences of the DBA LOCA were analyzed with the RADTRAD code, using the inputs and assumptions discussed in Section 4.3.1.1 above.

The postulated sources of activity in the control room include contributions from filtered intake, and unfiltered inleakage. Dose contributors include internal cloud immersion and inhalation, and gamma shine from sources outside the Control Room.

Table 9 presents the results of the LOCA radiological consequence analysis. As indicated, the control room, EAB, and LPZ calculated doses are within the regulatory limits for implementation of AST.

The post-accident doses are the result of the following four distinct contributors:

Primary Containment to Secondary Containment Leakaqe The leakage, captured by the secondary containment (reactor enclosure), is exhausted as an unfiltered, zero-velocity vent release as analyzed during the 15.5-minute drawdown period. After this period, this activity is collected by the SGTS, and then released to the environment through the North Stack as a zero-velocity vent (ground level equivalent) release with filter credit.

The primary leakage, secondary containment bypass pathway considers piping systems from primary containment to points outside secondary containment and then to the environment. Except for MSIV leakage, no other secondary containment bypass leakage pathways have been identified for LGS.

MSIV Leakage from the Primary Containment into the Environment The MSIV leakage is released via the alternate drain pathway, condenser and turbine seals, as an unfiltered, zero-velocity vent release from the North Stack. No credit for Turbine Enclosure holdup, upward velocity, or buoyancy at the North Stack is considered.

Request for License Amendments Page 24 of 76 Alternative Source Term February 27,2004 ECCS Leakage to Secondary Containment This leakage is assumed to start immediately after the onset of a LOCA and continue for 30 days. A flashing fraction of 1.39% determined using a methodology previously approved for use at the Clinton Power Station, is used in the analysis. The flashed activity is collected by the SGTS prior to release to the environment except during the 15.5-minute drawdown period.

Dose Assessment from Sources External to Control Room The doses from the following external sources were evaluated:

• Direct dose resulting from ECCS Core Spray and RHR piping adjacent to the control room.

• Direct shine from RERS, SGTS, and CREFAS filters, Secondary Containment cloud, and external clouds are negligible due to available shielding and/or distances.

4.4.1.3 MSLB Accident Radiological Consequence Analysis The radiological consequences of the postulated MSLB are given in Table 10. As indicated, the Control Room, EAB, and LPZ calculated doses are within regulatory limits after AST implementation.

4.4.1.4 FHA Radiological Consequence Analysis The radiological consequences of the postulated FHA are given in Table 11. As indicated, the Control Room, EAB, and LPZ calculated doses are within regulatory limits after AST implementation.

4.4.1.5 CRDA Radiological Consequence Analysis The radiological consequences of the postulated CRDA are given in Table 12. As indicated, the Control Room, EAB, and LPZ calculated doses are within regulatory limits after AST implementation.

4.4.2 Atmospheric Dispersion Factors Figure 1 illustrates the release and intake points for LGS. The X/Q values for these release-intake combinations are summarized in Tables 13 and 14a.

Table 13 lists X/Q values used for the control room dose assessments. The ground level release X/Q values (i.e., LOCA-MSIV and FHA release) were calculated by the ARCON96 computer code. The separate X/Q results of each of these two models were then analyzed according to the methodology in RG-1.194. These results are based on site-specific hourly meteorological data in a five-year period of record.

Tables 14a lists X/O values for the EAB and LPZ boundaries.

Request for License Amendments Page 25 of 76 Alternative Source Term February 27, 2004 4.4.3 Post-Accident Suppression Pool Water Chemistry Management The re-evolution of elemental iodine from the Suppression Pool is strongly dependent on pool pH. The analysis assumed that the borated solution was injected within 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br /> of the onset of a DBA LOCA and mixed within the Suppression pool.

The modeling of the LGS containment cabling maximized the production of hydrochloric acid. The analysis demonstrated that the Suppression Pool pH remains above 7 for the 30-day LOCA duration. The final pH and other related parameters are presented in Table 15.

4.4.4 Evaluation Conclusions As shown in Tables 9 through 12, the plant accident radiological consequence analyses demonstrate that the post-accident offsite and Control Room doses will be maintained within regulatory limits following AST implementation. Furthermore, it has been determined that continued compliance with NUREG-0737, Item II.B.2, will be maintained and that vital areas remain accessible.

4.5

SUMMARY

Implementation of the AST as the plant radiological consequence analyses licensing basis requires a license amendment pursuant to the requirements of 10 CFR 50.67.

The above described analyses demonstrate that the offsite and control room post-accident doses remain within the regulatory limits.

Implementation of the AST provides the basis for several changes to the licensing and design bases for LGS. The principal changes affect HEPA and charcoal filtration credits, MSIV closure times, and refueling and fuel handling activities.

In the dose consequence analyses for the Control Room occupants, the assumed unfiltered inleakage was increased to a value that would be expected to bound credible inleakage values. Further evaluation of the analyses performed in support of the AST implementation support the conclusion that exposures to onsite and offsite receptors would not result in doses exceeding the values specified in 10 CFR 50.67.

Attachment 1 LGS AST LAR Page 26 of 76 February 27, 2004 Figure 1: Layout of Release Points for LGS

-J

• 1 0

Ficure 1 Parameters: Dimensional Data For Dispersion Analvses Parameter Value Distance from North Stack to Control Room Intake 16.5 m Direction, Control Room Intake to North Stack 180 degrees Distance from South Stack to Control Room Intake 64.8 m Direction, Control Room Intake to South Stack 180 degrees Elevation at Plant Grade 217 feet above mean sea level (MSL)

(216 feet outside of building)

Elevation at Center of Control Room Intake 340 feet above MSL Elevation at Top of Exhaust Stacks 416 feet above MSL LGS AST LAR Page 27 of 76 February 27, 2004 C. tU I-.

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'1' Figure 2 LGS AST LAR Page 28 of 76 February 27, 2004 Radioactivity Transport Pathways Finvirp 9 Piiraimptprq UICA paksanA RatAq

.---.- anin RAtennfirv ri3-.- nntarinmprit Mivinn Pararmptupr_

Path Description Parameters & Values Leak Rate:

LI Primary Containment Leakage to L. = 0.50 %/day, 0 - 15.5 min after start of gap release.

Secondary Containment Release is unfiltered through the North Stack during drawdown period.

La = 0.50 %/day, 15.5 min - 24hr Release is SGTS filtered through North Stack.

0.5 x La= 0.25 %Iday, 1 - 30days Release is SGTS filtered through North Stack.

PI Release from Secondary Mixing by RERS in 50% of Secondary Containment volume.

Containment to Environment RERS and SGTS filtration not credited during drawdown.

through SGTS Filter L2 MSIV Leakage to Condenser Environment Leak Rate: Based on LLRT acceptance criterion of 200 scfh for all main steam (MS) lines; 100 scfh maximum for any one MS P2 Leakage Well Mixed in HP Turbine line when measured at greater than or equal to 22 psig; two steam Shell. No credit for transport to lines are each treated as two-node well-mixed volumes each with other shells through available 100 scih flow; leak rate reduced by 50% after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

opening. No credit for substantial plateout potential on Condenser Release is from Turbine Enclosure, unfiltered through North Tubing. Stack. Analytically, flow is direct from the condenser to the North Stack.

P3 Leak from Turbine Shaft Seals to Turbine enclosure Atmosphere P4 Flow from Turbine enclosure to North Stack. No credit for mixing or holdup or deposition in Turbine enclosure. No filtration is provided for this flow.

L3 ECCS Leakage (Supp. Pool Water Leak Rate:

Source) to Secondary Containment 5 gpm, 0 - 15.5 min {2x administrative limit) and release is unfiltered through North Stack during drawdown period.

5 gpm, 15.5 min - 30days {2x administrative limit) and release is SGTS filtered through North Stack.

PI Release from Secondary Mixing by RERS in 50% of Secondary Containment volume.

Containment to Environment RERS and SGTS filtration not credited during drawdown.

through SGTS Filter RI Release to Environment and then 2100 cfm normal unfiltered intake, plus 525 cfm of unfiltered Control Room Intake from the inleakage for the first 30 minutes; 525 cfm CREFAS filtered Environment intake, plus 525 cfm of unfiltered inlcakage thereafter.

CREFAS filters credited at 99% for aerosols (based on HEPA) and 80% for charcoal absorbers (for elemental and organic iodine.)

LGS AST LAR Page 29 of 76 February 27, 2004 Figure 2 Parameters: LOCA Leakage Rates and Secondary Containment Mixing Parameters Path Description Parameters & Values Control Room Exhaust to 2625 cfm for the first 30 min of LOCA, before CREFAS is Environment initiated, and 1050 cfm thereafter, to balance with intake and inleakage R2 Release to Environment for Offsite RI and R2 include Primary Containment to Secondary Dose Assessment Purposes Containment, ECCS Leakage, and MSIV Leakage related releases

- >Ta le 1: Personne D ose Inputs ;;

-InpuutAssumption' " Value Onsite Breathing Rate 3.5E-04 m3 /sec Offsite Breathing Rate 0-8 hours: 3.5E-04 m3 /sec 8-24 hours: 1.8E-04 m3 /sec 1-30 days: 2.3E-04 m3 /sec Control Room Occupancy Factors 0-1 day: 1.0 1-4 days: 0.6 4-30 days: 0.4 1 - ^ Table  :, e An-! is nputt and Assumptions  ;

- a- le lase e-n p t L C L

Rd S ce -i ..-. -

.erm Input Assumptioni- i Vl Core Fission Product Inventory ORIGEN-2.1 Only the 60 nuclides considered by RADTRAD are utilized in the analysis Core Power Level 3,527 MWt Core Burnup 711 EFPD per 2-year cycle LGS AST LAR Page 30 of 76 February 27, 2004 Fission Product Release Fractions for RG 1.183, Table 1 LOCA BWR Core Inventory Fraction Released Into Containment Gap Early Release In-vessel Group Phase Phase Total Noble Gases 0.05 0.95 1.0 Halogens 0.05 0.25 0.3 Alkali Metals 0.05 0.20 0.25 Tellurium Metals 0.00 0.05 0.05 Ba, Sr 0.00 0.02 0.02 Noble Metals 0.00 0.0025 0.0025 Cerium Group 0.00 0.0005 0.0005 Lanthanides 0.00 0.0002 0.0002 Fission Product Release Timing RG 1.183, Table 4 (Per RG 1.183, the release phases LOCA Release Phases are modeled sequentially) BWRs Phase Onset Duration Gap Release 2 min 0.5 hr Early In-Vessel 0.5 hr 1.5 hr Table 2b Key Analysis Inputs and Assurn to!ns

& t-.i .n Relea -Non-LOCA RadionuclideSource e

- Vnp alue ; ,;- !

Core Fission Product Inventory ORIGEN-2.1 Only the 60 nuclides considered by RADTRAD are utilized in the analysis Core Power Level 3,527 MWt Core Burnup 711 EFPD (per 2-year cycle)

Fission Product Gap Release RG 1.183, Table 3 Fractions for Non-LOCA Accidents Non-LOCA Fraction of Fission Product Inventory in Gap Group Fraction 1-131 0.08 Kr-85 0.10 Other Noble Gases 0.05 Other Halogens 0.05 Alkali Metals 0.12 LGS AST LAR Page 31 of 76 February 27, 2004

,e bOCA

3

Key als lt-and unpt ions ,

. I .

I , ; *Ipt-,'-"Release lp I I . . . I I. I . . .

soPrimay and Secondary Containment Parameters

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... I . . . ., , r - I . . . . . . . . . - .. .. - ......

Input/Assumtionn ' e ' ji Value Containment Free Volume Drywell: 243,580 ft3 Suppression Pool Airspace: 159,540 ft3 Total Calculated Volume: 403,120 ft3 Minimum Suppression Pool Water Volume 118,655 cubic feet Reactor Coolant Volume 13,108 cubic feet at 552.6 0F or 9,663 cubic feet at 95.00 F Primary Containment Total Leak Rate 0.5% per day for first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (La) 0.25% per day 24 - 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> Total MSIV leak rate 200 scfh total for all 4 steam lines; 100 scfh for any single line Secondary Containment Volume 1,800,000 cubic feet Fraction of Secondary Containment Available for 0.5 Mixing SGTS Flow Rate 3,000 cfm pre-drawdown 2,500 cfm post-drawdown SGTS Filter Efficiencies HEPA: 97.5 Charcoal: 97.5 RERS Flow Rate 60,000 cfm (rated) 30,000 cfm (credited in analysis)

RERS Filter Efficiencies HEPA: 70%

Charcoal: 70%

Secondary Containment Drawdown Time 15.5 minutes Secondary Containment Bypass None, except MSIV ECCS Systems Leak Rate Outside of Primary 5 gpm Containment (includes factor of 2)

ECCS Leakage Duration 0-30 days Release Pathways Location ECCS/Containment Leakage North Stack (ground release)

MSIV Leakage North Stack (ground release)

Release Pathways Duration ECCS/Containment Leakage 0-30 days MSIV Leakage 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> to 30 days LGS AST LAR Page 32 of 76 February 27, 2004 Table 4:' 'CA'An y!Is1nputs an ssumptions

- I ais; T poart Ius~- nrbe om - ar meterS, I:I:UinUil-:;Inpqf s;. TV auemin Nuclide Release Locations See Figure 2 CREFAS System Initiation Credit for the Control Room ventilation hi-radiation signal is removed. A 30-minute operator response to isolate the Control Room HVAC system is credited.

Control Room Free Volume 126,000 cubic feet Control Room Flow Rates Normal mode 2100 cfm unfiltered intake 525 cfm unfiltered inleakage 0 cfm recirculation flow 2625 cfm exfiltration Radiation isolation mode 525 cfm filtered intake 525 cfm unfiltered inleakage 2,475 cfm filtered recirculation 1050 cfm exfiltration Chlorine isolation mode 525 cfm unfiltered inleakage 3000 cfm filtered recirculation 525 cfm exfiltration Elemental and Organic Iodine Removal Efficiencies 80%

Aerosols Removal Efficiency 99%

LGS AST LAR Page 33 of 76 February 27, 2004

,.,Table 5: e LOCA'Anais IpIts and Assu-mptions, oval VT, i .7 , np' tpAss tion A-TVa lue-Containment Spray removal Not credited Aerosol Natural Deposition Coefficients Used in the Credit is taken for natural deposition of Containment aerosols based on equations for the Power's model in NUREG/CR 6189 and input directly by RADTRAD as natural deposition time dependent lambdas.

No credit is assumed for natural deposition of elemental or organic iodine, or for suppression pool scrubbing.

SGTS Filter Efficiencies - Elemental and Organic SGTS HEPA filters and charcoal adsorbers Iodine Aerosols are credited after drawdown.

HEPA: 97.5%

Charcoal: 97.5%

RERS Filter Efficiencies HEPA: 70%

Charcoal: 70%

Deposition/Plate-out (where credited) Deposition based on AEB-98-03 well-mixed model and associated median settling velocity. Only horizontal piping is credited, and the bottom half as the settling area. For elemental iodine, deposition velocities from AEB-98-03 are used and all piping and surfaces credited.

Main Steam Line and Condenser Holdup Modeling of aerosol settling and elemental Holdup Credit for MSIV Leakage iodine deposition is based on methodology used by NRC in AEB-98-03. For the two steam lines modeled, two nodes are used.

The first node is from the reactor pressure vessel to the inboard MSIV. The second node is from the inboard MSIV to the Turbine Stop Valve that provides the seismically designed boundary of the MSIV Leakage Control System. For aerosol settling, only horizontal piping runs are credited, and only the bottom surface area is considered available. Per AEB-98-03, a median settling velocity is used, given the conservatism in using a well-mixed treatment. For elemental iodine deposition both horizontal and vertical piping is credited, as well as all surfaces.

This is because this deposition is not gravity dependent LGS AST LAR Page 34 of 76 February 27, 2004

>'Table6:"Key Tab 6j MXL A ccidentAnalysis

' .6-~II - . inputs' . ndAssumptions.-

. ... , a..  :-- .... .

.- ,InpuVA tl sum IAItion i  :-  ; Value --

Break Discharge Mass Release For a 5.5 second MSIV closure time:

103,785 pounds (20,452 as steam and 83,333 as liquid)

For a 10.5 second MSIV closure time:

206,933 pounds (20,452 as steam and 186,481 as liquid)

Pre-Accident Spike Iodine Concentration 4.0 gCVgm 1-131 equivalent Maximum Equilibrium Iodine Concentration 0.2 [tCi/gm 1-131 equivalent Transport Model for Control Room Steam cloud moves past the Control Room intake at 1 m/sec Turbine Enclosure Holdup/Control Room Filtration Not credited

i. -. Tale 7: Key FHA`A.a.y..

FA AbalysIs Inpults and Assumptions ' ` .  ;;

Input/Assumption i l - ' Value Core Damage 172 fuel rods failed based on GE14 fuel and "Heavy Mast" Radial Peaking Factor 1.7 Fuel Decay Period 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Iodine Decontamination Factor DF = 200 Release Period 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Refuel Floor Air Removal Rate 6 air changes per hour to assure activity exhaust within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Control Room Filtration Not Credited Control Room Intake Flow 2100 cf in normal intake plus 1050 cfm unfiltered inleakage Release Location South Stack unfiltered, zero-velocity vent release (Ground Level equivalent)

CREFAS/SGT System Initiation Not Credited LGS AST LAR Page 35 of 76 February 27, 2004 Table 8: K CRDA Aalyis Inputs and A-su ti s;

-lnpuVtAssumption X. -

X Value'

Core Damage 1,200 fuel rods failed (66,720 fuel rods in core)

Percent of Damaged Fuel with Melt 0.77%

Radial Peaking Factor 1.7 Mechanical Vacuum Pump (MVP) Operation MVP isolation by MSLRM Hi Rad signal Before Isolation Condenser Leak Rate 1% of condenser activity per day throughout entire release period Release Period 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Forced Flow Paths None. Main Steam Line Radiation Monitor high radiation causes Mechanical Vacuum Pump trip and isolation, if required. LGS is a "clean sealing steam" plant so gland sealing steam is not a forced flow path.

If no reactor isolation (MSIVs are not closed), the forced flow path is through the Steam Jet Air Ejectors (SJAE) discharge to the off-gas system.

This pathway is assessed crediting elimination of iodine releases and a delay of noble gas releases by the off-gas system charcoal delay beds. This credit is as currently used and licensed in conformance with NEDO-31400A.

Release Location See Figure 1 Zero-velocity vent release through the North Stack CREFAS System Initiation Not credited At -*d'l o 4 'Table 9.: LOCA RadIogica Ic onseque ce alyis

'Location R -eglator'T.r Li '

Duration'- TEDE (rm)TEDE (Jremn)-

Control Room 30 days 4.01* 5 EAB Maximum, 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 0.90 25 LPZ 30 days 1.25 25

• The doses here are for the bounding radiation isolation mode with 525 cfm of filtered fresh air intake for pressurization and 525 cfm of unfiltered inleakage. The calculated total Control Room dose for the unpressurized chlorine isolation mode with 525 cfm of unfiltered inleakage is lower.

LGS AST LAR Page 36 of 76 February 27, 2004

. Table 10: MSLB Ac identR adiological Consequence"nalysis ,

'(10.5 Secnd!MSIV Closure Time)

'-.. .,,40 gCi/g nDose 0.2 'Ci/gm Dose' Equivalent 131: --. Equiqvalent 1-131 Reguatory 'Liit Location Duration .; ' TD (rem TEDE (rem)

Control 30-day 3.61 0.18 5 Room integrated dose EAB Worst 2.82 0.14 25 (4.0 tiCi/gm) intgra 2.5 (0.2 ,Ci/gm) integrated dose LPZ 30-day 1.11 0.056 25 (4.0 ItCVgm) integrated 2.5 (0.2 gCigm) dose

.,-Table 11i F.HA Raiologica Consequenc AnalysiS '

,Location ., u r Limit Control Room 30 days 2.52 5 EAB Maximum, 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 0.88 6.3 LPZ 30 days 0.32 6.3

-Tabl '12': CRD Radiological Con qu ce Ana lyis

i,'Regulatory imit

,Location ,Duration p 'EDE (e) TEDE (remj  ;

Control Room 30 days 1.62 5 EAB Maximum, 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 0.049 6.3 LPZ 30 days 0.034 6.3 LGS AST LAR Page 37 of 76 February 27, 2004

.- I l. '.3:'-;Contito R 6. 'RIC':f.-

DifferentR'elease~ nd ntake Combinaioni 2

~'XIQ (ec/rn)m_ _ _ _ _ _

-'.'Time-<' . ,f

• Period N,,.c FHA CD -A 0 - 2 hrs 6.88E-03 1.26E-03 6.88E-03 2 - 8 hrs 5.17E-03 5.17E-03 8 - 24 hrs 2.04E-03 2.04E-03 1 - 4 days 1.29E-03 4-30 days 9.63E-04 Notes:

1.Control room intake X/Q values are applicable for control room inleakage.

2.For MSLB, specific Control Room X/Q values are not calculated. For conservatism, the calculated hemispherical steam plume volume is transported, without dilution, over the Control Room intake location for the duration required for plume transit at a wind speed of 1 meter per second.

-.- :Table I14a: -North &'South'Stack (Ground Level Release) *?e,,t

' .; r,<, L5 Fdr LOCA -FHA; and.CRDA X -

'Q (secim 3 )TimePerid *2EAB Values'Lsin RG .l45 Methodoloav forth'"AB J~~iEII) . EAR~*<Ij>LZ X LPZ and

__________P______________ __________ (__'_ _:_ (sec/rn 0 - 2 hrs 3.18E-04 1.15E-04 0 - 8 hrs 5.79E-05 8 - 24 hrs 4.1 OE-05 1 - 4 days 1.95E-05 4 - 30 days 6.68E-06 Notes:

1. For MSLB, offsite X/Q values are determined using Regulatory Guide 1.5 methodology, and are 4.77E-04 sec/m 3 for the EAB and 1.89E-04 sec/m 3 for the LPZ.

Table 15

Su'pression Pool pH Results

,i,(Based on a otf

. ,~~~~~~t t .r n e g t 24 ,l. m ,...........

>Vauonedition c ' ,e Initial Suppression Pool pH 5.3 SLC injection time Required Sodium Pentaborate to be injected within 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br /> Suppression Pool pH throughout the 30-day Greater than 7 accident duration (with SLC injection)

Attachment 1 LGS AST LAR Page 38 of 76 February 27, 2004 REGULATORY GUIDE 1.183 COMPARISON

.Table:A: .Conformance with Reaulatorv.Guide (RG) 1.183 Main'Sections"A i%,s -. , ' , ,K'U !'i;. , ,

RG RG.Position  ;-;x,-. T -;-., LGS - Comments'. . -

Section -^i- Analysis: _ _ _ _ _ _ _ _

3.1 The inventory of fission products in the reactor core and available Conforms ORIGEN 2.1 based methodology was for release to the containment should be based on the maximum used to determine core inventory. These full power operation of the core with, as a minimum, current source terms were evaluated at end-of-licensed values for fuel enrichment, fuel burnup, and an assumed cycle and at beginning of cycle (100 core power equal to the current licensed rated thermal power times effective full power days (EFPD) to the ECCS evaluation uncertainty. The period of irradiation should achieve equilibrium) conditions and worst be of sufficient duration to allow the activity of dose-significant case inventory used for the selected radionuclides to reach equilibrium or to reach maximum values. isotopes. These values were then The core inventory should be determined using an appropriate converted to units of CVMWt. Accident isotope generation and depletion computer code such as ORIGEN analyses are based on a 3527 MWt power 2 or ORIGEN-ARP. Core inventory factors (CVMWt) provided in level, based on the current accident TID 14844 and used in some analysis computer codes were analysis design basis allowance for derived for low burnup, low enrichment fuel and should not be used instrument uncertainty.

with higher burnup and higher enrichment fuels. Source terms are based on a 2 year fuel cycle with a nominal 711 EFPD per cycle.

3.1 For the DBA LOCA, all fuel assemblies in the core are assumed to Conforms Peaking factors of 1.7 are used for DBA be affected and the core average inventory should be used. For events that do not involve the entire core, DBA events that do not involve the entire core, the fission product with fission product inventories for inventory of each of the damaged fuel rods is determined by damaged fuel rods determined by dividing dividing the total core inventory by the number of fuel rods in the the total core inventory by the number of core. To account for differences in power level across the core, fuel rods in the core.

radial peaking factors from the facility's core operating limits report (COLR) or technical specifications should be applied in determining the inventory of the damaged rods.

3.1 No adjustment to the fission product inventory should be made for Conforms No adjustments for less than full power are events postulated to occur during power operations at less than full made in any analyses.

rated power or those postulated to occur at the beginning of core life. For events postulated to occur while the facility is shutdown, e.g., a fuel handling accident, radioactive decay from the time of shutdown may be modeled.

3.2 The core inventory release fractions, by radionuclide groups, for the Conforms The fractions from Regulatory Position 3.1, gap release and early in-vessel damage phases for DBA LOCAs I Table 1 are used.

Attachment 1 LGS AST LAR Page 39 of 76 February 27, 2004 Table A: _Conformanc. wit. _Reaulatorv Gude (RG ...1..83Man ._.ions............. = . ..

RG o. -- GPosition .  ;  ; mments,.

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• Y Analysis ^, .  : .

are listed in Table 1 for BWRs and Table 2 for PWRs. These fractions are applied to the equilibrium core inventory described in Footnote 10 criteria are met.

Regulatory Position 3.1.

Table 1 BWR Core Inventory Fraction Released Into Containment Gap Early Release In-Vessel Group Phase Phase Total Noble Gases 0.05 0.95 1.0 Halogens 0.05 0.25 0.3 Alkali Metals 0.05 0.20 0.25 Tellurium Metals 0.00 0.05 0.05 Ba, Sr 0.00 0.02 0.02 Noble Metals 0.00 0.0025 0.0025 Cerium Group 0.00 0.0005 0.0005 Lanthanides 0.00 0.0002 0.0002 Footnote 10:

The release fractions listed here have been determined to be acceptable for use with currently approved LWR fuel with a peak rod burnup up to 62,000 MWD/MTU. The data in this section may not be applicable to cores containing mixed oxide (MOX) fuel.

LGS AST LAR Page 40 of 76 February 27, 2004 Table A:' Conformanbeewith Regulatory Guide'(RG) 1.183Main Sectionsh ;' --

RG RGPosition -. -.-. -LGSS Comments-Sect" '. Anal siys:'.;

3.2 For non-LOCA events, the fractions of the core inventory assumed Conforms Complies with Note 11 of Table 3.

to be in the gap for the various radionuclides are given in Table 3.

The release fractions from Table 3 are used in conjunction with the Peaking factor of 1.7 used for DBA events fission product inventory calculated with the maximum core radial that do not involve the entire core.

peaking factor.

Table 3 Non-LOCA Fraction of Fission Product Inventory in Gap Group Fraction 1-131 0.08 Kr-85 0.10 Other Noble Gases 0.05 Other Halogens 0.05 Alkali Metals 0.12 Footnote 11:

The release fractions listed here have been determined to be acceptable for use with currently approved LWR fuel with a peak burnup up to 62,000 MWD/MTU provided that the maximum linear heat generation rate does not exceed 6.3 kw/ft peak rod average power for rods with burnups that exceed 54 GWD/MTU. As an alternative, fission gas release calculations performed using NRC-approved methodologies may be considered on a case-by-case basis. To be acceptable, these calculations must use a projected power history that will bound the limiting projected plant-specific power history for the specific fuel load. For the BWR rod drop accident and the PWR rod ejection accident, the gap fractions are assumed to be 10% for iodines and noble gases.

3.3 Table 4 tabulates the onset and duration of each sequential release Conforms The BWR durations from Table 4 are phase for DBA LOCAs at PWRs and BWRs. The specified onset is used.

the time following the initiation of the accident (i.e., time = 0). The LOCA is modeled in a linear fashion.

early in-vessel phase immediately follows the gap release phase. Non-LOCA is modeled as an The activity released from the core during each release phase instantaneous release.

should be modeled as increasing in a linear fashion over the duration of the phase. For non-LOCA DBAs, in which fuel damage

Attachment 1 LGS AST LAR Page 41 of 76 February 27, 2004 Table A: Conformance With.Riaulator Guide (RG)Y1.183 Main Sections-\ I r%*. ' - I

+ . .I,

'> . --' . - . e RG RG Positionw, i'>t . ', . .;t -r LGS. Commens-  ; .. - ---.

Section . _

t'^-"

~~~~~~~~~~~~. .-

~.. pi

'iL^',.;,

' ".'m',^;.,'\>;t'A is projected, the release from the fuel gap and the fuel pellet should be assumed to occur instantaneously with the onset of the projected damage.

Table 4 LOCA Release Phases PWRs BWRs Phase Onset Duration Onset Duration Gap Release 30 sec 0.5 hr 2 min 0.5 hr Early In-Vessel 0.5 hr 1.3 hr 0.5 hr 1.5 hr 3.3 For facilities licensed with leak-before-break methodology, the Not LGS does not use leak-before-break onset of the gap release phase may be assumed to be 10 minutes. Applicable methodology for DBA analyses.

A licensee may propose an alternative time for the onset of the gap release phase, based on facility-specific calculations using suitable analysis codes or on an accepted topical report shown to be applicable for the specific facility. In the absence of approved alternatives, the gap release phase onsets in Table 4 should be 4-used. - .1, 3.4 Table 5 lists the elements in each radionuclide group that should be Conforms The nuclides used are the 60 identified as considered in design basis analyses. being potentially important dose Table 5 contributors to total effective dose Radionuclide Groups equivalent (TEDE) in the RADTRAD code, Group Elements which encompasses those listed in Noble Gases Xe, Kr RG 1.183, Table 5.

Halogens I, Br Alkali Metals Cs, Rb Tellurium Group Te, Sb, Se, Ba, Sr Noble Metals Ru, Rh, Pd, Mo, Tc, Co Lanthanides La, Zr, Nd, Eu, Nb, Pm, Pr, Sm, Y, Cm, Am Cerium Ce. Pu. ND 3.5 Of the radioiodine released from the reactor coolant system (RCS) Conforms This guidance is applied in the analyses.

to the containment in a postulated accident, 95 percent of the iodine released should be assumed to be cesium iodide (Csl), 4.85 percent elemental iodine, and 0.15 percent organic iodide. This

Attachment 1 LGS AST LAR Page 42 of 76 February 27, 2004 Table A: ConforrnanceWith Reaulatorv Guide (RG).1-.183 Main Sections ,'-'7,-, A If at . N

-- I -- ' -l --;  ;

RGRGPosition

_ _ _ _ _ _ _ _ _.,A Secton <j;..*.

.~ . ~ ~ .. ...

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-  ;'AnaYS is~ Iomments a i'

{ _ _ _ _ _ _ _ _ _ _ _

i i

includes releases from the gap and the fuel pellets. With the exception of elemental and organic iodine and noble gases, fission products should be assumed to be in particulate form. The same chemical form is assumed in releases from fuel pins in FHAs and from releases from the fuel pins through the RCS in DBAs other than FHAs or LOCAs. However, the transport of these iodine species following release from the fuel may affect these assumed fractions. The accident-specific appendices to this regulatory guide nrovide additional details.

3.6 The amount of fuel damage caused by non-LOCA design basis Conforms Fuel damage assessment for CRDA and events should be analyzed to determine, for the case resulting in FHA are based on GESTAR standard the highest radioactivity release, the fraction of the fuel that reaches analyses for GE14 fuel.

or exceeds the initiation temperature of fuel melt and the fraction of fuel elements for which the fuel clad is breached. Although the NRC staff has traditionally relied upon the departure from nucleate boiling ratio (DNBR) as a fuel damage criterion, licensees may propose other methods to the NRC staff, such as those based upon enthalpy deposition, for estimating fuel damage for the purpose of establishing radioactivity releases.

4.1.1 The dose calculations should determine the TEDE. TEDE is the Conforms TEDE is calculated, with significant sum of the committed effective dose equivalent (CEDE) from progeny included.

inhalation and the deep dose equivalent (DDE) from external exposure. The calculation of these two components of the TEDE should consider all radionuclides, including progeny from the decay of parent radionuclides that are significant with regard to dose consequences and the released radioactivity.

4.1.2 The exposure-to-CEDE factors for inhalation of radioactive material Conforms Federal Guidance Report 11 dose should be derived from the data provided in ICRP Publication 30, conversion factors (DCFs) are used.

"Limits for Intakes of Radionuclides by Workers" (Ref. 19). Table 2.1 of Federal Guidance Report 11, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion" (Ref. 20),

provides tables of conversion factors acceptable to the NRC staff.

Attachment 1 LGS AST LAR Page 43 of 76 February 27, 2004 Table A:.Conformance with.Reaulatorv Guide (RGY1;183 MainSections"x; n;  :-it I -  :..

RG RG Pos'ition-'_- 7'Y,'<. LGS Comments:.2 Sect* * - - Aa-s- - ..

The factors in the column headed "effective" yield doses corresponding to the CEDE.

4.1.3 For the first 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, the breathing rate of persons offsite should be Conforms The values that correspond to the rounded assumed to be 3.5 x 1O4 cubic meters per second. From 8 to 24 values in Section 4.1.3 of RG 1.183 are hours following the accident, the breathing rate should be assumed used.

to be 1.8 x 10' cubic meters per second. After that and until the end of the accident, the rate should be assumed to be 2.3 x 104 cubic meters per second.

4.1.4 The DDE should be calculated assuming submergence in semi- Conforms Federal Guidance Report 12 conversion infinite cloud assumptions with appropriate credit for attenuation by factors are used.

body tissue. The DDE is nominally equivalent to the effective dose equivalent (EDE) from external exposure if the whole body is irradiated uniformly. Since this is a reasonable assumption for submergence exposure situations, EDE may be used in lieu of DDE in determining the contribution of external dose to the TEDE. Table 111.1 of Federal Guidance Report 12, "External Exposure to Radionuclides in Air, Water, and Soil" (Ref. 21), provides external EDE conversion factors acceptable to the NRC staff. The factors in the column headed "effective" yield doses corresponding to the EDE.

4.1.5 The TEDE should be determined for the most limiting person at the Conforms The maximum two-hour LOCA EAB dose EAB. The maximum EAB TEDE for any two-hour period following starts as follows:

the start of the radioactivity release should be determined and used in determining compliance with the dose criteria in 10 CFR 50.67. PC Leakage: 3.2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> The maximum two-hour TEDE should be determined by calculating MSIV Leakage: 10.4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> the postulated dose for a series of small time increments and ECCS Leakage: 3.2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> performing a "sliding" sum over the increments for successive two-hour periods. The maximum TEDE obtained is submitted. The time Conservatively, the maximum 2-hour increments should appropriately reflect the progression of the period dose was determined by adding the accident to capture the peak dose interval between the start of the maximum 2-hour dose for each of the event and the end of radioactivity release (see also Table 6). components listed above even though they do not occur simultaneously.

Footnote 14:

With regard to the EAB TEDE, the maximum two-hour value is the LGS AST LAR Page 44 of 76 February 27, 2004 Table A:, Conformance with Regulatory Guide (RG) .1 '183'Main Sections::.,. -' ,-." , .- i '

RG >RG position . :LGS 'Comments'. ; "'--':

Section  : . - . Analysis basis for screening and evaluation under 10 CFR 50.59. Changes to doses outside of the two-hour window are only considered in the context of their impact on the maximum two-hour EAB TEDE.

4.1.6 TEDE should be determined for the most limiting receptor at the Conforms This guidance is applied in the analyses.

outer boundary of the low population zone (LPZ) and should be used in determining compliance with the dose criteria in 10 CFR 50.67.

4.1.7 No correction should be made for depletion of the effluent plume by Conforms No such corrections made in the analyses.

deposition on the ground.

4.2.1 The TEDE analysis should consider all sources of radiation that will Conforms The principal source of dose within the cause exposure to control room personnel. The applicable sources control room is due to airborne activity.

will vary from facility to facility, but typically will include: The dose estimates from post LOCA Contamination of the control room atmosphere by the intake or primary containment and sources external infiltration of the radioactive material contained in the radioactive to the control room indicate that plume released from the facility, contribution to dose is dominated by Contamination of the control room atmosphere by the intake or ECCS piping in the Reactor Enclosure infiltration of airborne radioactive material from areas and structures adjacent to the Control Room.

adjacent to the control room envelope, Radiation shine from the external radioactive plume released from the facility, Radiation shine from radioactive material in the reactor containment, Radiation shine from radioactive material in systems and components inside or external to the control room envelope, e.g.,

radioactive material buildup in recirculation filters.

4.2.2 The radioactive material releases and radiation levels used in the Conforms The source term, transport, and release control room dose analysis should be determined using the same methodology is the same for both the source term, transport, and release assumptions used for control room and offsite locations.

determining the EAB and the LPZ TEDE values, unless these assumptions would result in non-conservative results for the control room.

Attachment 1 LGS AST LAR Page 45 of 76 February 27, 2004 Table A: Conformance with Reablator Guide (RG) 1183 Main Sections-r-.- ;i .>..;'-'

RG Position;-- '- - -' -- --L LGS - Comments -

Section' -- i'  :

i.'*- -' - Analysis - -

4.2.3 The models used to transport radioactive material into and through Conforms This guidance is applied in the analyses.

the control room, and the shielding models used to determine radiation dose rates from external sources, should be structured to provide suitably conservative estimates of the exposure to control room personnel.

4.2.4 Credit for engineered safety features that mitigate airborne Conforms For the LOCA (after the drawdown period),

radioactive material within the control room may be assumed. Such credit is taken for SGTS HEPA and features may include control room isolation or pressurization, or charcoal adsorber filtration (97.5% each) intake or recirculation filtration. Refer to Section 6.5.1, "ESF and RERS HEPA and charcoal filtration Atmospheric Cleanup System," of the SRP (Ref. 3) and Regulatory (70% for each).

Guide 1.52, "Design, Testing, and Maintenance Criteria for Post-accident Engineered-Safety-Feature Atmosphere Cleanup System Control Room and intake and recirculation Air Filtration and Adsorption Units of Light-Water-Cooled Nuclear filtration by CREFAS are credited in the Power Plants" (Ref. 25), for guidance. LOCA accident analysis. Radiation isolation mode has been analyzed with manual initiation within 30 minutes. After this period, credit is taken for 99% HEPA and 80% charcoal adsorber efficiencies.

No filtration credit is taken in the FHA, MSLB or CRDA accident analyses.

4.2.5 Credit should generally not be taken for the use of personal Conforms Such credits are not taken.

protective equipment or prophylactic drugs. Deviations may be considered on a case-by-case basis.

4.2.6 The dose receptor for these analyses is the hypothetical maximum Conforms Standard occupancy factors and breathing exposed individual who is present in the control room for 100% of rate are used throughout the analyses.

the time during the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the event, 60% of the time between 1 and 4 days, and 40% of the time from 4 days to 30 days.

For the duration of the event, the breathing rate of this individual should be assumed to be 3.5 x 10 cubic meters per second.

4.2.7 Control room doses should be calculated using dose conversion Conforms The equation given is utilized for finite factors identified in Regulatory Position 4.1 above for use in offsite cloud correction when calculating external dose analyses. The DDE from photons may be corrected for the doses due to the airborne activity inside

Attachment 1 LGS AST LAR Page 46 of 76 February 27, 2004 Tnhl rA* 1 flln S * wt IV S ISJ8LA:I S )' A -1A*

I . 55 . ^ - I -,l . X -S.  : N - . - .: I - .. - I - , - I RG: RG Position LGS :Comments. :-

Section - - ;Analysis ' - ..-

difference between finite cloud geometry in the control room and the control room.

the semi-infinite cloud assumption used in calculating the dose conversion factors. The following expression may be used to correct the semi-infinite cloud dose, DDE., to a finite cloud dose, DDEfinite , where the control room is modeled as a hemisphere that has a volume, V, in cubic feet, equivalent to that of the control room (Ref. 22).

DDE = DDEV 0 '338 D flt = 1173 _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

4.3 The guidance provided in Regulatory Positions 4.1 and 4.2 should Conforms Based on an evaluation, the existing TID-be used, as applicable, in re-assessing the radiological analyses 14844 based analyses included in section identified in Regulatory Position 1.3.1, such as those in NUREG- 1.13 of the UFSAR are shown to be 0737 (Ref. 2). Design envelope source terms provided in NUREG- conservative and bounding. Given 0737 should be updated for consistency with the AST. In general, compliance with the GDC-19 limit of 5 radiation exposures to plant personnel identified in Regulatory REM when dose is based on TID-1 4844 Position 1.3.1 should be expressed in terms of TEDE. Integrated source terms, compliance with 10 CFR radiation exposure of plant equipment should be determined using 50.67 Control Room dose limits can be the guidance of Appendix I of this guide. expected with the AST-based analysis.

Therefore, the historically analyzed cases are sufficient and no additional analysis of vital areas of LGS are necessary.

5.1.1 The evaluations required by 10 CFR 50.67 are re-analyses of the Conforms These analyses were prepared as design basis safety analyses and evaluations required by 10 CFR specified in the guidance.

50.34; they are considered to be a significant input to the evaluations required by 10 CFR 50.92 or 10 CFR 50.59. These analyses should be prepared, reviewed, and maintained in accordance with quality assurance programs that comply with Appendix B, "Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants," to 10 CFR Part 50.

5.1.2 Credit may be taken for accident mitigation features that are Conforms The analyses take credit for SLC System classified as safety-related, are required to be operable by technical based on operation. The SLC System is safety-specifications, are powered by emergency power sources, and are acceptable related, required to be operable by either automatically actuated or, in limited cases, have actuation assessments Technical Specifications, and supplied with

Attachment 1 LGS AST LAR Page 47 of 76 February 27, 2004

.TableA: Conformancewith.Relatorv Guide,(RG1.183 MaiSections. .. ,, . , . , ,

RG ~. RGP PoiinGS..~ . ...

Section, AaLGsi .s requirements explicitly addressed in emergency operating emergency power. The SLC System is procedures. The single active component failure that results in the manually initiated from the main control most limiting radiological consequences should be assumed. room, as directed by the emergency Assumptions regarding the occurrence and timing of a loss of operating procedures. Due to having a offsite power should be selected with the objective of maximizing common flow path with inline check valves the postulated radiological consequences. located inside containment, SLC is not fully single-failure proof although it has a high level of redundancy regarding system flow paths and active components (e.g.,

multiple pumps, suction paths, and explosive iniection valves).

5.1.3 The numeric values that are chosen as inputs to the analyses Conforms Conservative assumptions are used.

required by 10 CFR 50.67 should be selected with the objective of determining a conservative postulated dose. In some instances, a particular parameter may be conservative in one portion of an analysis but be non-conservative in another portion of the same analysis.

5.1.4 Licensees should ensure that analysis assumptions and methods Conforms Analysis assumptions and methods were are compatible with the AST and the TEDE criteria, made per this guidance.

5.3 Atmospheric dispersion values (X/Q) for the EAB, the LPZ, and the Conforms New atmospheric dispersion values (X/Q) control room that were approved by the staff during initial facility for the EAB, the LPZ, and the control room licensing or in subsequent licensing proceedings may be used in were developed, using meteorological data performing the radiological analyses identified by this guide. for the years 1996-2000. ARCON96 and Methodologies that have been used for determining X/Q values are PAVAN were used with these data to documented in Regulatory Guides 1.3 and 1.4, Regulatory Guide determine control room and EAB/LPZ 1.145, "Atmospheric Dispersion Models for Potential Accident atmospheric dispersion values. Control Consequence Assessments at Nuclear Power Plants," and the room X/Qs from releases from the North paper, "Nuclear Power Plant Control Room Ventilation System and South Stacks were developed in Design for Meeting General Criterion 19". conformance with RG-1.194.

The NRC computer code PAVAN implements Regulatory Guide 1.145 and its use is acceptable to the NRC staff. The methodology of the NRC computer code ARCON96 is generally acceptable to the NRC staff for use in determining control room X/Q values.

LGS AST LAR Page 48 of 76 February 27, 2004 TableB:. Conformance with RG1.'183AppendixA(Loss-of-Coolant'Accident) _ -

RGG RGPosition ' . LGS. -Comments;",

Section l A sis Acceptable assumptions regarding core inventory and the release of Conforms Fission Product Inventory: Core radionuclides from the fuel are provided in Regulatory Position 3 of this source terms are developed using guide. ORIGEN-2.1 based methodology.

Release Fractions: Release fractions are per Table 1 of RG 1.183, and are implemented by RADTRAD.

Timing of Release Phases: Release Phases are per Table 4 of RG 1.183,.

and are implemented by RADTRAD.

Radionuclide Composition:

Radionuclide grouping is per Table 5 of RG 1.183, as implemented in RADTRAD.

Chemical Form: Treatment of release chemical form is per RG 1.183, Section 3.5.

2 If the sump or suppression pool pH is controlled at values of 7 or Conforms The stated distributions of iodine greater, the chemical form of radioiodine released to the containment chemical forms are used.

should be assumed to be 95% cesium iodide (Csl), 4.85 percent The post-LOCA suppression pool pH elemental iodine, and 0.15 percent organic iodide. Iodine species, has been evaluated, including including those from iodine re-evolution, for sump or suppression pool consideration of the effects of acids pH values less than 7 will be evaluated on a case-by-case basis. and bases created during the LOCA Evaluations of pH should consider the effect of acids and bases created event, the effects of key fission during the LOCA event, e.g., radiolysis products. With the exception of product releases, and the impact of elemental and organic iodine and noble gases, fission products should SLC injection. Suppression pool pH be assumed to be in particulate form. remains above 7 for at least 30 days.

3.1 The radioactivity released from the fuel should be assumed to mix Conforms The radioactivity release from the instantaneously and homogeneously throughout the free air volume of fuel is assumed to instantaneously the primary containment in PWRs or the drywell in BWRs as it is and homogeneously mix throughout released. This distribution should be adjusted if there are internal the drywell and suppression chamber compartments that have limited ventilation exchange. The suppression air space. The suppression chamber pool free air volume may be included provided there is a mechanism to free air volume is included based on ensure mixing between the drywell to the wetwell. The release into the expected steam flow from the drywell LGS AST LAR Page 49 of 76 February 27, 2004 Table'B: Conformance-with' RG'1.183AdcendixA'(Losst-of-Coolant"Ac'idenftl t --"' v - .S _^*- e - ..I ":

- , e RG nRG Position MAS-it , ,, Comm s

_ _ .A -. Analysis v>. . a.-.

containment or drywell should be assumed to terminate at the end of the to the suppression chamber, even early in-vessel phase. after the initial blowdown, and from the suppression chamber to the drywell through vacuum breakers as steam condensing reduces drywell pressure relative to that in the suppression chamber.

3.2 Reduction in airborne radioactivity in the containment by natural Conforms Credit is taken for natural deposition deposition within the containment may be credited. Acceptable models per the methodology of NUREG/CR-for removal of iodine and aerosols are described in Chapter 6.5.2, 6189, as implemented in RADTRAD.

"Containment Spray as a Fission Product Cleanup System," of the No deterministically assumed initial Standard Review Plan (SRP), NUREG-0800 (Ref. A-1) and in plateout is credited.

NUREG/CR-6189, "A Simplified Model of Aerosol Removal by Natural Processes in Reactor Containments" (Ref. A-2). The latter model is incorporated into the analysis code RADTRAD (Ref. A-3).

3.3 Reduction in airborne radioactivity in the containment by containment Not While containment spray is an spray systems that have been designed and are maintained in Applicable available design feature at LGS, no accordance with Chapter 6.5.2 of the SRP (Ref. A-1) may be credited. credit is taken for airborne activity Acceptable models for the removal of iodine and aerosols are described removal by sprays in the LOCA AST in Chapter 6.5.2 of the SRP and NUREG/CR-5966, "A Simplified Model analysis.

of Aerosol Removal by Containment Sprays"1 (Ref. A-4). This simplified model is incorporated into the analysis code RADTRAD (Refs. A-1 to A-3).

The evaluation of the containment sprays should address areas within the primary containment that are not covered by the spray drops. The mixing rate attributed to natural convection between sprayed and unsprayed regions of the containment building, provided that adequate flow exists between these regions, is assumed to be two turnovers of the unsprayed regions per hour, unless other rates are justified. The containment building atmosphere may be considered a single, well-mixed volume if the spray covers at least 90% of the volume and if adequate mixing of unsprayed compartments can be shown.

The SRP sets forth' a maximum decontamination factor (DF) for LGS AST LAR Page 50 of 76 February 27, 2004 Table B: -Conformance with'RG 1.183 Appendix A (Loss-of-Coolant-Accident) , -,,

RG,;, RG Position--. -. LGS

-- Comm'ents Section - -  :.->ys:s sI;YX2>.;-& An *i X elemental iodine based on the maximum iodine activity in the primary containment atmosphere when the sprays actuate, divided by the activity of iodine remaining at some time after decontamination. The SRP also states that the particulate iodine removal rate should be reduced by a factor of 10 when a DF of 50 is reached. The reduction in the removal rate is not required if the removal rate is based on the calculated time-dependent airborne aerosol mass. There Is no specified maximum DF for aerosol removal by sprays. The maximum activity to be used in determining the DF is defined as the iodine activity in the columns labeled "Total" in Tables 1 and 2 of this guide multiplied by 0.05 for elemental iodine and by 0.95 for particulate iodine (i.e., aerosol treated as particulate in SRP methodology).

3.4 Reduction in airborne radioactivity in the containment by in-containment Not No in-containment recirculation filter recirculation filter systems may be credited if these systems meet the Applicable systems exist at LGS.

guidance of Regulatory Guide 1.52 and Generic Letter 99-02 (Refs. A-5 and A-6). The filter media loading caused by the increased aerosol release associated with the revised source term should be addressed.

3.5 Reduction in airborne radioactivity in the containment by suppression Conforms No credit is taken for suppression pool scrubbing in BWRs should generally not be credited. However, the pool scrubbing in the LOCA AST staff may consider such reduction on an individual case basis. The reanalysis. Analyses have been evaluation should consider the relative timing of the blowdown and the performed that determined that the fission product release from the fuel, the force driving the release suppression pool liquid pH is through the pool, and the potential for any bypass of the suppression maintained greater than 7, and that, pool (Ref. 7). Analyses should consider iodine re-evolution if the therefore, iodine re-evolution is not suppression pool liquid pH is not maintained greater than 7. expected.

3.6 Reduction in airborne radioactivity in the containment by retention in ice Not LGS does not have ice condensers.

condensers, or other engineering safety features not addressed above, Applicable No other removal mechanisms are should be evaluated on an individual case basis. See Section 6.5.4 of credited other than natural the SRP (Ref. A-1). deposition.

3.7 The primary containment (i.e., drywell for Mark I and 11containment Conforms The LGS Mark II primary designs) should be assumed to leak at the peak pressure technical containment leakage is assumed to specification leak rate for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. For PWRs, the leak rate be 0.5% of containment mass per may be reduced after the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to 50% of the technical day for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and 0.25% per day LGS AST LAR Page 51 of 76 February 27, 2004 T-~kI AthW~I ~flfl~ff'h

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RG  ; i  : = LGS' Comments Section- --. _____--;_*________________________________________________.'".V __

'Ana __ __ __  : _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

specification leak rate. For BWRs, leakage may be reduced after the from 24 to 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> based on first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, if supported by plant configuration and analyses, to a containment pressure reductions.

value not less than 50% of the technical specification leak rate.

Leakage from subatmospheric containments is assumed to terminate when the containment is brought to and maintained at a subatmospheric condition as defined by technical specifications.

For BWRs with Mark IlIl containments, the leakage from the drywell into the primary containment should be based on the steaming rate of the heated reactor core, with no credit for core debris relocation. This leakage should be assumed during the two-hour period between the initial blowdown and termination of the fuel radioactivity release (gap and early in-vessel release phases). After two hours, the radioactivity is assumed to be uniformly distributed throughout the drywell and the primary containment.

3.8 If the primary containment is routinely purged during power operations, Conforms The LGS primary containment is not releases via the purge system prior to containment isolation should be routinely purged during power analyzed and the resulting doses summed with the postulated doses operation. Purging is limited to from other release paths. The purge release evaluation should assume inerting, de-inerting and occasional that 100% of the radionuclide inventory in the reactor coolant system short pressure control activities.

liquid is released to the containment at the initiation of the LOCA. This inventory should be based on the technical specification reactor coolant system equilibrium activity. Iodine spikes need not be considered. If the purge system is not isolated before the onset of the gap release phase, the release fractions associated with the gap release and early in-vessel phases should be considered as applicable.

4.1 Leakage from the primary containment should be considered to be Conforms Secondary Containment release is collected, processed by engineered safety feature (ESF) filters, if any, via the North Stack. Filtration credit and released to the environment via the secondary containment exhaust is taken after the 15.5-minute system during periods in which the secondary containment has a drawdown period. The Gap release negative pressure as defined in technical specifications. Credit for an begins at - 2 minutes after LOCA elevated release should be assumed only if the point of physical release initiation. For EAB and LPZ doses, is more than two and one-half times the height of any adjacent structure. ground level releases are assumed.

For Control Room doses, releases are based on zero-velocity vent LGS AST LAR Page 52 of 76 February 27, 2004 Table B: Conformancewith RG 1.183 Appendix A (Loss-of-Coolant Accident) ;___-i_;_>*-____i.;_<

RG RG Position'.- LGS Comments Section __:_:_.__;-_______;-.___-._;;_d;;_._____Anal Ysis: __

release assumptions, yielding ground level release equivalent dispersion

] factors.

4.2 Leakage from the primary containment is assumed to be released Conforms For EAB and LPZ doses, ground directly to the environment as a ground-level release during any period level releases are assumed. For in which the secondary containment does not have a negative pressure Control Room doses, releases are as defined in technical specifications. based on zero-velocity vent release assumptions.

4.3 The effect of high wind speeds on the ability of the secondary Conforms The wind speed exceeded only 5%

containment to maintain a negative pressure should be evaluated on an of the time at LGS in the secondary individual case basis. The wind speed to be assumed is the 1-hour containment vicinity is approximately average value that is exceeded only 5%of the total number of hours in 19.7 mph (175' elevation of the data set. Ambient temperatures used in these assessments should meteorological tower 2). It has been be the 1-hour average value that is exceeded only 5%or 95% of the determined that a wind speed of total numbers of hours in the data set, whichever is conservative for the greater than 35 mph would be intended use (e.g., if high temperatures are limiting, use those exceeded required before the secondary only 5%). containment pressures would be positive relative to outside air pressures at the downwind side of the reactor enclosure.

4.4 Credit for dilution in the secondary containment may be allowed when Conforms A 50% mixing credit is taken for adequate means to cause mixing can be demonstrated. Otherwise, the dilution/mixing in secondary leakage from the primary containment should be assumed to be containment. This mixing is transported directly to exhaust systems without mixing. Credit for attributed to the RERS flow network.

mixing, if found to be appropriate, should generally be limited to 50%.

This evaluation should consider the magnitude of the containment leakage in relation to contiguous building volume or exhaust rate, the location of exhaust plenums relative to projected release locations, the recirculation ventilation systems, and internal walls and floors that impede stream flow between the release and the exhaust.

LGS AST LAR Page 53 of 76 February 27, 2004 TableB: Conformancewith-RG1;183 Appendix A(Loss-of-CoolantAccident)  ; 7;. r * , -

RG ':]- RGPosition  ? ' ~- >  ;;' t 1 ~Comments

-Sectin':;

,, ; -* - t t at >

~~~~~~~~~~~~A66  ; Hi If;. -,wY- @ 'nI ....... i'--"

.y.s.......

s...... -w 4.b Primary containment leakage that bypasses the secondary containment Conforms No primary containment leakage, should be evaluated at the bypass leak rate incorporated in the technical with the exception of MSIV leakage, specifications. If the bypass leakage is through water, e.g., via a filled has been identified which bypasses piping run that is maintained full, credit for retention of iodine and the secondary containment. Only the aerosols may be considered on a case-by-case basis. Similarly, MSIV pathway leak rates are deposition of aerosol radioactivity in gas-filled lines may be considered incorporated into the Technical on a case-by-case basis. Specifications.

The AST analysis is based on an MSIV leakage limit of 200 scfh total leakage with not more that 100 scfh per line when tested at 2 22 psig.

Modeling of aerosol settling and elemental iodine deposition is based on methodology used by NRC in AEB-98-03 for both piping and the condenser. For the two steam lines modeled, two nodes are used. The first node is from the reactor pressure vessel to the inboard MSIV.

The second node is from the inboard MSIV to the Turbine Stop Valve that provides the seismically designed boundary of the MSIV alternate drain pathway. For aerosol settling, only horizontal piping runs are credited, and only the bottom surface area is considered available. A median settling velocity is used, given the conservatism in using a well-mixed treatment. For elemental iodine deposition both horizontal and vertical piping is credited, as well as I I all surfaces. This is because this LGS AST LAR Page 54 of 76 February 27, 2004

.TableB:. ,onformance with RG 1.183AppendixA'(Loss-of-Coolant'Accident) ' ::^' -  : :>

G o ssn;on.

oI irr1i ' LGI Comments' vi <7<

Section' _____--______._;__________-___________-_'_____ _A_'ls__;__-

JAnalysis deposition is not gravity dependent.

LGS has previously been analyzed and licensed to no longer credit a MSIV Leakage Control System, and to credit holdup, settling, and deposition in seismically rugged portions of the turbine condenser system. This historical evaluation is based on methodology described in NEDC-31858P, Rev. 2, "BWROG Report for Increasing MSIV Leakage Rate Limits and Elimination of Leakage Control Systems". That analysis was based on a design basis recirculation line break and TID-14844 based source terms. For this AST application, the analysis of MSIV leakage is updated to reflect AST parameters related to release timing, chemical makeup, and more recent approaches regarding fission product settling and deposition.

Credit is taken for deposition in the condenser, where the deposition area is the horizontal surface of the high pressure (HP) wetwell, and the HP condenser walls. By the time that activity has reached the condenser, the aerosols are essentially depleted. Therefore, vertical wall surfaces are credited for I I elemental iodine removal. No credit

Attachment 1 LGS AST LAR Page 55 of 76 February 27, 2004

'Table' B: ,Conformance with'RG .1.183 Appendix A (Loss-of-Coolant Accident) l'-'.'--*- .-

RG RG Position ... -- Comm' nts.'

Section 7 Anal 'sis '1 V is taken for any organic iodine removal in piping or the condenser.

Also, no credit is taken for condenser tubing even though this provides approximately 40 times the deposition area in the condenser

___________than what is credited.

4.6 Reduction in the amount of radioactive material released from the Conforms SGTS HEPA and charcoal adsorber secondary containment because of ESF filter systems may be taken into filters are credited in the evaluation account provided that these systems meet the guidance of Regulatory of a LOCA accident for onsite and Guide 1.52 (Ref. A-5) and Generic Letter 99-02 (Ref. A-6). offsite dose consequences. RERS HEPA filtration is also credited. Both of these systems meet the guidance of Regulatory Guide 1.52 and Generic Letter 99-02.

5.1 With the exception of noble gases, all the fission products released from Conforms With the exception of noble gases, all the fuel to the containment (as defined in Tables 1 and 2 of this guide) the fission products released from should be assumed to instantaneously and homogeneously mix in the the fuel to the containment are primary containment sump water (in PWRs) or suppression pool (in assumed to instantaneously and BWRs) at the time of release from the core. In lieu of this deterministic homogeneously mix in the approach, suitably conservative mechanistic models for the transport of suppression pool at the time of airborne activity in containment to the sump water may be used. Note release from the core.

that many of the parameters that make spray and deposition models conservative with regard to containment airborne leakage are non-conservative with regard to the buildup of sump activity.

5.2 The leakage should be taken as two times the sum of the simultaneous Conforms The 5-gpm leak rate is two times the leakage from all components in the ESF recirculation systems above sum of the allowed simultaneous which the technical specifications, or licensee commitments to item leakage from all ECCS components.

III.D.1.1 of NUREG-0737 (Ref. A-8), would require declaring such ECCS leakage is minimized at LGS systems inoperable. The leakage should be assumed to start at the through implementation of the earliest time the recirculation flow occurs in these systems and end at Program committed to in T.S.

the latest time the releases from these systems are terminated. 6.8.4.a, "Primary Coolant Sources Consideration should also be given to design leakage through valves Outside Containment".

LGS AST LAR Page 56 of 76 February 27, 2004 Table B: Conformrnce withRG1.183AcnendixA (Loss-of Coolant'Accident) 9 ";>-:. ' - .. :.<

RG - RG Position *'- -?  ; i' ; . j LGS - Comments-Section - - v; - ^ Analysis " -; - ' .

isolating ESF recirculation systems from tanks vented to atmosphere, Since certain ECCS systems take e.g., emergency core cooling system (ECCS) pump minflow return to the suction immediately from the refueling water storage tank. suppression pool, this leak path is assumed to start at time 0.

Leakage to atmospheric tanks is credible only for lines connecting from ECCS pump discharges to such a tank, because of relative elevations. The sole leakage paths to a tank vented to atmosphere meeting this condition are the High Pressure Coolant Injection / Reactor Core Isolation Cooling test lines that discharge to the Condensate Storage Tank (CST). These lines are isolated by two normally closed valves. Since the CST contents are demineralized water, ECCS leakage would quickly turn the water basic. Therefore, minimal elemental iodine is expected, and as a result, negligible iodine volatilization.

5.3 With the exception of iodine, all radioactive materials in the recirculating Conforms With the exception of iodine, all liquid should be assumed to be retained in the liquid phase. radioactive materials in ECCS liquids are assumed to be retained in the liquid phase.

5.4 If the temperature of the leakage exceeds 212 0F, the fraction of total Not The temperature of the leakage does iodine in the liquid that becomes airborne should be assumed equal to Applicable not exceed 212 0 F.

the fraction of the leakage that flashes to vapor. This flash fraction, FF, should be determined using a constant enthalpy, h, process, based on the maximum time-dependent temperature of the sump water circulating outside the containment:

LGS AST LAR Page 57 of 76 February 27, 2004 Table B: Conformance with RG 1.183 Aboendix A (Loss-of-Coolant Accident)  ; -- si;  ;';  ;'%a -  : ~ .' z - -

RGSe :RG Position -* <SA. .C LGs.

L omm ents FF= hf1l -h12 hfg Where: hf. is the enthalpy of liquid at system design temperature and pressure; hf2 is the enthalpy of liquid at saturation conditions (14.7 psia, 212'F); and hf, is the heat of vaporization at 212 0F.

5.5 If the temperature of the leakage is less than 212 0F or the calculated Conforms An airborne release fraction of 1.39%

flash fraction is less than 10%, the amount of iodine that becomes is used and was determined using a airborne should be assumed to be 10% of the total iodine activity in the methodology previously approved for leaked fluid, unless a smaller amount can be justified based on the use at the Clinton Power Station.

actual sump pH history and area ventilation rates. Suppression Pool water pH is maintained above 7 for the entire 30 days of the accident dose assessment period. Under these conditions virtually none of the iodine will be in elemental form, and organic iodine formation will be inhibited.

Because of the subcooled condition no flashing is expected.

5.6 The radioiodine that is postulated to be available for release to the Conforms The credited Control Room intake environment is assumed to be 97% elemental and 3% organic. charcoal and HEPA filters meet the Reduction in release activity by dilution or holdup within buildings, or by requirements of RG 1.52 and ESF ventilation filtration systems, may be credited where applicable. Generic Letter 99-02. These are Filter systems used in these applications should be evaluated against credited at 80% efficiency for the guidance of Regulatory Guide 1.52 (Ref. A-5) and Generic Letter 99- elemental and organic iodines.

02 (Ref. A-6). Aerosol removal efficiencies are assumed to be 99% based on the HEPA/charcoal combination.

6.1 For the purpose of this analysis, the activity available for release via Conforms The activity released through the MSIV leakage should be assumed to be that activity determined to be in MSIVs is the same concentration as the drywell for evaluating containment leakage (see Regulatory Position that used for evaluating Primary to 3). No credit should be assumed for activity reduction by the steam Secondary Containment leakage.

separators or by iodine partitioning in the reactor vessel. No credit is assumed for activity reduction by the steam separators or LGS AST LAR Page 58 of 76 February 27, 2004 Tablei 13-Conformance with R&1:183 Annendix A-(Ldss~-o-Cnolant Accident)'_ _ , -*_. I.-..* -'. .-1 -. ~-1 RG :. - RG Position .  ; . eS - . LGS " Comment Section . a *> 1/2 . .., t,,.'.. < -. 'i Analysis- -

by iodine partitioning in the reactor vessel.

6.2 All the MSIVs should be assumed to leak at the maximum leak rate Conforms MSIV leakage assumed in this above which the technical specifications would require declaring the accident analysis is 200 scfh for all MSIVs inoperable. The leakage should be assumed to continue for the steam lines and 100 scfh for any one duration of the accident. Postulated leakage may be reduced after the line when tested at 2 22 psig.

first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, if supported by site-specific analyses, to a value not less Reduction in leakage rates after 24 than 50% of the maximum leak rate. hours are based on post-accident containment pressures. No credit is taken for leakage rate reductions below 50% of the MSIV leakage limit.

6.3 Reduction of the amount of released radioactivity by deposition and Conforms See discussion under comments for plateout on steam system piping upstream of the outboard MSIVs may section 4.5 of Table B.

be credited, but the amount of reduction in concentration allowed will be evaluated on an individual case basis. Generally, the model should be based on the assumption of well-mixed volumes, but other models such as slug flow may be used if justified.

6.4 In the absence of collection and treatment of releases by ESFs such as Conforms Releases are assumed to be from the MSIV leakage control system, or as described in paragraph 6.5 the North Stack, without credit for below, the MSIV leakage should be assumed to be released to the holdup or dilution in the condenser or environment as an unprocessed, ground- level release. Holdup and Turbine Enclosure. The zero-velocity dilution in the turbine building should not be assumed. vent release assumption is effectively a ground level release assumption.

6.5 A reduction in MSIV releases that is due to holdup and deposition in Conforms See discussion under comments for main steam piping downstream of the MSIVs and in the main section 4.5 of Table B.

condenser, including the treatment of air ejector effluent by offgas systems, may be credited if the components and piping systems used in the release path are capable of performing their safety function during and following a safe shutdown earthquake (SSE). The amount of reduction allowed will be evaluated on an individual case basis.

References A-9 and A-10 provide guidance on acceptable models.

7.0 The radiological consequences from post-LOCA primary containment Conforms Containment purging as a purging as a combustible gas or pressure control measure should be combustible gas or pressure control LGS AST LAR Page 59 of 76 February 27, 2004 Table, B:.. Conformance with RG 1 . AppendixAp 8 (LooIan Accident)

(LndixA . . . M '  :;

RG:, RG Position ," -,,,,.-' LGS 'Comments '"ail-.

Section!-" ,; Anal sis '- - 7 analyzed. If the installed containment purging capabilities are measure is not required nor credited maintained for purposes of severe accident management and are not in any design basis analysis for 30 credited in any design basis analysis, radiological consequences need days following a design basis LOCA not be evaluated. If the primary containment purging is required within at LGS.

30 days of the LOCA, the results of this analysis should be combined with consequences postulated for other fission product release paths to determine the total calculated radiological consequences from the LOCA. Reduction in the amount of radioactive material released via ESF filter systems may be taken into account provided that these systems meet the guidance in Regulatory Guide 1.52 (Ref. A-5) and Generic Letter 99-02 (Ref. A-6).

Attachment 1 LGS AST LAR Page 60 of 76 February 27, 2004

-r-"I- t..:A-4-4013 0 1M.-I A ',I-. ' - - --- , - I-. 1".' ' " - "". -- - ;

RG~- 7,,,.;,t~ ;-.<* * ;v LGS , :.

Sectior: RG P ;Analysis§ Comments --

1 Acceptable assumptions regarding core inventory and the release of Conforms radionuclides from the fuel are provided in Regulatory Position 3 of this guide.

1.1 The number of fuel rods damaged during the accident should be based Conforms This is based on generic evaluation of on a conservative analysis that considers the most limiting case. This GEl1 and GE14 fuel, with heavy mast, analysis should consider parameters such as the weight of the dropped yielding 172 failed rods, based on heavy load or the weight of a dropped fuel assembly (plus any attached 87.33 rods per assembly and 764 handling grapples), the height of the drop, and the compression, torsion, assemblies in the core. Damage due and shear stresses on the irradiated fuel rods. Damage to adjacent fuel to a fuel assembly drop into the reactor assemblies, if applicable (e.g., events over the reactor vessel), should vessel bounds a drop in the spent fuel be considered. pool. This is due the greater distance of the drop in the vessel as opposed to the drop into the fuel pool.

1.2 The fission product release from the breached fuel is based on Conforms Gap activity assumed is per this Regulatory Position 3.2 of this guide and the estimate of the number of guidance.

fuel rods breached. All the gap activity in the damaged rods is assumed to be instantaneously released. Radionuclides that should be considered include xenons, kryptons, halogens, cesiums, and rubidiums.

1.3 The chemical form of radioiodine released from the fuel to the spent fuel Conforms All iodine added to pool assumed to pool should be assumed to be 95% cesium iodide (Csl), 4.85 percent dissociate and re-evolves as elemental elemental iodine, and 0.15 percent organic iodide. The Csi released iodine and treated appropriately.

from the fuel is assumed to completely dissociate in the pool water.

Because of the low pH of the pool water, the iodine re-evolves as elemental iodine. This is assumed to occur instantaneously. The NRC staff will consider, on a case-by-case basis, justifiable mechanistic treatment of the iodine release from the pool.

2 If the depth of water above the damaged fuel is 23 feet or greater, the Conforms The analyzed water depth above decontamination factors for the elemental and organic species are 500 damaged fuel is 23 feet. Although the and 1, respectively, giving an overall effective decontamination factor of actual water coverage over damaged 200 (i.e., 99.5% of the total iodine released from the damaged rods is fuel in the reactor vessel is 52 feet, no retained by the water). This difference in decontamination factors for further credit is applied for the elemental (99.85%) and organic iodine (0.15%) species results in the additional (i.e., >23 feet) water depth in

-iodine above the water being composed of 57% elemental and 43% accordance with regulatory guidance.

Attachment 1 LGS AST LAR Page 61 of 76 February 27, 2004 Table C: Conformance with Reulatorv Guide 1.183 ADDendix B (Fuel Handling Accident) G- en - "

Y--',e:' ' *

.RG ~...>- LGS Section- 'RG Positidn '; .. K<',. .. . A".^I ' C e organic species. If the depth of water is not 23 feet, the decontamination An overall DF of 200 is used.

factor will have to be determined on a case-by-case method (Ref. B-1).

For a drop over the spent fuel pool, coverage over the dropped fuel assembly, as it lies across the top of the bail handles of assemblies in the fuel rack, is 21.6 feet. The coverage over the fuel pins for assemblies in the racks is 22.6 feet. The calculated df, weighted by damaged fuel pin count, is 171. The conservatively determined damage over the spent fuel pool is 70% of that over the reactor vessel.

Therefore, the net effect (damage times df), is that a drop over the reactor is bounding.

3 The retention of noble gases in the water in the fuel pool or reactor Conforms DF = 1 for noble gas isotopes.

cavity is negligible (i.e., decontamination factor of 1). Particulate radionuclides are assumed to be retained by the water in the fuel pool or reactor cavity (i.e., infinite decontamination factor).

4.1 The radioactive material that escapes from the fuel pool to the fuel Conforms The release is assumed to occur over building is assumed to be released to the environment over a 2-hour a two hour period.

time period.

4.2 A reduction in the amount of radioactive material released from the fuel Conforms No credit is taken for the Standby Gas pool by engineered safety feature (ESF) filter systems may be taken into Treatment System.

account provided these systems meet the guidance of Regulatory Guide 1.52 and Generic Letter 99-02 (Refs. B-2, B-3). Delays in radiation detection, actuation of the ESF filtration system, or diversion of ventilation flow to the ESF filtration system(21) should be determined and accounted for in the radioactivity release analyses.

LGS AST LAR Page 62 of 76 February 27, 2004 lTable C: Conformancewith.ReaulatorVGuide1.183 ApDendixB.(FuelHan-dlinqAAccident)W-;.it.: I i

Sectio nRG itiIn . .  ;- .. A--na ysis Comments 4.3 The radioactivity release from the fuel pool should be assumed to be Conforms Two-hour release to the environment is drawn into the ESF filtration system without mixing or dilution in the fuel assumed, without SGTS or CREFAS building. If mixing can be demonstrated, credit for mixing and dilution filtration.

may be considered on a case-by-case basis. This evaluation should consider the magnitude of the building volume and exhaust rate, the potential for bypass to the environment, the location of exhaust plenums relative to the surface of the pool, recirculation ventilation systems, and internal walls and floors that impede stream flow between the surface of the pool and the exhaust plenums.

5.1 If the containment is isolated during fuel handling operations, no Conforms Secondary Containment isolation is not radiological consequences need to be analyzed. credited.

5.2 If the containment is open during fuel handling operations, but designed Conforms Automatic Secondary Containment to automatically isolate in the event of a fuel handling accident, the isolation is not credited.

release duration should be based on delays in radiation detection and completion of containment isolation. If it can be shown that containment isolation occurs before radioactivity is released to the environment, no radiological consequences need to be analyzed.

5.3 If the containment is open during fuel handling operations (e.g., Site- Refueling Area Secondary personnel air lock or equipment hatch is open), the radioactive material specific Containment closure will be that escapes from the reactor cavity pool to the containment is released exception accomplished within a 1-hour time to the environment over a 2-hour time period. period as opposed to the suggested 30 minutes. Administrative controls will Note 3: be in place associated with closure of The staff will generally require that technical specifications allowing such doors and penetrations.

operations include administrative controls to close the airlock, hatch, or penetrations within 30 minutes. Such administrative controls will generally require that a dedicated individual be present, with the necessary equipment available, to restore containment closure should a fuel handling accident occur. Radiological analyses should generally not credit this manual isolation.

LGS AST LAR Page 63 of 76 February 27, 2004 Table C. Conformance with RedulatorV Guide .1.1 83-AQpendixkB (Fuel Hannrdlin'Accident) *--

'- ." _l ,`,

Section. RG Posit Analysis Comments , -.

5.4 A reduction in the amount of radioactive material released from the Not No credit is being taken for filtration of containment by ESF filter systems may be taken into account provided Applicable release from the Reactor Enclosure.

that these systems meet the guidance of Regulatory Guide 1.52 and Generic Letter 99-02 (Refs. B-2 and B-3). Delays in radiation detection, actuation of the ESF filtration system, or diversion of ventilation flow to the ESF filtration system should be determined and accounted for in the radioactivity release analyses.

5.5 Credit for dilution or mixing of the activity released from the reactor Not Two-hour release to the environment is cavity by natural or forced convection inside the containment may be Applicable assumed.

considered on a case-by-case basis. Such credit is generally limited to 50% of the containment free volume. This evaluation should consider the magnitude of the containment volume and exhaust rate, the potential for bypass to the environment, the location of exhaust plenums relative to the surface of the reactor cavity, recirculation ventilation systems, and internal walls and floors that impede stream flow between the surface of the reactor cavity and the exhaust plenums.

Attachment 1 LGS AST LAR Page 64 of 76 February 27, 2004 e

'Table D:Conformancewith Regulator Guid 1;183:Appe'dix.:C (Control Rod Drop Accient. " .- '.

RG  ; 'RG Position L S.,,.,:, Comments ' -i:.' !.

Section . -.  : ..-  : - g . l ^ -_ _ _ ;_:_;-:

_ _ _ _ __I_ _-_Analysis-1 Assumptions acceptable to the NRC staff regarding core inventory are Conforms Breached/melted fuel rods and provided in Regulatory Position 3 of this guide. For the rod drop release fractions have been updated accident, the release from the breached fuel is based on the estimate of to reflect GE14 fuel, and release the number of fuel rods breached and the assumption that 10% of the fractions per RG 1.183. Releases are core inventory of the noble gases and iodines is in the fuel gap. The based on 1,200 fuel rods breached release attributed to fuel melting is based on the fraction of the fuel that and melting in 0.77% of the fuel reaches or exceeds the initiation temperature for fuel melting and on the contained in the breached rods.

assumption that 100% of the noble gases and 50% of the iodines A conservative radial peaking factor of contained in that fraction are released to the reactor coolant. 1.7 is used in agreement with the AST Calculation for the Fuel Handling Accident.

In addition to noble gas and iodine releases, releases of 12% of the core inventory of Cesium (an alkali metal, per Table 5 in Regulatory Position 3 of the guide) is assumed, based on Table 3 in Regulatory Position 3 of the guide.

Radionuclide grouping is per Table 5 in Regulatory Position 3 of the guide, as implemented in RADTRAD.

2 If no or minimal fuel damage is postulated for the limiting event, the Conforms Substantial fuel damage is postulated.

released activity should be the maximum coolant activity (typically 4 pCi/gm DE 1-1 31) allowed by the technical specifications.

3.1 The activity released from the fuel from either the gap or from fuel Conforms Instantaneous mixing is assumed per pellets is assumed to be instantaneously mixed in the reactor coolant this guidance.

within the pressure vessel.

3.2 Credit should not be assumed for partitioning in the pressure vessel or Conforms No partitioning is assumed.

for removal by the steam separators.

3.3 Of the activity released from the reactor coolant within the pressure Conforms Released activity is per this guidance.

vessel, 100% of the noble gases, 10% of the iodine, and 1% of the remaining radionuclides are assumed to reach the turbine and condensers.

LGS AST LAR Page 65 of 76 February 27, 2004

,i,+H Dt%^l J +pwmf fgt-2irG%' I 10'l

-AA.I-.'~ -Avm'r rr4trf ftn,~v#+,491 DeA fr,'rr~--A e X~ - I . -- ;4_ II ScG RG^:sitidny  ;* i , X.e LGS Commenits-,

3.4 Of the activity that reaches the turbine and condenser, 100% of the Conforms The condenser leak rate of 1%per noble gases, 10% of the iodine, and 1%of the particulate radionuclides day for a period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is are available for release to the environment. The turbine and assumed. All releases are assumed condensers leak to the atmosphere as a ground- level release at a rate to be at ground level and based on of 1%per day for a period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, at which time the leakage is zero-velocity vent release assumed to terminate. No credit should be assumed for dilution or assumptions. Radioactive decay holdup within the turbine building. Radioactive decay during holdup in during holdup in the condenser is the turbine and condenser may be assumed, assumed.

3.5 In lieu of the transport assumptions provided in paragraphs 3.2 through Not Sections 3.2 through 3.4 are used in 3.4 above, a more mechanistic analysis may be used on a case-by-case Applicable the analysis.

basis. Such analyses account for the quantity of contaminated steam carried from the pressure vessel to the turbine and condensers based on a review of the minimum transport time from the pressure vessel to the first main steam isolation (MSIV) and considers MSIV closure time.

3.6 The iodine species released from the reactor coolant within the pressure Conforms No credit for RERS, SGTS, or vessel should be assumed to be 95% CsI as an aerosol, 4.85% CREFAS filters is taken, and therefore elemental, and 0.15% organic. The release from the turbine and variation in iodine species has no condenser should be assumed to be 97% elemental and 3%organic. effect.

Foot- The activity assumed in the analysis should be based on the activity Conforms Projected fuel damage is the limiting note 1 associated with the projected fuel damage or the maximum technical case.

specification values, whichever maximizes the radiological consequences. In determining the dose equivalent 1-131 (DE 1-131), only the radioiodine associated with normal operations or iodine spikes should be included. Activity from projected fuel damage should not be included.

LGS AST LAR Page 66 of 76 February 27, 2004 Table D:., Conformance with Re ulatory'Guide 11 83 Appendix C (Control Rod Drop Accident) . .. - -' -

RG : RG Position - :t - - LGS- VA. Comments;' -;-

-  ; C - - ~~~: *An~alYSss- -* '*.

Foot- If there are forced flow paths from the turbine or condenser, such as Conforms Upon detection of high radiation levels note 2 unisolated motor vacuum pumps or unprocessed air ejectors, the by the Main Steam Line Radiation leakage rate should be assumed to be the flow rate associated with the Monitor system the mechanical most limiting of these paths. Credit for collection and processing of vacuum pump trips. Air ejector flows releases, such as by off gas or standby gas treatment, will be are processed through the offgas considered on a case-by-case basis. system, which removes all iodines and substantially delays noble gases.

LGS is a clean sealing steam system so gland seals are not considered a forced release pathway.

LGS AST LAR Page 67 of 76 February 27, 2004 Table E:- Conformance with ReOulatonr uiiiride .1 AnnendeiiD (Main Steam-L MrP ak '- . - -,  ;' .,' ' 8-X.... '.II RG '- '.1RG Position 'LGS:V.>,- I.... {.-.." .-. Comments-Section . Ae.

' :i: .'

1 Assumptions acceptable_-_to -"-the -- NRC staff reaardino core inventorv and

- -- _.-0 -- --- - . *.1 -_

Not N fuel mane releasetimate the release of radionuclides from the fuel are provided in Regulatory Applicable based on coolant activity.

Position 3 of this guide. The release from the breached fuel is based on Regulatory Position 3.2 of this guide and the estimate of the number of fuel rods breached.

2 If no or minimal fuel damage is postulated for the limiting event, the Conforms No fuel damage is expected even released activity should be the maximum coolant activity allowed by with the extension of the MSIV technical specification. The iodine concentration in the primary coolant is closure time to 10.5 seconds. The assumed to correspond to the following two cases in the nuclear steam 10.5 seconds is the same as that supply system vendor's standard technical specifications. currently licensed for Peach Bottom, where no fuel damage is postulated.

Per the current Limerick MSLB analysis, the core is uncovered for 60 seconds, which is well after the MSIV closure time. No fuel damage results during this event.

By changing the MSIV closure time by 5 seconds for AST, there is no concern of uncovering the core. This is due to the swell that results from reactor depressurization that will maintain adequate core coverage until the MSIV isolation.

LGS Technical Specifications limits the reactor coolant Dose Equivalent (DE) 1-131 specific activity to 0.2 pCi/gm, with action to isolate all main steam lines if the reactor coolant DE 1-131 specific activity exceeds 4.0 I_ .

pCi/gm during Power Operation or

Attachment 1 LGS AST LAR Page 68 of 76 February 27, 2004 Table-E: Conformance With Regulatory Guide1;183 Appendix-D (Main Steam Line Br e ', :aik)' .K-^

AI RG

. RG Position -. - - LGS-- Comments.'

S'cti~on;; - ;.. *r, ,

.,;,,,, , A

.,, a,\ 'Analysis - ' ^.*

Startup.

2.1 The concentration that is the maximum value (typically 4.0 pCi/gm DE I- Conforms See Item 2 above.

131) permitted and corresponds to the conditions of an assumed pre-accident spike, and 2.2 The concentration that is the maximum equilibrium value (typically 0.2 Conforms See Item 2 above.

pCi/gm DE 1-131) permitted for continued full power operation.

3 The activity released from the fuel should be assumed to mix Conforms Mixing is per this guidance.

instantaneously and homogeneously in the reactor coolant. Noble gases should be assumed to enter the steam phase instantaneously.

4.1 The main steam line isolation valves (MSIV) should be assumed to close Conforms See Item 2 above.

in the maximum time allowed by technical specifications.

4.2 The total mass of coolant released should be assumed to be that Conforms Mass of coolant released is per this amount in the steam line and connecting lines at the time of the break guidance.

plus the amount that passes through the valves prior to closure.

4.3 All the radioactivity in the released coolant should be assumed to be Conforms This guidance was used in the released to the atmosphere instantaneously as a ground-level release. analysis.

No credit should be assumed for plateout, holdup, or dilution within facility buildings.

4.4 The iodine species released from the main steam line should be Conforms No filtration is credited, so the iodine assumed to be 95% Csl as an aerosol, 4.85% elemental, and 0.15% species are irrelevant.

. _ organic.

LGS AST LAR Page 69 of 76 February 27, 2004

5.0 REGULATORY ANALYSIS

5.1 No Significant Hazards Consideration Exelon Generation Company, LLC (Exelon) is requesting a revision to the Facility Operating Licenses for Limerick Generating Station, Units 1 and 2. Specifically, we are requesting a revision to the Technical Specifications and licensing and design bases to reflect the application of alternative source term (AST) assumptions.

The AST analyses were performed in accordance with the guidance in Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," dated July 2000, and Standard Review Plan Section 15.0.1, "Radiological Consequence Analyses Using Alternative Source Terms."

According to 10 CFR 50.92, "Issuance of amendment," paragraph (c), a proposed amendment to an operating license involves no significant hazards consideration if operation of the facility in accordance with the proposed amendment would not:

(1) Involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) Create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) Involve a significant reduction in a margin of safety.

In support of this determination, an evaluation of each of the three criteria set forth in 10 CFR 50.92 is provided below regarding the proposed license amendment.

5.1.1 The proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

The implementation of AST assumptions has been evaluated in revisions to the analyses of the following limiting design basis accidents (DBAs) at Limerick Generating Station (LGS):

• Loss-of-Coolant Accident,

• Main Steam Line Break Accident,

• Fuel Handling Accident, and

• Control Rod Drop Accident.

Based upon the results of these analyses, it has been demonstrated that, with the requested changes, the dose consequences of these limiting events are within the regulatory guidance provided by the NRC for use with the AST. This guidance is presented in 10 CFR 50.67, Regulatory Guide 1.183, and Standard Review Plan Section 15.0.1. The AST is an input to calculations used to evaluate the consequences of an accident, and does not by itself affect the plant response, or the actual pathway of the radiation released from the fuel. It does however; better LGS AST LAR Page 70 of 76 February 27, 2004 represent the physical characteristics of the release, so that appropriate mitigation techniques may be applied.

For the proposed change to increase the MSIV closure time by 5 seconds (5.5 to 10.5 seconds), although nearly twice as much steam mass is released from the current assumptions, there is no significant net increase of dose to Control Room or offsite personnel. The effect of the increased release time is offset by the dose methodologies applied using AST. The increased closure time also impacts the volume of reactor coolant available to maintain core coverage; however, the swell that results from the reactor depressurization adequately covers the core throughout the time it takes to complete the MSIV isolation.

The AST methodology follows the guidance provided in Regulatory Guide 1.183 and conforms to the dose limits in 10 CFR 50.67. Even though these limits are not directly comparable to the previously specified whole body and thyroid requirements of GDC 19 and 10 CFR 100.1 1, the results of the AST analyses have demonstrated that the 10 CFR 50.67 limits are satisfied. Therefore, it is concluded that AST does not involve a significant increase in the consequences of an accident previously evaluated.

Implementation of AST provides increased operating margins for: RERS, SGTS, and CREFAS filtration efficiencies; MSIV closure time; and RERS flow. It also relaxes secondary containment integrity requirements while handling irradiated fuel that has decayed for greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and during core alterations. Automatic initiation of the radiation isolation mode for the control room is no longer credited in the accident analysis, which can relax some Technical Specification surveillance requirements.

The equipment affected by the proposed changes is mitigative in nature, and relied upon after an accident has been initiated. Application of the AST does result in changes to Updated Final Safety Analysis Report (UFSAR) functions (e.g., MSIV closure time, SLC system) and operation of various filtration systems. As a condition of application of AST, LGS is proposing to use the SLC system to control the Suppression Pool pH following a LOCA only. These changes have been included within the evaluations for these proposed changes. While the operation of various systems does change with the implementation of AST, the affected systems are not accident initiators. Application of the AST is not an initiator of a design basis accident. The proposed changes to the Technical Specifications (TS), while they revise certain performance requirements, do not require any physical changes to the plant.

Additionally, the proposed change to the SLC system surveillance requirement (SR), to verify the weight of Boron-1 0, is equivalent to the current requirement to determine the minimum available weight of sodium pentaborate. This change performs the same chemical analysis determination as the current TS requirement, but more clearly identifies the boron weight available to mitigate the ATWS event, and, given the possible range of tank enrichments, also supports determining the minimum required weight of boron needed to control suppression pool pH for AST.

LGS AST LAR Page 71 of 76 February 27, 2004 As a result, the proposed changes do not affect any of the parameters or conditions that could contribute to the initiation of any accidents. Relaxation of operability requirements during the specified conditions will not significantly increase the probability of occurrence of an accident previously analyzed. Since design basis accident initiators are not being altered by adoption of the AST, the probability of an accident previously evaluated is not affected.

Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

5.1.2 The proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

The proposed changes do not involve a physical change to the plant.

Implementation of AST provides increased operating margins for: RERS, SGTS, AND CREFAS filtration efficiencies; MSIV closure time; and RERS flow. It also relaxes secondary containment integrity requirements while handling irradiated fuel that has decayed for greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and during core alterations. Automatic initiation of the radiation isolation mode for the control room is no longer credited in the accident analysis, which can relax some Technical Specification surveillance requirements.

Similarly, the proposed changes do not require any physical changes to any structures, systems or components involved in the mitigation of any accidents. The sodium pentaborate requirement for the SLC system does not change. Therefore, no new initiators or precursors of a new or different kind of accident are created. New equipment or personnel failure modes that might initiate a new type of accident are not created as a result of the proposed changes. I As such, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

5.1.3 The proposed change does not involve a significant reduction in a margin of safety.

Safety margins and analytical conservatisms have been evaluated and have been found acceptable. The analyzed events have been carefully selected and margin has been retained to ensure that the analyses adequately bound postulated event scenarios. The dose consequences due to design basis accidents comply with the requirements of 10 CFR 50.67 and the guidance of Regulatory Guide 1.183.

The proposed changes are associated with the implementation of a new licensing basis for LGS Design Basis Accidents (DBAs). Approval of the change from the original source term to a new source term taken from Regulatory Guide 1.183 is being requested. The results of the accident analyses, revised in support of the proposed changes, are subject to revised acceptance criteria. The analyses have been performed using conservative methodologies, as specified in Regulatory Guide 1.183. The dose consequences of these DBAs remain within the acceptance criteria presented in 10 CFR 50.67, "Accident Source Term", and Regulatory Guide 1.183.

Attachment I LGS AST LAR Page 72 of 76 February 27, 2004 The proposed changes continue to ensure that the doses at the exclusion area boundary (EAB) and low population zone boundary (LPZ), as well as the Control Room, are within corresponding regulatory limits.

Therefore, the proposed changes do not involve a significant reduction in a margin of safety.

Conclusion Exelon Generation Company, LLC (Exelon) concludes that the proposed changes present no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of no significant hazards consideration is justified.

5.2 Applicable Regulatory Requirements/Criteria The NRC's traditional methods (prior to the AST) for calculating the radiological consequences of design basis accidents are described in a series of regulatory guides and Standard Review Plan (SRP) chapters. That guidance was developed to be consistent with the TID-14844 source term and the whole body and thyroid dose guidelines stated in 10 CFR 100.11. Many of those analysis assumptions and methods are inconsistent with the ASTs and with the Total Effective Dose Equivalent (TEDE) criteria provided in 10 CFR 50.67. Regulatory Guide 1.183 provides assumptions and methods that are acceptable to the NRC staff for performing design basis radiological analyses using an AST. This guidance supersedes corresponding radiological analysis assumptions provided in the older regulatory guides and SRP chapters when used in conjunction with an approved AST and the TEDE criteria provided in 10 CFR 50.67.

Due to the comprehensive nature of Regulatory Guide 1.183, the Tables in Section 4 above were incorporated into this submittal to show how each section of the new guidance is being addressed.

Also, the NRC published a new SRP section to address AST. It is Standard Review Plan Section 15.0.1, Rev. 0, entitled "Radiological Consequence Analyses Using Alternative Source Terms". It provides guidance on which NRC branches will review various aspects of an AST license amendment request, but otherwise is consistent with the guidance found in Regulatory Guide 1.183. The plant-specific information provided in this license amendment request is consistent with the guidance found in SRP 15.0.1.

In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or the health and safety of the public.

6.0 ENVIRONMENTAL CONSIDERATION

Exelon Generation Company, LLC (Exelon) has evaluated the proposed changes against the criteria for identification of licensing and regulatory actions requiring LGS AST LAR Page 73 of 76 February 27, 2004 environmental assessment in accordance with 10 CFR 51.21, "Criteria for and identification of licensing and regulatory actions requiring environmental assessments."

Exelon has determined that the proposed changes meet the criteria for a categorical exclusion as set forth in 10 CFR 51.22, "Criterion for categorical exclusion; identification of licensing and regulatory actions eligible for categorical exclusion or otherwise not requiring environmental review," paragraph (c)(9), and as such, has determined that no irreversible consequences exist in accordance with 10 CFR 50.92, "Issuance of amendment," paragraph (b). This determination is based on the fact that this change is being proposed as an amendment to a license issued pursuant to 10 CFR 50, "Domestic Licensing of Production and Utilization Facilities," which changes a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20," Standards for Protection Against Radiation,"

or that changes an inspection or a surveillance requirement, and the amendment meets the following specific criteria.

(i) The amendment involves no significant hazards consideration.

As demonstrated in Section 5.1 above, the proposed changes do not involve a significant hazards consideration.

(ii) There is no significant change in the types or significant increase in the amounts of any effluents that may be released offsite.

The following table demonstrates that Exelon meets the radiological criteria described in 10 CFR 50.67 for the exclusion area boundary (EAB) and the low population zone (LPZ). The EAB and LPZ doses represent a small fraction of the dose limits.

'-iI-lt D' r1

D-(re -TEDE)'

,, .. Ac-l.ntEAB Doses L Doses and Limits

,Dose e Limit _?_, DDs I.,,, .L Loss of Coolant 0.90 25 1.25 25 Accident Main Steam 2.82(1) 25(t) 1.11(1) 25(1)

Line Break 0.14(2) 2.5(2) 0.056(2) 2.5(2)

Control Rod 0.049 6.3 0.034 6.3 Drop Accident Fuel Handling 0.88 6.3 0.32 6.3 Accident I I I Notes: (1) Based on a pre-accident spike concentration of 4.0 iCVgm dose equivalent 1-131.

Attachment I LGS AST LAR Page 74 of 76 February 27, 2004 (2) Based on a maximum equilibrium concentration of 0.2 gCVgm dose equivalent 1-131. i Adoption of the alternative source term and Technical Specification changes which implement certain conservative assumptions in the alternative source term analyses will not result in physical changes to the plant that could significantly alter the type or amounts of effluents that may be released offsite.

Changes to operational parameters that could affect effluent releases have been demonstrated through analysis to satisfy regulatory requirements.

(iii) There is no significant increase in individual or cumulative occupational radiation exposure.

The following table demonstrates that Exelon meets the radiological criteria described in 10 CFR 50.67 for the Control Room. Control Room exposure to operators is less than the five rem total effective dose equivalent (TEDE) over 30 days for all accidents.

Control Room DoseResults (remTE1E

- <Accident l Dos Lmit Loss of Coolant Accident 4.02 5.0 Main Steam Line Break 3.61 ') 5.0 0.18(2)

Control Rod Drop Accident 1.62 5.0 Fuel Handling Accident 2.52 5.0 Notes: (1) Based on a pre-accident spike concentration of 4.0 giCVgm dose equivalent 1-131.

(2) Based on a maximum equilibrium concentration of 0.2 piCVgm dose equivalent 1-131.

The implementation of the alternative source term has been evaluated in revisions to the analyses of the limiting design basis accidents at Limerick Generating Station, Units 1 and 2. These accidents include the control rod drop accident, fuel handling accident, loss of coolant accident, and main steam line break accident. Based upon the results of these analyses, it has been demonstrated that with the proposed changes, the dose consequences of these limiting events are within the regulatory guidance provided by the NRC for use with alternative source term (i.e., 10 CFR 50.67 and 10 CFR 50, Appendix A, General Design Criterion 19). Thus, there will be no significant increase in either individual or cumulative occupational radiation exposure.

LGS AST LAR Page 75 of 76 February 27, 2004

7.0 REFERENCES

7.1 U. S. Atomic Energy Commission, Technical Information Document (TID) 14844, "Calculation of Distance Factors for Power and Test Reactor Sites," March 23, 1962 7.2 U. S. Nuclear Regulatory Commission Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors,' July 2000 7.3 U. S. Nuclear Regulatory Commission Standard Review Plan 15.0.1, "Radiological Consequence Analyses Using Alternative Source Terms," Revision 0, July 2000 7.4 NUREG-1465, "Accident Source Terms for Light-Water Nuclear Power Plants,"

February 1995 7.5 A. G. Croff, "A User's Manual for the ORIGEN 2 Computer Code," ORNL/TM-7175, Oak Ridge National Laboratory, July 1980 7.6 S. L. Humphreys et al., "RADTRAD: A Simplified Model for Radionuclide Transport and Removal and Dose Estimation," NUREG/CR-6604, U. S. Nuclear Regulatory Commission, April 1998 7.7 U. S. Nuclear Regulatory Commission Regulatory Guide 1.145, "Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants," Revision 1, November 1982 7.8 T. J. Bander, "PAVAN: An Atmospheric Dispersion Program for Evaluating Design Basis Accidental Releases of Radioactive Materials from Nuclear Power Stations,"

NUREG-2858, U. S. Nuclear Regulatory Commission, November 1982 7.9 J. V. Ramsdell and C. A. Simonen, "Atmospheric Relative Concentrations in Building Wakes," NUREG-6331, Revision 1, U. S. Nuclear Regulatory Commission, May 1997. (ARCON96) 7.10 ANSI/ANS-2.5-1984, "Standard for Determining Meteorological Information at Nuclear Power Sites", 1984 7.11 U. S. Nuclear Regulatory Commission Regulatory Guide 1.23 (Safety Guide 23),

"Onsite Meteorological Programs," February 17, 1972 7.12 U. S. Nuclear Regulatory Commission Regulatory Guide 1.3, "Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss-of-Coolant Accident for Boiling Water Reactors," Revision 2, June 1974 7.13 U. S. Nuclear Regulatory Commission Standard Review Plan 6.4, "Control Room Habitability Systems," Revision 2, July 1981 LGS AST LAR Page 76 of 76 February 27, 2004 7.14 U. S. Nuclear Regulatory Commission Regulatory Guide 1.5, "Assumptions Used for Evaluating the Potential Radiological Consequences of a Steam Line Break Accident for Boiling Water Reactors," March 1971 7.15 Federal Guidance Report No. 11, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion", 1988 7.16 Federal Guidance Report No. 12, "External Exposure to Radionuclides in Air, Water, and Soil", 1993.

7.17 Generic Letter 99-02, Laboratory Testing of Nuclear-Grade Activated Charcoal, June 3,1999 7.18 GE Report NEDE-31152P, "General Electric Fuel Bundle Designs", Revision 7, June, 2000.

7.19 Regulatory Guide 1.194; Atmospheric Relative Concentrations for Control Room Radiological Habitability Assessments at Nuclear Power Plants; U.S. Nuclear Regulatory Commission; December 2001.

7.20 Technical Specification Task Force (TSTF) Traveler, TSTF-51, "Revise Containment Requirements During Handling Irradiated Fuel and Core Alterations," Revision 2 7.21 NUREG-0737, "Clarification of TMI Action Plan Requirements", October 1980 7.22 10 CFR 50.67, "Accident source term," December 23, 1999 7.23 10 CFR 50.90, "Application for amendment of license or construction permit",

October 4, 1999 7.24 10 CFR 50, Appendix A, General Design Criterion 19 7.25 ICRP Publication 30, "Limits for Intakes of Radionuclides by Workers, 7.26 AEB-98-03, "Assessment of Radiological Consequences for the Perry Pilot Plant Application using the Revised (NUREG-1 465) Source Term", December 9, 1998 7.27 NUREG/CR-6189, D.A. Powers et al, "A simplified Model of Aerosol Removal by Natural Processes in Reactor Containments," July 1996 7.28 NEDO-31400A, "Safety Evaluation for Eliminating the BWR MSIV Closure Function and Scram Function of the MSLRM", October 1992 7.29 NEDC-31858P, Rev. 2, "BWROG Report for Increasing MSIV Leakage Rate Limits and Elimination of Leakage Control Systems", September 1993

ATTACHMENT 2 LIMERICK GENERATING STATION UNITS 1 AND 2 Docket Nos. 50-352 & 50-353 License Nos. NPF-39 & NPF-85 License Amendment Request "LGS Alternative Source Term Implementation" Markup of Technical Specification Pages UNIT 1 UNIT 2 1-2 1-2 1-6 1-6 1-7 1-7 3/4 1-19 3/4 1-19 3/4 1-20 3/4 1-20 3/4 3-16 3/4 3-16 3/4 3-31 3/4 3-31 3/4 3-64 3/4 3-64 3/4 3-65 3/4 3-65 3/4 3-66 3/4 3-66 3/4 3-67 3/4 3-67 3/4 4-23 3/4 4-23 3/4 6-3 3/4 6-3 3/4 6-47 3/4 6-47 3/4 6-50 3/4 6-50 3/4 6-52 3/4 6-52 3/4 6-53 3/4 6-53 3/4 6-55 3/4 6-55 3/4 6-56 3/4 6-56 3/4 7-3 3/4 7-3 3/4 7-4 3/4 7-4 3/4 7-5 3/4 7-5 3/4 7-6 3/4 7-6 3/4 7-7 3/4 7-7 3/4 8-9 3/4 8-9 3/4 8-14 3/4 8-14 3/4 8-14A 3/4 8-14A 3/4 8-20 3/4 8-19 3/4 8-20

It.

DEFINITIONS CORE ALTERATION 1.7 CORE ALTERATION shall be the movement of any fuel, sources, or reactivity control components within the reactor vessel with the vessel head removed and fuel in the vessel. The following exceptions are not considered to be CORE ALTERATIONS:

a) Movement of source range monitors, local power range monitors, intermediate range monitors, traversing incore probes, or special moveable detectors (including undervessel replacement); and b) Control rod movement, provided there are no fuel assemblies in the associated core cell.

Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe position.

CORE OPERATING LIMITS REPORT 1.7a The CORE OPERATING LIMITS REPORT (COLR) is the unit-specific document that provides the core operating limits for the current operating reload cycle. These cycle-specific core operating limits shall be determined for each reload cycle in accordance with Specifications 6.9.1.9 thru 6.9.1.12. Plant operation within these limits is addressed in individual specifications.

CRITICAL POWER RATIO 1.8 The CRITICAL POWER RATIO (CPR) shall be the ratio of that power in the assembly which is calculated by application of the (GEXL) correlation to cause some point in the assembly to experience boiling transition, divided by the act al assembl operating power.

DOSE EQUIVALENT 1-131 tnqcCin onmee4ecau dose ecto;.4qlenf(&.EE) 1.9 DOSE EQUIVALENT I-13 shall be that concentration of I-131, microcuries per gram, which alon ould produce the same s the quantity and isotopic mixture1of I-131, 1-132, 1-133, I-D , an -135 actually present.

Thed3bi1Ebov~sa fat ~ur 5 ed for this-calculation sall be DOWNSCALE TRIP SETPOINT (DTSP) 1.9a The downscale trip setpoint associated with the Rod Block Monitor (RBM) rod block trip setting.

E-AVERAGE DISINTEGRATION ENERGY 1.10 E shall be the average, weighted in proportion to the concentration of each radionuclide in the reactor coolant at the time of sampling, of the sum of the average beta and gamma energies per disintegration, in MeV, for isotopes, with half lives greater than 15 minutes,' making up at least 95% of the total noniodine activity in the coolant.

EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME 1.11 The EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ECCS actuation set-point at the channel sensor until the ECCS equipment is capable of performing its safety function, i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc. Times shall include diesel generator starting and sequence loading delays where applicable. The response time may be measured by any series of sequential, overlapping or total steps such that the entire response time is measured.

LIMERICK - UNIT 1 1-2 Amendment No. 37. BB. 87 JAN 3 1 1995

V DEFINITIONS PURGE - PURGING 1.31 PURGE or PURGING shall be the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is required to purify the confinement.

RATED THERMAL POWER 1.32 RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 3458 MWt.

REACTOR ENCLOSURE SECONDARY CONTAINMENT INTEGRITY 1.33 REACTOR ENCLOSURE SECONDARY CONTAINMENT INTEGRITY shall exist when:

a. All reactor enclosure secondary containment penetrations required to be closed during accident conditions are either:

1. Capable of being closed by an OPERABLE secondary containment automatic isolation system, or

2. Closed by at least one manual valve, blind flange, slide gate damper, or deactivated automatic valve secured in its closed position, except as provided by Specification 3.6.5.2.1.

b. All reactor enclosure secondary containment hatches and blowout panels are closed and sealed.

c. The standby gas treatment system is in compliance with the requirements of Specification 3.6.5.3.

d. The reactor enclosure recirculation system is in compliance with the requirements of Specification 3.6.5.4.

e. At least one door in each access to the reactor enclosure secondary containment is closed.

f. The sealing mechanism associated with each reactor enclosure secondary containment penetration, e.g., welds, bellows, or O-rings, is OPERABLE.

9. The pressure within the reactor enclosure secondary containment is less than or equal to the value required by Specification 4.6.5.1.1a.

REACTOR PROTECTION SYSTEM RESPONSE TIME 1.34 REACTOR PROTECTION SYSTEM RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its trip setpoint at the channel sensor until de-energization of the scram pilot valve solenoids. The

/ response time may be measured by any series of sequential, overlapping or total steps such that the entire response time is measured.

REFUELING FLOOR SECONDARY CONTAINMENT INTEGRITY REFUELING FLOOR SECONDARY CONTAINMENT INTEGRITY shall exist when:

a. All refueling floor secondary containment penetrations required to 3 be closed during accident conditions are either:

LIMERICK - UNIT 1 1-6 Amendment No. 53,00,105,106 FEB 1 219

DEFINITIONS REFUELING FLOOR SECONDARY CONTAINMENT INTEGRITY (Continued)

1. Capable of being closed by an OPERABLE secondary containment automatic isolation system, or

2. Closed by at least one manual valve, blind flange, slide gate damper, or deactivated automatic valve secured in its closed position, except as provided by Specification 3.6.5.2.2.

b. All refueling floor secondary containment hatches and blowout panels are closed and sealed.

c. The standby gas treatment system is in compliance with the requirements of specification 3.6.5.3.

d. At least one door in each access to the refueling floor secondary containment is closed.

e. The sealing mechanism associated with each refueling floor secondary containment penetration, e.g., welds, bellows, or O-rings; is OPERABLE.

f. The pressure within the refueling floor secondary containment is less than or equal to the value required by Specification 4.6.5.1.2a.

A ENT eZ ORA REPORTABLE EVENT shall be any of those conditions specified in Section 50.73 to 10 CFR Part 50.

g_\>_~ROD DENSITY ROD DENSITY shall be the number of control rod notches inserted as a fraction of the total number of control rod notches. All rods fully inserted is equivalent to 100% ROD DENSITY.

Jo><_

SHTDOW MARIN SHUTDON MARGIN shall be the amount of reactivity by which the reactor is subcritical or would be subcritical assuming all control rods are fully inserted except for the single control rod of highest reactivity worth which is assumed to be fully withdrawn and the reactor is in the shutdown condition; cold, i.e. 680 F; and xenon free.

J^-5_ STE OUNDARY SITE BOUNDARY shall be that line as defined in Figure 5.1.3-la.

SOURCE CHECK 1.41 A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a radioactive source.

DEC 2 0 1995 LIMERICK - UNIT I 1-7 Amendment No. 40,90,105

REACTIVITY CONTROL SYSTEMS 314.1.5 STANDBY LIQUID CONTROL SYSTEM LIMITING CONDITION FOR OPERATION

.1.5- The Standby liquid control system conssetiin of a mtniirum-rf two. pium ¶D-l corresponding fi ow paths, shall be OPERABLE. .

~

<~APPIC  !:LTY.OPERATIONAL EONDITIONS 1 -e-n 2--

AGTION.-

a. in OPERATIONAL CONDITIOI i-en2)

T. With only one pump and corresponding explosive valve OPERABLE, restore one inoperablc pump and corresponding explosive valve to OPERABLE status within 7 day: or be in a least [lOT 5:lUTDOWI within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

2. With standby liquid control-:sy:tem otherwise inoperable, restore f 1 < the system %o QPERABLE statu4 witi " a hMu-rb be im-b#4cat k0a nsee SHLT4'w.N within the ,irt 1 hnt'rs l SURVEILLANCE REQUIREMENTS 4.1.5 The standby liquid control system shall be demonstrated OPERABLE:

a. At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by verifying that:

I. The temperature of the sodium pentaborate solution is within the limits of Figure 3.1.5-1.

2. The available volume of sodium pentaborate solution is at least 3160 gallons.

3. The temperature of the pump suction piping is within the limits of Figure 3.1.5-1 for the most recent concentration analysis.

Amendmert No . Si . 0 . 87 LIMERICK - UNIT I 3/4 1-19 JAN 3 1 1995

REACTIVITY CONTROL SYSTEMS SURVEILLANCE REOUIREMENTS (Continued)

b. At least once per 31 days by:

1. Verifying the continuity of the explosive charge.

2. Determining by chemical calculation* that the available weight o ff is greater than or equal lbs; the concentra on of sodium pentaborate in solution is less than or equal to 13.8% and within the limits of Figure 3.1.5-1 and; the following equation is satisfied:

C x . E - x 0 2 1 13% wt. 29 atom % 86 gpm where C - Sodium pentaborate solution (% by weight)

Q a Two pump flowrate, as determined per surveillance requirement 4.1.5.c.

E - Boron 10 enrichment (atom % Boron 10)

3. Verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.

c. Demonstrating that, when tested pursuant to Specification 4.0.5, the minimum flow requirement of 41.2 gpm per pump at a pressure of greater than or equal to 1230 +/- 25 psig is met.

d. At least once per 24 months during shutdown by:

1. Initiating at least one of the standby liquid control system loops, including an explosive valve, and verifying that a flow path from the pumps to the reactor pressure vessel is available by pumping demineralized water into the reactor vessel. The replacement charge for the explosive valve shall be from the same manufactured batch as the one fired or from another batch which has been certified by having one of the batch success-fully fired. All injection loops shall be tested in 3 operating cycles.

2. Verify all heat-treated piping between storage tank and pump suction is unblocked.**

e. Prior to addition of Boron to storage tank verify sodium pentaborate enrichment to be added is > 29 atom % Boron 10.

• This test shall also be performed anytime water or boron is added to the solu-tion or when the solution temperature drops below the limits of Figure 3.1.5-1 for the most recent concentration analysis, within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after water or boron addition or solution temperature is restored.

• • This test shall also be performed whenever suction piping temperature drops below the limits of Figure 3.1.5-1 for the most recent concentration analysis, within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after solution temperature is restored.

LIMERICK - UNIT 1 3/4 1-20 Amendment No. %9,f1,O6,91,106 FEB 1 2 1N96

TABLE 3.3.2-1 (Continued)

ISOLATION ACTUATION INSTRUMENTATION ACTION STATEMENTS CTION 20 - Be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

CTION 21 - Be in at least STARTUP with the associated isolation valves closed within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> or be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

CTION 22 - Be in at least STARTUP within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

CTION 23 - In OPERATIONAL CONDITION I or 2, verify the affected system isolation valves are closed within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and declare the affected system inoperable. In OPERATIONAL CONDITION 3, be in at least COLD SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

CTION 24 - Restore the manual initiation function to OPERABLE status within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or

.close the affected system isolation valves within the next hour and declare the affected system inoperable or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

CTION 25 - Establish SECONDARY CONTAINMENT INTEGRITY with the standby gas treatment system operating within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

TION 26 - Close the :affected system isolation valves within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

RECENTL IRAIATE TABLE NOTATIONS Required when (1) handlin the secondary containment, or (2) during operations with a potential for draining the reactor vesse wi t e vessel head removed and fuel in the vessel.

t May be bypassed under administrative control, with all- turbine stop valves closed.

f During operation of the associated .Unit 1 or Unit 2 ventilation exhaust system.

a) DELETED v) A channel may be placed in an inoperable status for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for required surveillance without placing the trip system in the tripped condition provided at least one OPERABLE channel in the same trip system is monitoring that parameter. Trip functions common to RPS Actuation Instrumentation are shown in Table 4.3.2.1-1. In addition, for the HPCI system and RCIC system isolation, provided that the redundant isolation valve, inboard or outboard, as applicable, in each line is OPERABLE and all required actuation instrumentation for that valve is OPERABLE, one channel may be placed in an inoperable status for up to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> for required surveillance without placing the channel or trip system in the tripped condition.

MCT 18 aio IMERICK - UNIT I 3/4 3-16 Amendment No. 2@, 40, 63, fi, 146

TABLE 4.3.2.1-1 (Continued)

ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL OPERATIONAL 3

CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH JTRIP FUNCTIN1 CHECK TEST CALIBRATION SURVEILLANCE REQUIRED

7. SECONDARY CONTAINMENT ISOLATION

a. Reactor Vessel Water Level 1, 2, 3 Low, Low - Level 2 S q R S Q R 1, 2, 3

b. Drywell Pressure## - High c.l. Refueling Area Unit I Ventilation Exhaust Duct Radiation - High S Q R q

2. Refueling Area Unit 2 Ventilation *1 Exhaust Duct Radiation - High S Q R

d. Reactor Enclosure Ventilation S Q R 1, 2, 3 Exhaust Duct Radiation - High

e. Deleted 4-

f. Deleted

g. Reactor Enclosure 1, 2, 3 Manual Initiation N.A. R N.A.

h. Refueling Area

• nitiation N.A. R N.A.

° lgCe NTLY IRRu1*v rf. fLJE RJ u ce when

• R handling in the N. econdary containment, or (2)

M during operations w a poten or draining the reactor vessel with th Iihead remove and uel in the vessel.

• • When not administratively bypassed and/or when any turbine stop valve is open.

1. During operation of the associated Unit I or Unit 2 ventilation exhaust system.

I#These trip functions (Za, 6b,. and 7b) are common to the RPS actuation trip function.

389 1002560 TABLE 3.3.7.1-1 RADIATION MONITORING INSTRUMENTATION MINIMUM CHANNELS APPLICABIE ALARM/TRI P I INSTRUMENTATION OPERABLE CONDITIONS SETPOINT 'ACTION t3 1. Main Control roow ormal 4 I1,2,3 ) Ix lo-5 PCi/c 70 Fresh Air Supply Radiation and * '-

monitor

2. Area Monitors

a. Criticality Monitors

1) Spent Fuel 2 (a) > 5 mR/h and <20R/h(b) 71 Storage Pool

b. Control Room Direct 1 At All tles N.A. (b) 73 Id Radiation Monitor

3. Reactor Enclosure Cooling 72 Water Radiation Monitor 1 At All Times < 3 x Background(b)

Rf.CeN11 TABLE 3.3.7.1-1 (Continued)

RADIATION MONITORING INSTRUMENTATJR TABLE NOTATIONS

• Whe is being handled in the secondary containmen t (a) With fuel in the spent fuel storage pool.

(b) Alarm only.

ACTION STATEMENTS ACTION 70 - With one monitor inoperable, restore the inoperable monitor to the OPERABLE status within 7 days or, within -the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, initiate and maintain operation of the control room emergency filtration system in the radiation isolation mode of operation.

• With two or more of the monitors inoperable,..wthin one hour, initiate and maintain operation of the control room emergency

• filtration system in the radiation mode of operation.

ACTION 7 - With one of the required monitor inoperable, assure a portable continuous monitor with the seae alarm setpofnt is OPERABLE in the vicinity of the installed monitor during any fuel movement.

If no fuel movement is being made, perform area surveys of the monitored area with portable monitoring instrumentation at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

ACTION 72 - With the required monitor inoperable, obtain and analyze at least one grab saMple of the monitored parameter at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

ACTION 73 - With the required monitor inoperable, assure a portable alaraing

• monitor is OPERABLE in the vicinity of the installed monitor or

• perform area surveys of the monitored area with portable monitor-ing instrumentation at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

or cburir%5 opero,+ionS wal-kl ck Pokvvh'eJ ~Pr 011rAM5Il~

feqcdor tesseI uWtfkt .e esse-et he Ile

( Qe( Fot1*nt sseIl LIMERICK - UNIT 1 3/4 3-65 A., i 8 1915

I p

nIAnmTI Fff'Wm" 4.3.7.1h1 TABLE ulliTnolue INCt"INFNTATInU YFIWI1 I ASrF Orniffl"NUTC I OPERATIONAL CHANNEL CONDITIONS FOR CHANNEL FUNCTIONAL CHANNEL WHICH SURVEILLANCEI CHECK- TEST CAL IBRAT ION ISREQUIRED I

1. Hain Control Rom Normal Fresh Air Supply Radiation Monitor S Q R 1, 2, 3,[And

• I

2. Area Monitors

a. Criticality Monitors

1) Spent Fuel Storage S N R (a)

Pool la I

aw b. Control Room Direct S M R At All Times ch Radiation Monitor

3. Reactor Enclosure Cooling Water Radiation Monitor S R(b) At All Times 20 f 0 0"

3901038720 TABLE 4.3.7.1-1 (Continued)

RADIATION MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS TABLE NOTATIONS is being handled in the secondary containren/

(a) With fuel in the spent fuel storage pool.

(b) The initial CHANNEL CALIBRATION shall be performed using one or more of the reference standards certified by the National Bureau of Standards (NBS) or using standards that have been obtained from suppliers that participate in measurement assurance activfties with NBS. These standards shall permit calibrating the system over its intended range of energy and measurement range. For subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration shall be used.

or REC~hl oRR 4Mnsf RED lu f C~cicr less'el wAh tfee rUcsse hencic remaved LIMERICK - UNIT 1 3/4 3-67 AMSr 8L

REACTOR COOLANT SYSTEM 389,002$560 3/4.4.7 MAIN STEAM LINE ISOLATION VALVES LIMITING CONDITION FOR OPERATION 3.4.7 Two main steam line isolation valves (MSIVs) per main steam line shall be OPERAB E with closing times greater than or equal to 3 and less than or equal t

- seconds.

PPLICABILITY: OPERATIONAL CONDITIONS .1,2, and 3.

ACTION:

a. With one or more MSIVs inoperable:

1. Maintain at least one MSIV OPERABLE in each affected main steam line that is open and within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, either:

a) Restore the inoperable valve(s) to OPERABLE status, or b) Isolate the affected main steam line by use of a deactivated MSIV in the closed position.

2. Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />..

b. The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.4.7 Each of the above required MSIVs shall be demonstrated OPERABLE by verifying full closure between 3 and seconds when tested pursuant to Specification 4.0.5.

- B1"TT 1 .1A tm A/ 9 g Lartna$6h - Una I L .01

• S L)

CONTAINMENT SYSTEMS LIMITING CONDITION FOR OPERATION (Continued)

ACTION: (Continued)

b. The combined leakage rate to be in accordance with the Primary Containment Leakage Rate Testing Program for all primary containment penetrations and all primary containment isolation valves that are subject to Type B and C tests, except for: main steam line isolation valves*, valves which are hydrostatically tested, and those valves where an exemption to Appendix J of 10 CFR 50 has been granted, and '00

c. The leakage rate scf per hour for any main steam isolation valve that exceeds 100 scf per hour, and restore the combined maximum pathway leakage to <200 scf per hour, and

d. The combined leakage rate for all containment isolation valves in hydrostatically tested lines which penetrate the primary containment to less than or equal to 1 gpm times the total number of such valves, prior to increasing the reactor coolant system temperature above 200 0 F.

SURVEILLANCE REQUIREMENTS 4.6.1.2 The primary containment leakage rates shall be demonstrated to be in accordance with the Primary Containment Leakage Rate Testing Program, or approved exemptions, for the following:.

a. Type A Test

b. Type B and C Tests (including air locks)

c. Main Steam Line Isolation Valves d; Hydrostatically tested Containment Isolation Valves

• Exemption to Appendix "'J" to 10 CFR Part 50.

OCT 18 2x00 LIMERICK - UNIT I 3/4 6-3 Amendment No. 107, 148, 146

CONTAINMENT SYSTEMS 3/4.6.5 SECONDARY CONTAINMENT REFUELING AREA SECONDARY CONTAINMENT INTEGRITY LIMTTING CONDITION FOR OPERATION 3.6.5.1.2 REFUELING AREA SECONDARY CONTAINMENT INTEGRITY shall be maintained..

APPLICABILTTY: CTID-i9 IN6I-P R q FCTION:

Withoti ~EFUELING AREA SECONDARY CONTAINeENT IHdEGRUY+_uspd handling of in the secondary containment and operations wiEfi apotential for draining the reactor vessel. The provisions of Specifica-tion 3.0.3 are not applicable.

SURVEILLANCE REOUTREMENTS 4.6.5.1.2 REFUELING AREA SECONDARY CONTAINMENT INTEGRITY shall be demonstrated by:

a. Verifying at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> that the pressure within the refueling area secondary containment is greater than or equal to 0.25 inch of vacuum water gauge.

b. Verifying at least once per 31 days that:

1. All refueling area secondary containment equipment hatches and blowout panels are closed and sealed.

2. At least one door in each access to the refueling area secondary containment is closed.

3. All refueling area secondary containment penetrations not capable of being closed by OPERABLE secondary containment automatic iso-lation dampers/valves and required to be closed during accident conditions are closed by valves, blind flanges, slide gate dampers or deactivated automatic dampers/valves secured in position.

c. At least once per 24 months:

Operating one standby gas treatment subsystem for one hour and main-taining greater than or equal to 0.25 inch of vacuum water gauge in the refueling area secondary containment at a flow rate not exceeding 764 cfm.

"'Required _tten (1) irirhidiaeina handled in the refue

~.4&ry ontainment, or (2) during CORE ALTERAZIONS, or. (3)1 driticprt~n Twith : pctcnnt'l for draining th: reactor vc::eli-with-tbe-yessel-head-emoved a aF-ftd i. the tnsecl LIMERICK - UNIT I 3/4 6-47 Anmdmmt ]io. 29, 71 JUL 28 1994

CONTAINMENT SYSTEMS REFUELING AREA SECONDARY CONTAINMENT AUTOMATIC ISOLATION VALVES LIMITING CONDITION FOR OPERATION 3.6.5.2.2 The refueling area secondary containment ventilation system automatic isolation valves shall be OPERABLE.

APPLICABILITY: PETIER73 ACTION:

With one or more of the refueling area secondary containment ventilation system automatic isolation valves inoperable, maintain at least one isolation valve OPERABLE in each affected penetration that is open and within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> either:

a. Restore the inoperable valves to OPERABLE status, or

b. Isolate each affected penetration by use of at least one deactivated valve secured in the isolation position, or

c. Isolate each affected penetration by use of at least one closed manual valve, blind flange or slide gate damper. .- - - r ------.

Otherwise, I e suspend hand 4 in the rfuelin g area secondary containment, and operations with a potential for draining the reactor vessel. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.6.5.2.2 Each refueling area secondary containment ventilation system auto-matic isolation valve shall be demonstrated OPERABLE:

a. Prior to returning the valve to service after maintenance, repair or replacement work is performed on the valve or its associated actuator, control or power circuit by cycling the valve through at least one complete cycle of full travel and verifying the specified isolation time.

b. At least once per 24 months by verifying that on a containment isolation test signal each isolation valve actuates to its isolation position.

c. By verifying the isolation time to be within its limit at least once per 92 days.

LIEns with a pNtential for draining the reaAtor vessel with the Yesel hea

-DEC 2 05 LIMERICK -.UNIT 1 .3/4 6-50 Amendment No. COX07J?

CONTAINMENT SYSTEMS STANDBY GAS TREATMENT SYSTEMCOMMON SYSTEM

• LIMITING CONDITION FOR OPERATION 3.6.5.3 Two independent standby gas treatment subsystems shall be OPERABLE.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3, and ACTION:

a. With one standby gas treatment subsystem inoperable, restore the inoperable subsystem to OPERABLE status within 7 days, or:

1. In OPERATIONAL CONDITION 1, 2, or 3, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. TAJSE7' - ECEWrLY 1RRADUIAroFI

2. su eng-handi f n he secondary con inment* and operations with a potential for draining the reactor vessel. The provisions of Specification 3.0.3 are not applicable.

b. With both standby gas treatment substems inperable dlno of in th e a con or opera ions with a potential for draining the reitor vessel. The provisions of Specification 3.0.3.

are not applicable.

SURVEILLANCE REQUIREMENTS 4.6.5.3 Each standby gas treatment subsystem shall be demonstrated OPERABLE:

a. At least once per 31 days by initiating, from the control room, flow through the HEPA filters and charcoal adsorbers and verifying that the subsystem operates with the heaters OPERABLE.-

e/wfte__.d_ .58  :-4+df222its4 ben hanye-4-h P--

l remswed h- th dandi the1 sn wihehevese-b AA-the LIMERICK - UNIT 1 3/4 6-52 Amendment No. 29, Fa 3 0 MA 40 r

CONTAINMENT SYSTEMS SURVEILLANCE RFOUtRFMFNTS (Continued)

b. At least once per 24* months or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire, or chemical release in any ventilation zone communicating with the subsystem by:

1. Verifying that the subsystem satisfies the in-place penetration and bypass leakage testing acceptance criteria of less than 0.05%

and uses the test procedure guidance in Regulatory Positions C.5.a, C.5.c and C.5.d of Regulatory Guide 1.52, Revision 2, March 1978, and the system flow rate is 5764 cfm + 10%.

2. Verifying within 31 days after removal that a laboratory analysis l of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2 March 1978, shows the methyl iodide penetration of less than______

when tested in accordance with ASTM D3803-1989 at a temperature or 300 C (860F), at a relative humidity of 70% and at a face velocity of 66 fpm.

3. Verify that when the fan is running the subsystem flowrate is 2800 cfm minimum from each reactor enclosure (Zones I and II) and 2200 cfm minimum from the refueling area (Zone III) when tested in accordance with ANSI N510-1980.

4. Verify that the pressure drop across the refueling area to SGTS prefilter is less than 0.25 inches water gage while operating at a flow rate of 2400 cfm +/- 10%.

c. After every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of charcoal adsorber operation by verifying within 31 days after removal that a laboratory analysis of a repre-sentative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, shows the methyl iodide penetration of less than when tested in accordance with ASTM 03803-1989 at a temperature of 300 C (860F), at a relative humidity of 70% and at a face velocity of 66 fpm.

d. At least once per 24 months by:

1. Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than 9.1 inches water gauge while operating the filter train at a flow rate of 8400 cfm +/- 10%.

• Surveillance interval is an exception to the guidance provided in Regulatory Guide 1.52, Revision 2, March 1978. SEP 22 aDO LIMERICK - UNIT 1 3/4 6-53 Amendment No. r, 33, A4, 4-2, 144

CONTAINMENT SYSTEMS REACTOR ENCLOSURE RECIRCULATION SYSTEM LIMITING CONDITION FOR OPERATION 3.6.5.4 Two independent reactor enclosure recirculation subsystems shall be OPERABLE.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3.

ACTION:

a. With one reactor enclosure recirculation subsystem inoperable, restore the inoperable subsystem to OPERABLE status within 7 days, or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

b. With both reactor enclosure recirculation subsystems inoperable, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hors.

SURVEILLANCE REQUIREMENTS 4.6.5.4 Each reactor enclosure recirculation subsystem shall be demonstrated OPERABLE:

a. At least once per 31 days by initiating, from the control room, flow through the HEPA filters and charcoal d verifyin th the subsystem operates properl . D Q+ t r"14l040M Of 30 O0cO

b. At least once per 24* months or (1) after any structural maintena on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire, or chemical release in any ventilation zone communi-cating with the subsystem by:

1. Verifying that the subsystem satisfies the in-place penetration and bypass leakage testing acceptance criteria of less than 0.05%

and uses the test procedure guidance in Regulatory Positions C.5.a, of 9

. 2, March 1978, loww0, =

2. Verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accordance with £I570i Regulatory Position C.6.b of Regulatory Guide 1.52, Revision ( .

March 1978, shows the methyl iodide penetration of less than when tested in accordance with ATM D3803-1989 at a temperature ot 300C (86*F) and a relative humidity of 70%.

3. Verifying a subsystem flow rate during system operation when tested in accordance with ANS N510-1980.

ra~eof 30, 000 M IoO0C/

• Surveillance interval is an exception to the guidance-provi d in Regulatory Guide 1.52, Revision 2, March 1978.

SEP 22 a0o LIMERICK - UNIT 1 3/4 6-55 Amendment No. .74, 144

CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

c. After every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of charcoal adsorber operation by verifying within 31 days after removal that a laboratory analysis of a repre-sentative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2 March 1978, shows, the methyl iodide penetration of less than when tested in accordance with ASTM D3803-1989 at a temperatu of 300C (860F) and a relative humidity of 70%.

d. At least once per 24 months by:

1. Verifying that the pressure drop across the combined prefilter, upstream and downstream HEPA filters, and charcoal adsorber banks is less than 6 in 1.1n^ r natia >jwhilp erating the filter train atj flow cfm + 10) verifying that the prefilter pressure r S T'TrF-1.b water gauge and that the pressure drop across each HEPA is less than 2 inches water gauge.

2. Verifying that the filter train starts and the isolation valves which take suction on and return to the reactor enclosure open on each of the following test signals:

a. Manual initiation from the control room, and

b. Simulated automatic initiation signal.

e. After each complete or partial replacement of a HEPA filter bank by verifying that the HEPA filter bank satisfies the inplace penetration and leakage testing acceptance criteria of less than 0.05% in 0cfj0 980 while operating the system at T21o

f. After each complete or partial replacement of a charcoal adsorber bank by verifying that the charcoal adsorber bank satisfies the inplace penetration and leakage testing acceptance criteria of less than 0.05% in accordance with ANSI N510-1980 for a halogenqt d hydro-

,^ - rt ss while operating the system at t low

(.ac 4f60,000 ox cf2 +/- 109 )

SEP 22 3J00 LIMERICK - UNIT I 3/4 6-56 Amendment No. 4X1, At, 144

PLANT SYSTEMS EMERGENCY SERVICE WATER SYSTEM - COMMON SYSTEM LIMITING CONDITION FOR OPERATION 3.7.1.2 At least the following independent emergency service water system loops, with each loop comprised of:

a. Two OPERABLE emergency service water pumps, and

b. An OPERABLE flow path capable of taking suction from the emergency service water pumps wet pits which are supplied from the spray pond or the cooling tower basin and transferring the water to the associated Unit 1 and common safety-related equipment, shall be OPERABLE

a. ffi 7 TIONAL CONDITIONS 1, 2, and 3 ~ s b 4OPERATIONAL CONDITIONS 4, 5, and APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3, 4, 5, and ACTION:

a. In OPERATION CONDITION 1, 2, or 3:

1. With one emergency service water pump inoperable, restore the inoperable pump to OPERABLE status within 45 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

2. With one emergency service water pump in each loop inoperable, restore at least one inoperable pump to OPERABLE status within 30 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

3. With one emergency service water system loop otherwise inoperable, declare all equipment aligned to the inoperable loop inoperable**, restore the inoperable loop to OPERABLE status with at least one OPERABLE pump within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

ncy service water system loop provided confirmatory flow testing has been performed. Those diesel generators no aligned to the OPERABLE emergency service water system loop shall be declared inoperable and the actions of 3.8.1.1 taken.

I LIMERICK - UNIT 1 3/4 7-3 Amendment No. F,48, 6, 131 NOV 1 6 1998

PLANT SYSTEMS LIMITING CONDITION FOR OPERATION (Continued)

ACTION: (Continued)

4. With three ESW pump/diesel generator pairs** inoperable, restore at least one inoperable ESW pump/diesel generator pair** to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

5. With four ESW pump/diesel generator pairs" inoperable, restore at least one inoperable ESW pump/diesel generator pair** to OPERABLE status within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

b. In OPERATIONAL CONDITION 4 or 5:

1. With only one emergency service water pump and its associated flowpath OPERABLE, restore at least two pumps with at least one flow path to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or declare the associated safety related-equipment inoperable and take the ACTION required by Specifications 3.5.2 and 3.8.1.2.

c. In T

1. With only one emergency service water pump and its associated flow path OPERABLE, restore at least two pumps with at least one flow path to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or verify adequate cooling remains available for the diesel generators required to be OPERABLE or declare the associated diesel generator(s) inoperable and take the ACTION required by Specification 3.8.1.2. The provisions of Specification 3.0.3

_are

_ not applicable.

SURVEILLANCE REQUIREMENT 4.7.1.2 At least the above required emergency service water-system loop(s) shall be demonstrated OPERABLE:

a. At least once per 31 days by verifying that each valve (manual, power-operated, or automatic) that is not locked, sealed, or otherwise secured in position, is in its correct position.

b. At least once per 24 months by verifying that:

1. Each automatic valve actuates to its correct position on its appropriate ESW pump start signal.

2. Each pump starts automatically when its associated diesel generator starts.

• • An _ESW-pumpdf se Tgenierator pis s n p and its associated diesel generator. If either an ESW pump or its associated diesel generator becomes inoperable, then the ESW pump/diesel generator pair. is inoperable.

LIMERICK - UNIT 1 3/4 7-4 Amendment No. V7, i0,71 JUL 28 1994

PLANT SYSTEMS ULTIMATE HEAT SINK LIMITING CONDITION FOR OPERATION 3.7.1.3 The spray pond shall be OPERABLE with:

a. A minimum pond water level at or above elevation 250' 10" Mean Sea Level, and

b. A pond water temperature of less than or equal to 880F.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3, 4, 5, and ACTION:

With the requirements of the above specification not satisfied:

a. In OPERATIONAL CONDITION 1, 2, or 3, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

b. In OPERATIONAL CONDITION 4 or 5, declare the RHRSW system and the emergency service water system inoperable and take the ACTION required by Specifications 3.7.1.1 and 3.7.1.2.

c.

system inopera eantake the ACTION required by Specification 3.7.1.2. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.7.1.3 The spray pond shall be determined OPERABLE:

a. By verifying the pond water level to be greater than its limit at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

b. By verifying the water surface temperature (within the upper two feet of the surface) to be less than or equal to 88'F:

1. at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> when the spray pond temperature is greater than or equal to 800F; and

2. at least once per 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> when the spray pond temperature is greater than or equal to 85'F; and

3. at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when the spray pond temperature is greater than 32*F.

c. By verifying all piping above the frost line is drained:

1. within one (1) hour after being used when ambient air temperature is below 40'F; or

2. when ambient air temperature falls below 40*F if the piping has not been previously drained.

pnL. ng H-i-edieted Fu-ir. -the sondavuy.- 1 4 JU14 I99 LIMERICK - UNIT I 3/4 7-5 Amendment No. 4&, 44), 90

314.7.2 CONTROL ROOM EKERGENCY FRESH AIR SUPPLY SYSTEM - COMMON SYSTM LIMITING CONDTTION FOR OPERATION 3.7.2 Two independent control room emergency fresh air supply system subsystems shall be OPERABLE.

APPLICABILITY: All OPERATIONAL CONDITIONS and5 ACTION:

a. In OPERATIONAL CONDITION 1, 2, or 3 with one control room emergency fresh air supply subsystem inoperable, restore the inoperable subsystem to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

b. In OPERATIONAL CONDITION 4, 5, o

1. With one control room emergency fresh air supply subsystems inoperable, restore the inoperable subsystem to OPERABLE status within 7 days or initiate and maintain operation of the OPERABLE subsystem in the radiation isolation mode of operation.

2. With both control ro enc sh air supply subsyt in rable, suspend handling of i~idcatT idh M~i) W the secondary^cni9Ent and operations fhi

~bAEin p ential for draining the reactor vessel.I ions of 4ecification 3.0:3 are'not applicable 2 SURVEILLANCE REQUIREMENTS 4.7.2 Each control room emergency fresh air supply subsystem shall be demonstrated OPERABLE:

a. At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by verifying the control room air tempera-ture to be less than or equal to 85 F effective temperature.

b. At least once per 31 days on a STAGGERED TEST BASIS by initiating, from the control room, flow through the HEPA filters and charcoal adsorbers and verifying that the subsystem operates with the heaters OPERABLE.

c. At least once per 24** months or (1) after any structural maintenance l on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire, or chemical release in any ventilation zone communicating with the subsystem by:

1. Verifying that the subsystem satisfies the in-place penetration and bypass leakage testing acceptance criteria of less than 0.05% and uses the test procedure guidance in Regulatory Positions C.5.a, C.5.c, and C.S.d of Regulatory Guide 1.52, Revision 2, March 1978, and the system flow rate is 3000 cfm 10%.

Regulatory Guide 1.52, Revision 2, March 1978.

LIMERICK - UNIT 1 3/4 7-6 Amendment No. f0, 71 JUL 28 B9

PLANT SYSTEMS SURVEILLA H E-RE IREMETS (C.ntin'ued) -

2. Verifying within'31 days after removal that a laboratory analysis (IO7 of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2 March 1978, shows the methyl iodide penetration of less than when tested in accordance with ASTM D3803-1989 at a temperature of 300 C (860F) and a relative humidity of 70%.

3. Verifying a subsystem flow rate of 3000 cfm + 10% during subsystem operation when tested in accordance with ANSI N510-1980.

d. After every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of charcoal adsorber operation by verifying within 31 days after removal that a laboratory analysis of a repre-sentative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revisions March 1978, shows the methyl iodide penetration of less than t..&31=-.-_ .- 1-4) when tested in accordance with ASTM D3803-1989 at a emperature of 300C (860F) and a relative humidity of 70%.

e. At least once per 24 months by:

1. Verifying that the pressure drop across the combined prefilter, upstream and downstream HEPA filters, and charcoal adsorber banks is.less than 6 inches water gauge while operating the subsystem at a flow rate of 3000 cfm + 10%; verifying that the prefilter pressure drop is less than 0.8 inch water gauge and that the pressure drop across each HEPA is less than 2 inches water gauge.

2. Verifying that on each of the below chlorine isolation mode actuation test signals, the subsystem automatically switches to the chlorine isolation mode of operation and the isolation valves close within 5 seconds:

a) Outside air intake high chlorine, and b) Manual initiation from the control room.

3. Verifyin that o onas-r~mFgy~ysem iASh~iSE;liswitches to isolation mode of operatn n t e control room is maintained at a positive pressure of at least 1/8 inch water gauge relative to the turbine enclosure and auxiliary equipment room and outside atmosphere during subsystem operation with an outdoor I air flow rate less than or equal to 525 cJir r W)nQfl1c Inla+,i torx frovt., e Con rArv Fo~

SEP 2 2 144 LIMERICK - UNIT 1 3/4 7-7 Amendment No.: 422, A,O, 144

ELECTRICAL POWER SYSTEMS A.C. SOURCES - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.8.1.2 As a minimum, the following A.C. electrical por sources shall be OPERABLE:

a. One circuit between the offsite transmission network and the onsite Class 1E distribution systeu,-and

b. Two diesel generators each with:

1. A day fuel tank containing a minimtu of 200 gallons of fuel.

2. A fuel storage systm containing a minimm of 33,500 gallons of fuel.

3. A fuel transfer pup.

APPLICABILITY: OPERATIONAL CONDITIONS 4, 5, a F1 ACTION: TfRRADIAIE

a. With less than the above required A.C. electrical r s urces OPERABLE, suspend CORE ALTERATIONS, handling ot in the secondary contain ent, operations with a potent4 l for dra ning the reactor vessel and crane operations over the spent fuel storag pool when fuel assemblies are stored therein. In addition, when In OPERATIONAL CONDITION 5 with the water level less than 22 feet above the reactor pressure vessel flange, imadiately initiate corrective action to restore the required power sources to OPERABLE status as soon as practical.

b. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.8.1.2 At least the above required A.C. electrical power sourc s shall be demonstrated OPERABLE per Surveillance Requirements 4.8.1.1.1, 4.8.1.1.2, and 4.8.1.1.3.

LIMERICK - UNIT 1 3/4 8-9 Amndment No1.32 *1

.SEP 2 8 IN

ELECTRICAL POWER SYSTEMS DC. SOURCES - SHUTDOWN

/

LIMITING CONQIIION FOR OPERATION 3.8.2.2 As a minimum, two of the following four divisions of the D.C.

electrical power sources system shall be OPERABLE with:

a. Division 1, Consisting of:

1. 125-Volt Battery lAI (lAlD101).

2. 125-Volt Battery 1A2 (lA2D101).

3. 125-Volt Battery Charger 1BCA1 (1A1D103).

4. 125-Volt Battery Charger 1BCA2 (1A2D103).

b. Division 2, Consisting of:

1. 125-Volt Battery iBi (iBiD101).

2. 125-Volt Battery 1B2 (1B2D101).

3. 125-Volt Battery Charger 1BCB1 (IB1D103).

4. 125-Volt Battery Charger 1BCB2 (1B2D103).

c. Division 3, Consisting of:

1. 125-Volt Battery IC (1CD101).

2. 125-Volt Battery Charger 1BCC (1CD103).

d. Division 4, Consisting of:

1. 125-Volt Battery 1D (100101).

2. 125-Volt Battery Charger IBCD (IDD103).

APPLICABILITY: OPERATIONAL CONDITIONS 4, 5, and&- FT ACTION:

a. With one or two required battery chargers on one required division inoperable:

1. Restore battery terminal voltage to greater than or equal to the minimum established float voltage within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />,

2. Verify associated Division 1 or 2 float current < 2 amps, or Division 3 or 4 float current < 1 amp within 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br /> and once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter, and

3. Restore battery charger(s) to OPERABLE status within 7 days.

b. With one or more required batteries inoperable due to:

1. One or two batteries on one division with one or more battery cells float voltage < 2.07 volts, perform 4.8.2.1.a.1 and 4.8.2.1.a.2 within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for affected battery(s) and restore affected cell(s) voltage > 2.07 volts within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

!~i~ ir~d+~v~fIl hndlij ~th~tocndary Got~n-~t LIMERICK - UNIT 1 3/4 8-14 Amendment No. 164

ELECTRICAl POWER SYSTEMS LIMITING CONDITION FOR OPERATION (Continued)

ACTION: (Continued)

2. Division 1 or 2 with float current > 2 amps, or with Division 3 or 4 with float current > 1 amp, perform 4.8.2.1.a.2 within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for affected *battery(s) and restore battery float current to within limits within 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br />.

3. One or two batteries on one division with one or more cells electrolyte level less than minimum established design limits, if electrolyte level was below the top of the plates restore electrolyte level to above top of plates within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and verify no evidence of leakage(*) within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. In all cases, restore electrolyte level to greater than or equal to minimum established design limits within 31 days.

4. One or two batteries on one division with pilot cell electrolyte temperature less than minimum established design limits, restore battery pilot cell temperature to greater than or equal to minimum established design limits within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

5. Batteries in more than one division affected, restore battery parameters for all batteries in one division to within limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

6. (i) Any battery having both (Action b.1) one or more battery cells float voltage < 2.07 volts and (Action b.2) float current not within limits, and/or (ii) Any battery not meeting any Action b.1 through b.5, Restore the battery parameters to within limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

c. 1. With the requirements of Action a. and/or Action b. not met, or

2. With less than two divisions of the above required D.C. electrical power sources OPERABLE for reasons other than Actions a. and/or b.,

Suspend CORE ALTERATIONS, handling of in the secondary containment and operations with a potentia r draining the reactor vessel

d. The provisions of Specification 3.0.3 are not ap licable REQUIREMENTSRA AURVEILLANCE 4.8.2.2 At least the above required battery and charger shall be demonstrated OPERABLE per Surveillance Requirement 4.8.2.1.

(*) Contrary to the provisions of Specification 3.0.2, if electrolyte level was below the top of the plates, the verification that there is no evidence of leakage is required to be completed regardless of when electrolyte level is restored.

LIMERICK - UNIT 1 3/4 8-14a Amendment No. 164 l

3892300520 .

ELECTRICAL POWER SYSTEMS LIMITING CONDITION FOR OPERATION (Continued)

APPLICABILITY: OPERATIONAL CONDITIONS 4, .5, andy ACTION:

a. With less than two divisions of the above required Unit 1 A.C. dis-ts energized, suspend CORE ALTERATIONS, handling of in the secondary containment and operations with a

) potefft~iaTi draining the reactor vessel.

b. With less than two divisions of the above required Unit 1 D.C. dis-tr~i~butionystems energized, suspend CORE ALTERATIONS, handling of in the secondary containment and operations with a pot iial~7ibr draining the reactor vessel.

c. With any of the above required Unit 2 and common AC and/or DC distribution system divisions not energized, declare the associated common equipment inoperable, and take the appropriate ACTION for that system.

d. The provisions of Specification-3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.8.3.2 At least the above required power distribution system divisions shall be determined energized at least once per 7 days by verifying correct breaker alignment and voltage on the busses/MCCs/panels.

(,

UNIT 1 3/4 8-20 jj 2 2 139 LIMERICK - Amendment No. 24 I'

,. j;  ! '.:t; .

Unit 1 Limerick Generating Station Technical Specification Insert Insert 1 (Page 1-2):

Table 2.1 of Federal Guidelines Report 11, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion," ORNL, 1989, as described in Regulatory Guide 1.183. The factors in the column headed "effective" yield doses corresponding to the CEDE.

Insert 2 (Page 1-6):

RECENTLY IRRADIATED FUEL 1.35 RECENTLY IRRADIATED FUEL is fuel that has occupied part of a critical reactor core within the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Insert 3 (Page 3/4 1-19) 3.1.5 The standby liquid control system shall be OPERABLE and consist of a minimum of the following:

a. In OPERATIONAL CONDITIONS 1 and 2, two pumps and corresponding flow paths,

b. In OPERATIONAL CONDITION 3, one pump and corresponding flow path.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3 ACTION:

a. With only one pump and corresponding explosive valve OPERABLE, in OPERATIONAL CONDITION 1 or 2, restore one inoperable pump and corresponding explosive valve to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

b. With standby liquid control system otherwise inoperable, in OPERATIONAL CONDITION 1, 2, or 3, restore the system to OPERABLE status within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Page 1 of 2

Insert 4 (Paqes 3/4 6-47, 3/4 6-50, 3/4 6-52,)

When RECENTLY IRRADIATED FUEL is being handled in the secondary containment, or during operations with a potential for draining the reactor vessel, with the vessel head removed and fuel in the vessel Insert 5 (Pages 3/4 7-3, 3/4 7-4, 3/4 7-5, 3/4 7-6, 3/4 8-9, 3/4 8-14, 3/4 8-20)

When RECENTLY IRRADIATED FUEL is being handled in the secondary containment, or during operations with a potential for draining the reactor vessel Page 2 of 2

DEFINITIONS CORE 'ALTERATION 1.7 CORE ALTERATION shall be the movement of any fuel, sources, or reactivity control components within the reactor vessel with the vessel head removed and fuel in the vessel. The following exceptions are not considered to be CORE ALTERATIONS:

a) Movement of source range monitors, local power range monitors, intermediate range monitors, traversing incore probes, or special moveable detectors (including undervessel replacement); and b) Control rod movement, provided there are no fuel assemblies in the associated core cell.

Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe position.

CORE OPERATING LIMITS REPORT 1.7a The CORE OPERATING LIMITS REPORT (COLR) is the unit-specific document that provides the core operating limits for the current operating reload cycle. These cycle-specific core operating limits shall be determined for each reload cycle in accordance with Specifications 6.9.1.9 thru 6.9.12. Plant operation within these limits is addressed in individual specifications.

CRITICAL POWER RATIO 1.8 The CRITICAL POWER RATIO (CPR) shall be the ratio of that power in the assembly which is calculated by application of the (GEXL) correlation to cause some point in the assembly to experience boiling transition, divided by the actal assembly operating power.

DOSE EQUIVALENT 1-131 n mniec Jrve cosi co 1.9 DOSE EQUIVALENT I-131 shall be that concenriI-131, of microcuries per gram, which alone(wou d produce the same as the quantity and isotoni- mixture of 1-131, 1-132, 1-133, an 1-135 actually present.

The cojwersion factors usedo-fr this calculationstaL thasls e$-r-:

IIv11W~.~f Lle DOWNSCALE TRIP SETPOINT (DTSP) 1.9a The downscale trip setpoint associated with the Rod Block Monitor (RBM) rod block trip setting.

E-AVERAGE DISINTEGRATION ENERGY 1.10 E shall be the average, weighted in proportion to the concentration of each radionuclide in the reactor coolant at the time of sampling, of the sum of the average beta and gamma energies per disintegration, in MeV, for isotopes, with half lives greater than 15 minutes, making up at least 95% of the total noniodine activity in the coolant.

EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME 1.11 The EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ECCS actuation set-point at the channel sensor until the ECCS equipment is capable of performing its safety function, i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc. Times shall include diesel generator starting and sequence loading delays where applicable. The response time may be measured by any series of sequential, overlapping or total steps such that the entire response time is measured.

JAN 3 1 1995 LIMERICK - UNIT 2 1-2 Amendment No. 4.4?, 49

.DEFINITIONS PURGE-- PURGING 1.31 PURGE or PURGING shall be the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is required to purify the confinement.

RATED THERMAL POWER 1.32 RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 3458 MWt.

REACTOR ENCLOSURE SECONDARY CONTAINMENT INTEGRITY 1.33 REACTOR ENCLOSURE SECONDARY CONTAINMENT INTEGRITY shall exist when:

a. All reactor enclosure secondary containment penetrations required to be closed during accident conditions are either:

1. Capable of being closed by an OPERABLE secondary containment automatic isolation system, or

2. Closed by at least one manual valve, blind flange, slide gate damper or deactivated automatic valve secured in its closed position, except as provided by Specification 3.6.5.2.1.

b. All reactor enclosure secondary containment hatches and blowout panels are closed and sealed.

c. The standby gas treatment system is in compliance with the requirements __

of Specification 3.6.5.3.

d. The reactor enclosure recirculation system is in compliance with the requirements of Specification 3.6.5.4.

e. At least one door in each access to the reactor enclosure secondary containment is closed.

f. The sealing mechanism associated with each reactor enclosure secondary containment penetration, e.g., welds, bellows, or O-rings, is OPERABLE.

g. The pressure within the reactor enclosure secondary containment is less than or equal to the value required by Specification 4.6.5.1.1a.

REACTOR PROTECTION SYSTEM RESPONSE TIME 1.34 REACTOR PROTECTION SYSTEM RESPONSE TIME shall be the time interval from hen the monitored parameter exceeds its trip setpoint at the channel ensor until de-energization of the scram pilot valve solenoids. The tj response time may be measured by any series of sequential, overlapping or total steps such that the entire response time is measured.

REFUELING FLOOR SECONDARY CONTAINMENT INTEGRITY REFUELING FLOOR SECONDARY CONTAINMENT INTEGRITY shall exist when:

a. All refueling floor secondary containment penetrations required to be closed during accident conditions are either:

LIMERICK - UNIT 2 1-6 Amendment No. 4$,1,6s DEC 2 0 195

DEFINITIONS REFUELING FLOOR SECONDARY CONTAINMENT INTEGRITY (Continued)

1. Capable of being closed by an OPERABLE secondary containment automatic isolation system, or

2. Closed by at least one manual valve, blind flange, slide gate damper or deactivated automatic valve secured in its closed position, except as provided by Specification 3.6.5.2.2.

b. All refueling floor secondary containment hatches and blowout panels are closed and sealed.

c. The standby gas treatment system is in compliance with the requirements of Specification 3.6.5.3.

d. At least one door in each access to the refueling floor secondary containment is closed.

e. The sealing mechanism associated with each refueling floor secondary containment penetration, e.g., welds, bellows, or O-rings, is OPERABLE.

f. The pressure within the refueling floor secondary containment is less than or equal to the value required by Specification 4.6.5.1.2a.

REPORTABLE EVENT A REPORTABLE EVENT shall be any of those conditions specified in Section 50.73 to 10 CFR Part 50.

ROD DENSITY X 3S-ROD DENSITY shall be the number of control rod notches inserted as a fraction 3of the total number of control rod notches. All rods fully inserted is equivalent to 100% ROD DENSITY.

SHUTDOWN MARGIN SHUTDOWN MARGIN shall be the amount of reactivity by which the reactor is subcritical or would be subcritical assuming all control rods are fully inserted except for the single control rod of highest reactivity worth which is assumed to be fully withdrawn and the reactor is in the shutdown condition; cold, i.e. 680 F; and xenon free.

SITE BOUNDARY The SITE BOUNDARY shall be that line as defined in Figure 5.1.3-la.

{14 Dciced3 SOURCE CHECK 1.41 A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a radioactive source.

LIMERICK - UNIT 2 1-7 Amendment No. 11,{8,69 DEC 2 0 195

REACTIVITY CONTROL SYSTEMS 314.1.5 STANDBY LIQUID CONTROL SYSTEM LIMITING CONDITION FOR OPERATION LL

\-1.5 1cztandby liquid control system, consisting of e 1 ;I,3 ., r,1 twu pumpF ell corrc ~'onding flow paths, shell be OPEABLE.

APPLICABILIT.. OPERATIOIAL COHDITIONS i-e ndI

a. In OPERATIONAL CONDITION 1 or 2:

Ia With only one pump and corresponding explosive valve OPERABLE, restore one inoperable p'mp and corrosponding explosiWe valve to 4OPRALEsttus within 7 days or- be in at leas IIOTHTDWN wit!,;

4th next-12 hours.

2. With standby liquid control system otherwise inoperable, restere the system to OPERABLE status within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or be in at least HOT)

SMTDO'N within the next 12 hour:.s SURVEILLANCE REQUIREMENTS 4.1.5 The standby liquid control system shall be demonstrated OPERABLE:

a. At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by verifying that:

1. The temperature of the sodium pentaborate solution is within the limits of Figure 3.1.5-1.

2. The available volume of sodium pentaborate solution is at least 3160 gallons.

3. The temperature of the pump suction piping is within the limits of Figure 3.1.5-1 for the most recent concentration analysis.

JAN 3 1 1995 LIMERICK - UNIT 2 3/4 1-19 Amendment No 26 usa ia

REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

b. At least once per 31 days by:

1. Verifying the continuity of the e losive charge.

2. Determining by chemical an(> si sa calculation* that the ava e weight of i greater than or equal fslbs; the conc e trat1io isdium pentaborate in solution ess than or equal to 13.8% and within the limits of Figure 3.1.5-1 and; the following equation is satisfied:

C x E x 0 > 1 13% wt. 29 atom % 86 gpm where C - Sodium pentaborate solution (% by weight)

Q - Two pump flowrate, as determined per surveillance requirement 4.1.5.c.

E = Boron 10 enrichment (atom % Boron 10)

3. Verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.

c. Demonstrating that, when tested pursuant to Specification 4.0.5, the minimum flow requirement of 41.2 gpm per pump at a pressure of greater than or equal to 1230+25 psig is met.

d. At least once per 24 months during shutdown by:

1. Initiating at least one of the standby liquid control system loops, including an explosive valve, and verifying that a flow path from the pumps to the reactor pressure vessel is available by pumping demineralized water into the reactor vessel. The replacement charge for the explosive valve shall be from the same manufactured batch as the one fired or from another batch which has been certified by having one of the batch success-fully fired. All injection loops shall be tested in 3 operating cycles.

2. Verify all heat-treated piping between storage tank and pump suction is unblocked.**

e. Prior to addition of Boron to storage tank verify sodium pentaborate enrichment to be added is > 29 atom % Boron 10.

• This test shall also be performed anytime water or boron is added to the solu-tion or when the solution temperature drops below the limits of Figure 3.1.5-1 for the most recent concentration analysis, within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after water or boron addition or solution temperature is restored.

• • This test shall also be performed whenever suction piping temperature drops below the limits of Figure 3.1.5-1 for the most recent concentration analysis, within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after solution temperature is restored.

LIMERICK - UNIT 2 3/4 1-20 Amendment No. Z4, ZB, (,.4, 51 FEB 16 1995

TABLE 3.3.2-1. (Continued)

ISOLATION ACTUATION INSTRUMENTATION

.ACTION STATEMENTS kCTION 20 Be in at least HOT SHUTDOWN within .12'hours and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.:'

ACTION Z1 - Be in at least STARTUP with the Associated isolation valves closed within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> or be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the next 24'hours.

ICTION 22 - Be in at least STARTUP within' 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

%CTION 23 In OPERATIONAL CONDITION 1 or 2, verify the affected systeq tiso ation-valves are'.tlosed~witbin'..1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and declare the,.affedted..system inopeable. In OPERATIOVAL-CONDITION 3, be in at'leist COLD`V SHUTDOWN within'12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. '

kCTION 24 -Restore the manual initiation function "to OPERABLE -status.within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or

'close'the affected systemisolation valves withih'tthe next.hour and .declare the affected system inoperable bor be in'at least HOT. SHUTDOWN'within the, next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> atid in COLD SHUTDOWN within the-following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.-

ICTION 25 - Establish SECONDARY"CONTAINMENT I.NTEGRIP w'iith the standby gas treatment system operating within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. .

tCTION 26 - Clo'se the affected system isolation valves within i'hour.

R ~L .TABLE NOTATIONS

• Required when (1) handling econdary containment, or (2) gering operations with a l potential 'for drain or vesse wit e vessel head removed and fuel1 in the vessel.'

-*May be bypassed under administrative control, with all turbine stop valves closed.

1. During operation of the associated Unit i or.Unit 2 ventilation exhaust system.

(a) DELETED I (b) A channel may be placed in an inoperable status for. up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />

• for required surveillance without placing the trip system in the tripped condition provided at least one OPERABLE-channel in the samejtrip systemis monitoring that parameter. Trip functions common to RPS'Actuation Instrumentation are shown in Table 4.3.2.1-1. In addition,'for the HPCI'system'and RCIC system isolation, provided that the redundant isolation Valve, inboard or.'outboard, as applicable, in each line is OPERABLE and all required actuation instrumentation for that valve is OPERABLE, one channel may be placed in an inoperable status for up to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> for required surveillance without placing-the channel or trip system in the tripped condition.

OCT 18 2DOO LIMERICK - UNIT 2 3/4 3-16 Amendment No. 44, 2, 107

TABLE 4.3.2.1-1 (Continued) ..

ISOLATION ACTUATION'INSTRUMENTATION'.SURVEILLANCEWREOUIREMENTS I-

• .HANNEL.. ..

I.OPERATIONAL .

0-4 CHANNE - FUNCTIONAL CHANNEL CONDrTIONSFOR WHICH TIPLLFUNCTION CHECX b .'TEST' . CAI-IBRAiTON SURVEILLANCE'REOUIRE -

C-4 7. SECONDARY.CONTAINMENT'. ISOLATION

a. Reactor Vessel Water Level Low, Low - Level 2 S R 1, 2, 3

b. Orywell Pressure## - High S- .Q.. R 1, 2,.3 c.l. Refueling Area.Unit l.Vent'ilation Exhaust Duct Radiation.- High S .Q. . .  : *#

2. Refueling Area:Unit'2 Ventilation Q Exhaust DuctVRidfiatfon - High' S.

". S.- " - R I ... ,Q

d. Reactor Enclostfre-rVentilati6n ' .. .. ...

(,

Exhaust'.Duct Radiation - Hight R I, 2, 3

.5t.

C-. e. Deleted CI

f. Deleted I

g.9.. .Reactor Enclosure M~anual Initi-ation ' - N.A. '-. R- - N.iA. lj, 2, 3 en 0D h. Refuel ing Area .

en -Maud

-Initatio N.A. - 'R: ,' N.'A. .*

0.

Na equlxeJhen- I and]ing f" *the secondary' containment, or..(2 (D

w a poten I -drain fngth'e reactor.vessel with- the-T:ve se

-.. a heaB~remove' fuel inh the'vessel.;

m1 **When not idministratively bypassed and/or when any turbine. stop valve is.open.'

!-4

1. During operation of the associated Unit-l or Unit 2 venti-lation exhaust system.,

1. 1. These trip functions (2a, 6bo a'nd 7b) are common to 'the RPS actuation -trip .'function.

TABLE 3.3.7.1-1 RADIATIONMORITORINGINSTRUMENTATION I.I MINIMUM CHANNELS APPLICABLE ALARM/TRIP bN INSTRUMENTATION OPERABLE CONDITIONS SETPOINT ACTION' 1.- Main Control Room Normal 4 1 x lo-o PI 70 Fresh Air Supply Radiation and *l Monitor

2. Area Monitors

a. Criticality Monitors

1) Spent Fuel 2 * (a) > 5 R/hAhd <20R/h(b) 71 Storage Pool

b. Control Room Direct 1 At All Times .A.(b) 73 wA Radiation Monitor wa

3. Reactor Enclosure Cooling 6

Water Radiation Monitor 1 At All Times < 3 x Backbround(b) 72 Ot

.11

TABLE 3.3.7.1-1 (Continued)

RADIATION MONITORING INSTRUMENTATION TABLE NOTATIONS is being handled in the secondary containmen  ;

(a) With fuel in the sspent fuel storage pool. -

(b) Alarm only.

ACTION STATEMENTS ACTION 70 - With one monitor inoperable, restore the inoperable monitor to the OPERABLE status within 7X days or, within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, initiate and maintain operation of the control room emergency filtration system in the radiation isolation mode of operation.

With two or more of the monitors inoperable, within one hour, initiate and maintain operation of the control room emergency filtration system in the radiation mode of operation.

ACTION 71 - With one of the required monitor inoperable, assure a portable continuous monitor with the same alarm setpoint is OPERABLE in the vicinity of the installed monitor during any fuel movement.

If no fuei movement is being made, perform area surveys of the monitored area with portable monitoring instrumentation at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

ACTION 72 - With the required monitor inoperable, obtain and analyze at least one grab sample of the monitored parameter at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

ACTION 73 - With the required monitor inoperable, assure a portable alarming monitor is OPERABLE in the vicinity of the installed monitor or perform area surveys of the monitored area with portable monitor-ing instrumentation at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

C411d 4

-de Ose tec,,j remo _e AUG 2 5 WS LIMERICK- UNI 2 3/4 3-655

IABLE JiLL1 RMDIATION MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS OPERATIONAL CHANNEL CONDITIONS FOR

'1 CHANNEL FUNCTIONAL CHANNEL WIHICH SURVEILLANCE n CLCK TEST- CALIBRATION IS REQUIRED n

I. Main Control Room Nomal Fresh Air Supply Radiation R 1, 2, 3,o ?and

• Monitor S q w 2. Area Monitors P.&

a. Criticality Monitors S N R (a)

I) Spent Fuel Storage Pool L S R At All TIies

b. Control RooM Direct ata w

MI Radiation Monitor

3. Reactor Enclosure Cooling S At All Tites Vater Radiation Monitor t.1

.)

Lo

TABLE 4.3.7.1-1 (Continued)

RADIATION MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS RECEiA'rLy WI/bA QW TABLE NOTATIONS tWherinlrrFdi~tcd =ui is being handled in the secondarycontainm e (a) With fuel in the spent fuel storage pool.

(b) The initial CHANNEL CALIBRATION shall be performed using one or more of the reference standards certified by the National Bureau of Standards (NBS) -or using standards that have been -obtained from suppliers that participate in measurement assurance activities with HBS. These standards shall permit calibrating the system over its intended range of energy and measurement range. For subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration shall be used.

Aor cLhflctJ -fc,;%

e;pLuc(- ti(.sr6 I)sA pcdemt I geG

( wbc ILe IictIrve.s - aA ~c~(.{,l dns{seX LIMERICK - UNIT 2 3/4 3-67 AUG 25 DU

REACTOR COOLANT SYSTEM 3/4.4.7 MAIN STEAM LINE ISOLATION VALVES LIMITING CONDITION FOR OPERATION -

3.4.7 Two main steam line isolation valves (MSIVs) per main steam line shall be OPERABLE with closing times greater than or equal to 3 and less than or equal tVPiseconds.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3.

ACTION:

a. With one or more MSIVs inoperable:

1. Maintain at least one MSIV OPERABLE in each affected main steam line that is open and within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, either:

a) Restore the inoperable valve(s) to OPERABLE status, or b) Isolate the affected main steam line by use of a deactivated MSIV in the closed position.

2. Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

b. The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.4.7 Each of the above required MSIVs shall be demonstrated OPERABLE by verifying full closure between 3 andf $seconds when tested pursuant to Specification 4.0.5.

AU6 2 5 HIS LIMERICK - UNIT 2 3/4 4-23

CONTAINMENT SYSTEMS LIMITING CONDITION FOR OPERATION (Continued)

TION: (Continued)

b. The combined leakage rate to be in accordance with the Primary Containment Leakage Rate Testing Program for all primary containment penetrations and all primary containment isolation valves that are subject to Type B and C tests, except for: main steam line isolation valves*, valves which are hydrostatically tested, and those valves where an exemption to Appendix J of 10 CFR 50 has been granted, and

c. The leakage rate to 11.5-scf per hour for any main steam isolation valve that exceeds 100 sc hour, and restore the combined maximum pathway leakage to <200 scf per hour, and

d. The combined leakage rate for all containment isolation valves in hydrostatically tested lines which penetrate the primary containment to less than or equal to 1 gpm times the total number of such valves, prior to increasing reactor coolant system temperature above 2000F.

SURVEILLANCE REQUIREMENTS.

.4.6.1.2 The primary containment leakage rates shall be demonstrated to be in accordance with the Primary-Containment Leakage Rate Testing Program, or approved exemptions, for the following:

a. Type A Test

b. Type B and C Tests (including air locks)

c. Main Steam Line Isolation Valves

• d. Hydrostatically tested Containment Isolation Valves

• Exemption to Appendix "J" to 10 CFR Part 50.

OCT 18 MMO LIMERICK - UNIT 2 3/4 6-3 Amendment No. -&3, q4, 84-, 107

CONTAINMENT SYSTEMS 314.6.5 SECONDARY CONTAWNMENT REFUELING AREA SECONDARY CONTAINMENT INTErRtm LIMITING CONDITION FOR OPERATION 3.6.5.1.2 REFUELING AREA SECONDARY CONTAINMENT INTEGRITY shall be maintained.

APPLICABILITY:

ACTION-JiFhout UELING AREA SECONDARY CONTAINMENT INTEGRITY susnd handling of

{fi~gRati3 iinthe -secondaryvcontainment, a -DF0ib3nd-operations t~inal for draining the reactor vesse

• sions of Specifica-tion 3.0.3 are not applicable.

SURVEILLANCE-REQUIREMENTS 4.6.5.1.2 REFUELING AREA SECONDARY CONTAINMENT INTEGRITY shall be demonstrated by:

a. Verifying at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> that the pressure within the refueling area secondary containment is greater than or equal to 0.25 inch of vacuum water gauge.

b. Verifying at least once per 31 days that:

1. All refueling area secondary containment equipment hatches and blowout panels are closed and sealed.

2. At least one door in each access to the refueling area secondary containment is closed.

3. All refueling area secondary containment penetrations not capable of being closed by OPERABLE secondary containment automatic iso-lation dampers/valves and required to be closed during accident conditions are closed by valves, blind flanges, slide gate dampers or deactivated automatic dampers/valves secured in position.

c. At least once per 24 months:

Operating one standby gas treatment subsystem for one hour and main-taining greater than or equal to 0.25 inch of vacuum water gauge in the refueling area secondary containment at a flow rate not exceeding 764 cfm.

'keu--L-al

cecondr cota ant, or (2 dui CEa1&TJr()9Irn o rtions

-or Arninimjthe-reactor essel with-the esselhead removed.

l n fas iran ths ac men ci. ..

LIMERICK - UNIT 2 3/4 6-47 L3menmt No. 34 JUL 28 Em

CONTAINMENT SYSTEMS REFUELING AREA SECONDARY CONTAINMENT AUTOMATIC ISOLATION VALVES LIMITING CONDITION FOR OPERATION 3.6.5.2.2 The refueling area secondary containment ventilation system automatic isolation valves shall be OPERABLE.

APPLICABILITY:

ACTION:

With one or more of the refueling area secondary containment ventilation system automatic isolation valves inoperable, maintain at least one isolation valve OPERABLE in each affected penetration that is open and within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> either:

a. Restore the inoperable valves to OPERABLE status, or

b. Isolate each affected penetration by use of at least one deactivated valve secured in the isolation position, or

c. Isolate each affected penetration by use of at least one closed manual valve, blind flange or slide gate damper. Rf'&,Vfl-rpisetental suspend handling aoh"' in'the rgel fifogs an-Rdao y contimnt 'and operations with a potential for draining the reactor vessel. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.6.5.2.2 Each refueling area secondary containment ventilation system auto-matic isolation valve shall be demonstrated OPERABLE:

a. Prior to returning the valve to service after maintenance, repair or replacement work is performed on the valve or its associated actuator, control or power circuit by cycling the valve through at least one complete cycle of full travel and verifying the specified isolation time.,

b. At least once per 24 months by verifying that on a containment isolation test signal each isolation valve actuates to its isolation position.

c. By verifying the isolation time to be within its limit at least once per 92 days.

-*Required when (}) ir-radiate ade4sSmR~~gwe- auli-en

-secondary contei~mnt, or (2)durinq COEAT~LTOS r(2 during epcratin

.with a -oteta1 for-draining the reactor upor-sel with the 'ere ha remo'.ed 12nd fu9l in the essel .

DEC 2 0 M LIMERICK - UNIT 2 3/4 6-50 Amnendment -No. 14,69

CONTAINMENT SYSTEMS STANDBY GAS TREATMENT.,SYSTEM -. COMMON SYSTEM LIMITING CONDITION FOR OPERATION 3.6.5.3' Two6Andependeht standby .gas'treatment subsystems 'shall be OPERABLE.

APPLICABILITY: OPERATIONAL.CbNDITIONS 1, 2, 3, andl9 T, ii

.ACTION:

a. In OPERATIONAL CONDITION 1,'2, or 3: '

1. With-the Unit 1 diese'ge'nerator'.for one standby .gas treatment

'sub'sys;tem`apopeiablefor '-more 'than30das,be in'at least HOT SHUTDOWN within thenext 1Žhours and in 'COLD SHUTDOWN within the',.fol&i8wing '24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />'. The :'rpvisions:OT',3ecification,3.0.4

• are not applicable, r o e.,

2. With o'nestandby ga'sA'reatment:;subsystem inoperable,' restore

~

the' nop '6 e-subsyst"e"'to"OPERABLE;:statu' within'7 days, or be in.at'.-least ,,HOT SUTDOWN',v'ithin the next' 1 hours1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and in' COLD SHUTDOWN wi thin the`Iiollow'ing;24 'h .- - "

3. Witih one' standby gas t;en-t subsystem inerabie and the other standby.gas tteatsent subsystem with an inoperable,.Unit'1 diesel generator., restor"e'thie'noperable'sub'system 'to OPERABLE' status .or'.rest'orethe inoperable Unit":diesel'g'eneiato:tor OPERABLE status'w'ithin72.hours, ;or b'e inNat'least HOTSHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

4. With.the Unit.1 diesel generators for both.-standby gas treatment'-system.'subsystems,-inoperable.for more than..

72:hours, be iin at.leas't"IOT.'SHUTDOWN'within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD. SHUTDOWN within'the following,'24 'hours.

b. PTTINL~TI~

1. With one standby gas'treatment subsystem inoperable, restore E a nri'Y the inoperable subsystem to 'OPERABLE status within 7 days, or qr6 suspeund_.,angnof in the secondary containment,

'2anoperatons with a.potential for,'draining he reac or vessel. The provisions of Specification 3.0.3 are not applicable.

2. With both s&tandb gas treatment subsystems inoperable, suspend iizadlinq of in the secondary containment,?

' and operations with a potential for draining reactor ssel. The provisions of Specification 3.0.3'are not applicable.

d-Reqtired-when-(-1-)-i rrada te+ el-4s bei-ng-handled4rn-the-refue-1ing-area-ccndaiy G containmcnt, or (2) during CORE ALTERAIONS, or (3) during-operatieos K -~h--otent~a! _

for dranig heracor"ge with tha uessclhe-the _ _*eid LIMr IoCKd -nd fu2l in th-3 6-5c2. _ _ _

LIMERICK ~-UIRIT 2 "' 3/6-2AUS 2 5 W~

CONTAINMENT SYSTEMS SURVEILLZANCE RFOUIREMENIS (Conti nued)

b. At least once per 24*.months or .(1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire, or-chemical release in any ventilation zone communicating with the subsystem by:

1. Verifying that the subsystem satisfies the in-place penetration and bypass leakage testing acceptance criteria of less than 0.05%

and uses the test procedure guidance in Regulatory Positions C.5.a, C.5.c and C.5.d of Regulatory.Ouide 1.52, Revision 2-, March 1978, and the system flow rate is 5764 cfm +/-+10%.

2. Verifying within3 days afte rremoval tha6t alaboratory analysis

. of a representative'car.bon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide' 1;52, Revision 2 ,

March 1978, showsthe.methy iodide penetration of lessthar(.*

wheh te~sted in .ac'cor`.da1e 'with ASTM D3803-1989 at a temperatureof

.300 C(86°F),.at a rative.,htufnidity .of 70% and at -aface-velocity of 3b6. MY86!F,. .at a . 14" ., umidity ' of ;

3. Verify that when.the fan is running the subsystem:flowrate is 2800 cfm minimum from each reactor enclosure (Zones I and II) and 2200 cfm minimum'from the refueling are'a (Zone III) When tested in accordance with ANSI N510-1980.

4. 'Verify that the pressure urop r e refueling area to SGTS prefilter is less thdh 0.25 inches water gage while operating at a flow rate of 2400 cfm + 10%.

c. After every 720'hours of charcoal adsorber operation by verifying within 31 days after removal that a laboratory analysis of a repre-sentative carbon sample obt'i6ind in accordance with Regulatory Position C.6.b of Regulatory' Guide 1.52, Revision 2, March-1978.

shows the methyl iodide 'peetration of less than 5when tested in accordance with ASTM D3803-1989 at a temperature of 300C C86 0F),

at a relative humidity of 70% and at a face velocity of 66 fpm.

d. At least once per 24 months by:

1. Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than 9.1 inches water gauge while operating the filter train at a flow rate of 8400 cfm +/- 10%.

• Surveillance interval is an exception to the guidance provided in Regulatory Guide 1.52, Revision 2, March 1978. SEP 22 2000 LIMERICK - UNIT 2 3/4 6-53 Amendment -34,9ra,106

CONTAINMENT SYSTEMS REACTOR ENCLOSURE RECIRCULATION SYSTEM LIMITING CONDITION FOR OPERATION 3.6.5.4 Two independent reactor enclosure recirculation subsystems shall be OPERABLE.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3.

ACTION:

a. With one reactor enclosure recirculation subsystem inoperable, restore the inoperable subsystem to OPERABLE status within 7 days, or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

b. With both reactor enclosure recirculation subsystems inoperable, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

SURVEILLANCE REQUIREMENTS .

4.6.5.4 Each reactor enclosure recirculation subsystem shall be demonstrated OPERABLE:

a. At least once per 31 days by initiating, from the control room, flow through the HEPA filters and charcoal adsorbers and verifyingdveify the subsystem operates properly 0*0

b. At least once per 24* months or (1) aft ructural maintenances -

on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire, or chemical release in any ventilation zone communi-cating with the subsystem by:'

1. Verifying that the subsystem satisfies the in-place penetration and bypass leakage testing acceptance criteria of less than 0.05%

and uses the test procedure guidance in Regulatory Positions C.5.a, of Rggul' w iujde=52 vision 2, March 1978, flow - 60,000 cfm 10%.)

2. Verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in'accordance with }52 Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2 March 1978, shows the methyl iodide penetration of less than2.-

when tested in accordance with ASTM D3803-1989 at a temperature of 30'C (86°F) and a relative humidity of 70%.

3. Verifying a subsystem flow rat during system operation when tested in accordance with AlISI ZF=510-19 80.

cjitlin r+tj oS30,c' Cffb to L 0()7Cf )

• Surveillance interval is an exception to the guidance provided in Regulatory Guide 1.52 Revision 2, March 1978.

SEP 22 200 LIMERICK - UNIT 2 3/4 6-55 Amendment No. a4, 106

CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

c. After every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of charcoal adsorber operation by verifying within 31 days after removal that a laboratory analysis of a repre-sentative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, shows the methyl iodide penetration of less tharN when tested in accordance with ASTM D3803-1989 at a temperatur of 300 C (860F) an a relative humidity of 70%.

d. At least once per 24 months by:

1. Verifying that the pressure drop across the combined prefilter, upstream and downstream HEPA filters, and charcoal adsorber banks is less than 6 inch~teAaugeyhle.operating the filter train atj6flow§o- f(60,000 cfm + 10J) verifying that the prefilter pressure drop 15eSS'tnhn-Z-.81" water gauge and that the pressure drop across each HEPA is less than 2 inches water gauge.

2. Verifying that the filter train starts and the isolation valves which take suction on and return to the reactor enclosure open on each of the following test signals:

a. Manual initiation from the control room, and

b. Simulated automatic initiation signal.

e. After each complete or partial replacement of a HEPA filter bank by verifying that the HEPA filter bank satisfies the inplace penetration and leakage testing acceptance criteria of less than 0.05% in cgei hgS;lEfiW-1980 while operating the system at }J low cfm,.) 0_.

f. After each complete or partial replacement of a charcoal .adsorber { l bank by verifying that the charcoal adsorber bank satisfies the inplace penetration and leakage testing acceptance criteria of less than 0.05% in accordance with ANSI N510-1980 for a halogenated hydro-as while operating the system at(a)flow I USEP 2 2 2406 LIMERICK - UNIT 2 3/4 6-56 Amendment No. 34, 106

PLANT SYSTEMS EMERGENCY SERVICE WATER SYSTEM - COMMON SYSTEM LIMITING CONDITION FOR OPERATION 3.7.1.2 At least the following independent emergency service water system loops, with each loop comprised of:

a. Two OPERABLE emergency service water pumps, and

b. An OPERABLE flow path capable of taking suction from the emergency service water pumps wet pits which are supplied from the spray pond or the cooling tower basin and transferring the water to the associated Unit 2 and common safety-related equipment, shall be OPERABL `rwo

. OPERATIONAL CONDITIONS 1, 2, and loopr b -- PERATIONAL CONDITIONS 4, 5, and X R APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3, 4, 5, andi ' '

ACTION:

a. In OPERATION CONDITION 1, 2, or 3:

1. With one emergency service water pump inoperable, restore the inoperable pump to OPERABLE status within 45 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

2. With one emergency service water pump in each loop inoperable, restore at least one inoperable pump to OPERABLE status within 30 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

3. With one emergency service water system loop otherwise inoperable, declare all equipment aligned to the inoperable loop inoperable**, restore the inoperable loop to OPERABLE status with at least one OPERABLE pump within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

( when handling irradiatcd f ie'ci int.

• • The diesel generators may be aligned to the OPERABLE emergency service water system loop provided confirmatory flow testing has been performed. Those diesel generators not aligned to the OPERABLE emergency service water system loop shall be declared inoperable and the actions of 3.8.1.1 taken.

An I Amendment No. 48, O, 92 LIMERICK - UNIT 2 3/4 7-3

PLANT SYSTEMS LIMITING CONDITION FOR OPERATION (Continued)

ACTION: (Continued)

4. With three ESW pump/diesel generator pairs** inoperable, restore at-least one Inoperable. ESW pump/diesel generator-pair** to OPERABLE status'within, 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT.SHUTDOWN within the next 1-.hotirs and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

5. With four ESW pump/diesel generator pairs** inoperable, restore at least one inoperable ESW pump/diesel generator pair** to.

OPERABLE status within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or be in at least HOT SHUTDOWN within the next 12-4h6urs and in -COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

b. In OPERATIONAL CONDITION 4 or 5:

1. With only one.emergency service water pump and its associated .

flow path OPERABLE, restore at least two..pumps with at least one flow path':tb:OPERABLE status within.72.hours or declare the associated sAfety related equipment lnop'eeable'and take'the ACTION required:by Specifications 3.52 'and 3.8.1.2.

c. J:5 s ..

1. WiT .on y oneie ncy service water pump and its associated flow path OPERABLE, restore at least two pumps with at least one flow path'toOPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or verify adequate cooling remains available for the diesel generators required to be OPERABLE or declare the associated diesel genera-tor(s) inoperable and take the ACTION required by Specifica-tion 3.8.1.2. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENT 4.7.1.2 At least the above required emergency service water system loop(s) shall be demonstrated OPERABLE:

a. At least once per 31 days by verifying that each valve (manual, power-operated, or automatic) that is not locked, sealed, or otherwise secured in position, is in its correct position.

b. At least once per 24 months by verifying that:

1. Each automatic valve actuates to its correct position on its appropriate ESW pump start signal.

2. Each pump starts automatically when its associated diesel generator starts.

• AnESWp dUfpigenerator lp t6nitsocfrESF p ump itsandassociated diesel generator. If either an ESW pump or its associated diesel generator becomes inoperable, then the ESW pump/diesel generator pair is.inoperable.

LIMERICK - UNIT 2 3/4 7-4 Amendment No.34 JUL 28 1994

PLANT SYSTEMS ULTIMATE HEAT SINK -

LIMITING CONDITION FOR OPERATION 3.7.1.3 The spray pond shall be OPERABLE with:

a. A minimum pond water level at or above elevation 250-10" Mean Sea Level, and

b. A pond water temperature of less than or equal to 880 F.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3,34, 5, a a ACTION:

With the requirements of the above specification not satisfied:

a. In OPERATIONAL CONDITION 1, 2, or 3, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

b. In OPERATIONAL CONDITION 4 or 5, declare the RHRSW system and the emergency service water system inoperable and take the ACTION required by Specifications 3.7.1.1 and 3.7.1.2.

c. eclare the emergency service water system inopefrafilte-aikd the ACTION required by Specification 3.7.1.2.

The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.7.1.3 The spray pond shall be determined OPERABLE:

a. By verifying the pond water level to be greater than its limit at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

b. By verifying the-water surface temperature (within the upper two feet of the surface) to be less than or equal to 880F:

1. at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> when the spray pond temperature is greater than or equal to 800 F; and

2. at least once per 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> when the spray pond temperature is greater than or equal to 850F; and

3. at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when the spray pond temperature is greater than 320F.

c. By verifying all piping above the frost line is drained:

1. within one (1) hour after being used when ambient air temperature is below 40'F; or

2. when ambient air temperature falls below 400 F if the piping has not been previously drained.

k. -aerem -h44eg-irrad-iatedfueli-in-tbe-jecondtr-eonta-iinment-. 3

&t 1 4199 LIMERICK - UNIT 2 3/4 7-5 Amendment No. 54

I PLAnT SYSTEMS.

3/4.7.2 CONTROL ROOM EMERGENCY FRESH AIR SUPPLY SYSTEM - COMMON SYSTEM LIMITING CONDITION FOR OPERATION 3.7.2 Two independent control room emergency fresh air supply system subsystems shall be OPERABLE. -

APPLICABILITY: All OPERATIONAL CONDITIONS and -

ACTION:

a. In OPERATIONAL CONDITION 1, 2, or 3:

1. With the Unit 1 diesel generator for one control room emergency fresh air supply subsystem inoperable for sore than 30 days, be in at least HOT SHUTOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The provisions of Specification 3.0.4 are not applicable.

2. With one control room emergency fresh air supply subsystem inoperable, restore the inoperable subsystem to OPERABLE status within 7 days, or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

3. With one control room emergency fresh air supply subsystem inoperable and the other control room emergency fresh air supply subsystem with an inoperable Unit 1 diesel generator, restore the inoperable subsystem to OPERABLE status or restore the Unit 1 diesel generator to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

4. With the Unit 1 diesel generators for both control room emergency fresh air.supply subsystems inoperable for more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24.hours.

b. In OPERATIONAL CONDITION 4, 5 or: 3 jj )

1. With one control room emergency fresh air supply subsystem inoperable, restore the inoperable subsystem to OPERABLE status within 7 days, or initiate and maintain operation of the OPERABLE subsystem in the radiation isolation mode of operation.

006 2. With both both cont n fresh air sPly subsystem rable, suspen handling of Mtential ar in the secondary con irn operations a potential for draining the reactor vessel. 2 (ication 3.0.3 are not applicable )

nple0ia:ta LIMERICK - UNIT 2 3/4 7-6 25 is

PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

2. Verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision ,42',-, -',

March 1978, shows the methyl iodide penetration of less thanC )

when tested in accordance with ASTM D3803-1989 at a temperature of 300C (860F) and a relative humidity of 70%.

3. Verifying a subsystem flow rate of 3000 cfm +/- 10% during subsystem operation when tested in accordance with ANSI N510-1980.

d. After every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of charcoal adsorber operation by verifying within 31 days after removal that a laboratory analysis of a repre-sentative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2. March 1978, shows the methyl iodide penetration of less than- when tested (E '

in accordance with ASTM D3803-1989 at a temperature of 300C (860F) and a relative humidity of 70%.

e. At least once per 24 months by:

1. Verifying that the pressure drop across the combined prefilter, upstream and downstream HEPA filters, and charcoal adsorber banks is less than 6 inches water gauge while operating the subsystem at a flow rate of 3000 cfm +/- 10%; verifying that the prefilter pressure drop is less than 0.8 inch water gauge and that the pressure drop across each HEPA is less than 2 inches water gauge..

2. Verifying that on each of the below chlorine isolation mode actuation test signals, the subsystem automatically switches to the chlorine isolation mode of operation and the isolation valves close within 5 seconds:

a) Outside air intake high chlorine, and b) Manual initiation from the control room.

3. Verif i i that 04- Iie 3-e-s-0 y e. Us e to fhe radfiatn io ation mode of operation and the control room is maintained at a positive pressure of at least 1/8 inch water gauge relative to the turbine enclosure and auxiliary equipment room and outside atmosphere during subsystem operation with an outdoor air flow rate less than or equal to 525 cf Ia) Outsid^ air intakc high radGiaitno , )nd l ( b) nnun1 initiation from control-Fee 8 I 11VC%t ' I n i Et ot a n - ro -'n Av- Cc,--ii-p,)' I A o SEP 2 2 2000 LIMERICK - UNIT 2 3/4 7-7 Amendment No. .34, 106

.ELECTRICAL POWER SYSTEMS A.C. SOURCES.- SHUTDOWN LIMITING CONDITION FOR ,OPERTION E - .-

'38.1.2As a'inimum, the following A.C. electrical power;sources shall be:

OPERABLE:

a.: Dne.;circuit'.between.4he-Of it a nsmistion network and the onsite Class lE distribution tsystfe And;.-. . -

b. Two diesel generators each wi'th

. 1. Aday fuel tank containing a .ini mm of 200 gallons of fuel.  :

2. A fuel storage ;stm cotaininga n .:of 3.n5 of -fuel.

• 0 aln

• 3. A fuel transfer-pump. ...

APPLCABILITY: OPERATIONAL CONDItiONS 4, 15, a d / .. ,--.-----;

• ACTION: -

a. With lesg than te:abovwe i

'OPERABLE, ..suspend CORE ALTERATIONS, handling l e calypwe sburcessp of in the secondary containment, operations w'ith'.a Pino ng thiereactor- vessel Kand crane -p'erat'iois.'rver the spent fuel storage pool6 wheni fuel asse blie's, are'stored.therein. In additibn, W4en in OPERATIONAL'C-O=PION' 5' .wjtth he i at"eevel ,less.haini,22-b feet above

• the'reactorpressure vessel.flange1ii ediately initiate cbrrective actiontrestore t reqUitred power'sour ces to.OPERABLE status as

.soon as .practical.

b. The provisions of Specification'3.0.3 are not appl icable..

SURVEILLANCE REQUIREMENTS 4.8.1.2 At least the above required A;C. electrical power sources shall be demonstrated OPERABLE per Surveillance Requiirement~s 4.8.1.1.1, 4.8.1.1.2,.

and 4.8.1.1.3. --

twhe -hafid449I4at dited-fue-4n-t4-e--se-ndary-cot ntn}.

LIMERICK - UNIT 2 3/4 8-9 AUS 2 5 MO

FELECTRICAL POWER SYSTEMS D.C. SOURCES - SHUTDOWN LIMIING CONDITION FOR OPERATION 3.8.2.2 As a minimum, two of the following four divisions of the D.C.

electrical power sources system shall be OPERABLE with:

a. Division 1, Consisting of:

1. 125-Volt Battery 2A1 (2A1D101).

2. 125-Volt Battery 2A2 (2A2D101).

3. 125-Volt Battery Charger 2BCA1 (2AID103).

4. 125-Volt Battery Charger 2BCA2 (2A2D103).

b. Division 2, Consisting of:

1. 125-Volt Battery 2B1 (2B1D101).

2. 125-Volt Battery 2B2 (2B2D101).

3. 125-Volt Battery Charger 2BCB1 (2lD103).

4. 125-Volt Battery Charger 2BCB2 (2B2D103).

c. Division 3, Consisting of:

1. .125-Volt Battery 2C (2CD101).

2. 125-Volt Battery Charger 2BCC (2CD103).

d. Division 4, Consisting of:

1. 125-Volt Battery 2D (2DD101).

2. 125-Volt Battery Charger 2BCD (2DD103).

APPLICABILITY: OPERATIONAL CONDITIONS 4, 5, and ./5  ?

ACTION:

a. With one or two required battery chargers on one required division inoperable:

1. Restore battery terminal voltage to greater than or equal to the minimum established float voltage within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />,

2. Verify associated Division 1 or 2 float current < 2 amps, or Division 3 or 4 float current < 1 amp within 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br /> and once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter, and

3. Restore battery charger(s) to OPERABLE status within 7 days.

b. With one or more required batteries inoperable due to:

1. One or two batteries on one division with one or more battery cells float voltage < 2.07 volts, perform 4.8.2.1.a.1 and 4.8.2.1.a.2 within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for affected battery(s) and restore affected cell(s) voltage > 2.07 volts within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

I.ihcn handling irradiated f9el in tha secondary onirvent.

LIMERICK - UNIT 2 3/4 8-14 Amendment No. 126!

ELECTRICAL POWER SYSTEMS 1-IMITING CONDITION FOR OPERATION ACTION: (Continued)

2. Division 1 or 2 with float current > 2 amps, or with Division 3 or 4 with float current > 1 amp, perform 4.8.2.1.a.2 within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for affected battery(s) and restore battery float current to within limits within 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br />.

3. One or two batteries on one division with one or more cells electrolyte level less than minimum established design limits, if electrolyte level was below the top of the plates restore electrolyte level to above top of plates within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and verify no evidence of leakage(*) within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. In all cases, restore electrolyte level to greater than or equal to minimum established design limits within 31 days.

4. One or two batteries on one division with pilot cell electrolyte temperature less than minimum established design limits, restore battery pilot cell temperature to greater than or equal to minimum established design limits within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

5. Batteries in more than one division affected, restore battery parameters for all batteries in one division to within limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

6. (i) Any battery having both (Action b.1) one or more battery cells float voltage < 2.07 volts and (Action b.2) float current not within limits, and/or (ii) Any battery not meeting any Action b.1 through b.5, Restore the battery parameters to within limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

c. 1. With the requirements of Action a. and/or Action b. not met, or

2. With less than two divisions of the above required D.C. electrical power sources OPERABLE for reasons other than Actions a. and/or b.,

Suspend CORE ALTERATIONS, handling ofn containment and operations with a pote the secondary raining the reactor vessel.

I.

d. The provisions of Specification 3.0.3 are not applicable.

Li LEL SURVEILLANCE RFOIJREMENTS .

4.8.2.2 At least the above required batteries and chargers shall be demonstrated OPERABLE per Surveillance Requirement 4.8.2.1.

(*) Contrary to the provisions of Specification 3.0.2, if electrolyte level was below the top of the plates, the verification that there is no evidence of leakage is required to be completed regardless of when electrolyte level is restored.

LIMERICK - UNIT 2 3/4 8-14a Amendment No. 126 l

.ELECTRICAL POWER SYSTEMS LIMITING CONDITION FOR OPERATION.(Continued) c) 125-V DC Distribution:.Panels: . 2PPA1  :(2AD102) 2PPA2z (...412AD501) 2PPA3 (2AD162)

2. Unit 2 Divyision 2, Consisting of:

a) 250-V DC:Fuse.Box.-- - 2FB BDiO) b) 250-'V DC," otorf,.ntro, Centers: ':g20Ds 2DB-i c) 5-V.Dt 'DistribBtion Panels:4 2pPBl';..(2BD102)

- .. ,P!PB2.. 2D501) 2PPB3 '."(2BD162)

3. Unit 2 ZDivisjor 3,o Consistingof - i. '

b 125-.'DC "isDi tiolnPAne s-: .2PPCC 02) 2PPC2 ~~2CD5O1

4. -Uni:.t ' i, ' ?PPC3 ' sa(2CD162)I

• .-.b.--'5YDivision a) 125- DFus Box.- i:;.'. -'!.ZFD, bb )',

D1 -,ui

~ZP.DZ 1D51

-s- , PPD3 ~t(2DD162

' uni ind Como visnI, onsitingpof:. -

a) 250-V.DC FuseB'ox FA (AD105)'

b) 125-VDC Dist'r'ib'.tion Panels: IPPAl. (1AD102) 1PPA2 (1AD501)

6. Unit 1 and Common'Division 2; Consisting of:

• a) .250-V DC Fuse Box: . 1FB (lBD105) b) ',125-V DC *DistribUtipn Panels: *PPB1 ,(lBD102)-'

1PPB2 'QPBD50.1).

7. Unit I and Common Division 3, Consisting of:

a) 125-V DC Fuse.Box: 1FC. *(lCD105) b) 125-V DC Distribution Panels: 1PPC1 I' (lCD102).

1PPC2 *(1CD501)

8. . Unit 1 and Common Division 4, Consisting of:

a) 125-V DC Fuse Box; 1FD (lDD105) b) 125-V DC Distribution Panels: 1PPD1 (lDD102) 1PPD2 (IDD5O0).

APPLICABILITY: OPERATIONAL CONDITIONS 4, 5, an ACTION:

a. With less than two divisions of the above required Unit 2 A.C.

distribution systems energized, suspend CORE ALTERATIONS, handling of

-A cgs.,in the secondary containment and operations with a

( Dotential for draining the reactor vessel.

_- -x - ~-

indling irin th c n nd a 4)

LIMERICK - UNIT 2 3/4-8-19 Amendment No. 102

,, +/-.'v>

• ELECTRICAL POWER -SYSTEMS . '

LIMITING CONDITION FOR 'OPERATION '(Conti nue d) . . . . .

, "... .; , ... ,I I:.- I:. - . , ,.. t1l S . ... ... . . .

,6- 1. - - . . . - - . . . .

. b. With le'ss' i

• d is Of ~ab- 'rbeUi ; Uit . 0.C-d' t 'b ;emsdeneroze di sOpend CORE 'ALTERATIONS-i,ardling of distribiiti:o ysttem ople'd'iev sns' 1,n-t kenrI.. rast 'the Ssnr ibeq .r6Q' .

pyte e**,,not s  ;!b'*-'.""rj~tiACTION

' !. 'i'a

.t . -for that SUThR EpILCOviontRO R Vs.M ofE:Specificat on 3.0.3 t.4'p I:ale .

SURV ILLNC REQUIREMENTS.-'-.:. ,..- --

¢' cs.s'rt'i.'*;9 ',.

4.8 ' At le',ast th , i striti on:

4..3ab1~t hlove required' powoersysteon m -divsi on shall be determined energizedddays atleaston~e per 7 by verinfyingtorrec breaker alignment and voltige on. the .busses/ItCCs/panels. ..

LIUERICK- UNIT 2 3/4 8-20 ABE 2 5 0

Unit 2 Limerick Generating Station Technical Specification Insert Insert 1 (Page 1-2):

Table 2.1 of Federal Guidelines Report 11, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion," ORNL, 1989, as described in Regulatory Guide 1.183. The factors in the column headed "effective" yield doses corresponding to the CEDE.

Insert 2 (Page 1-6):

RECENTLY IRRADIATED FUEL 1.35 RECENTLY IRRADIATED FUEL is fuel that has occupied part of a critical reactor core within the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Insert 3 (Page 314 1-19) 3.1.5 The standby liquid control system shall be OPERABLE and consist of a minimum of the following:

a. In OPERATIONAL CONDITIONS 1 and 2, two pumps and corresponding flow paths,

b. In OPERATIONAL CONDITION 3, one pump and corresponding flow path.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3 ACTION:

a. With only one pump and corresponding explosive valve OPERABLE, in OPERATIONAL CONDITION 1 or 2, restore one inoperable pump and corresponding explosive valve to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

b. With standby liquid control system otherwise inoperable, in OPERATIONAL CONDITION 1, 2, or 3, restore the system to OPERABLE status within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Page 1 of 2

Insert 4 (Pages 3/4 6-47, 3/4 6-50, 3/4 6-52,)

When RECENTLY IRRADIATED FUEL is being handled in the secondary containment, or during operations with a potential for draining the reactor vessel, with the vessel head removed and fuel in the vessel Insert 5 (Pages 3/4 7-3. 3/4 7-4, 3/4 7-5, 3/4 7-6, 3/4 8-9, 3/4 8-14, 3/4 8-19)

When RECENTLY IRRADIATED FUEL is being handled in the secondary containment, or during operations with a potential for draining the reactor vessel Page 2 of 2

ATTACHMENT 3 LIMERICK GENERATING STATION UNITS 1 AND 2 Docket Nos. 50-352 & 50-353 License Nos. NPF-39 & NPF-85 License Amendment Request "LGS Alternative Source Term Implementation" Markup of Technical Specification Bases Pages (For information only)

UNITS 1 B 3/4 1-4 B 3/4 1-5 B 3/4 4-6 B 3/4 6-5 B 3/4 7-1 a UNITS 2 B 3/4 1-4 B 3/4 1-5 B 3/4 4-6 B 3/4 6-5 B 3/4 7-1

REACTIVITY CONTROL SYSTEMS BASES 3/4.1.5 STANDBY LIQUID CONTROL SYSTEM The standby liquid control system provides a backup capability for bringing the reactor from full power to a cold, Xenon-free shutdown, assuming that the withdrawn control rods remain fixed in the rated power pattern. To meet this objective it is necessary to inject a quantity of boron-which produces a concen-ration of 660 ppm in the reactor core and other piping systems connected to r the reactor vessel. To allow for potential leakage and improper mixing, this concentration is increased by 25%. The required concentration is achieved by having available a minimum quantity of 3,160 gallons of sodium pentaborate solution g inng a minimum of 3,754 lbs of sodium pentaborate having the r- equisite atom % enrichment of 29% as determined from Reference 5.

.;.'f1 his quantity of solution is a net amount which is above the pump suction

'1 i'hutoff level setpoint thus allowing for the portion which cannot be injected.

.iThe pumping rate of 41.2 gpm provides a negative reactivity insertion rate over

_ the permissible solution volume range, which adequately compensates for the positive reactivity effects due to elimination of steam voids, increased water density from hot to cold, reduced doppler effect in uranium, reduced neutron leakage from boiling to cold, decreased control rod worth as the moderator cools, and xenon decay. The temperature requirement ensures that the sodium pentaborate always remains in solution.

With redundant pumps and explosive injection valves and with a highly reliable control rod scram system, operation of the reactor is permitted to continue for short periods of time with the system inoperable or for longer periods of time with one of the redundant components inoperable.

The SLCS system consists of three separate and independent pumps and explosive valves. Two of the separate and independent pumps and explosive valves are required to meet the minimum requirements of this technical specification and, where applicable, satisfy the single failure criterion.

The SLCS must have an equivalent control capacity of 86 gpm of 13% weight sodium pentaborate in order to satisfy 10 CFR 50.62 (Requirements for reduction of risk from anticipated transients without scram (ATWS) events for light-water-cooled nuclear power plants). As part of the ARTS/MELLL program the ATWS analysis was updated to reflect the new rod line. As a result'of this it was determined that the Boron 10 enrichment was required to be increased to 29% to prevent exceeding a suppression pool temperature of 190@F. This equivalency requirement is fulfilled by having a system which satisfies the equation given in 4.1.5.b.2.

The upper limit concentration of 13.8% has been established as a reasonable limit to prevent precipitation of sodium pentaborate in the event of a loss of tank heating, which allow the solution to cool.

LIMERICK - UNIT 1 B 3/4 1-4 Amendment No. 72,.6 6 FEB 10 1994

REACTIVITY CONTROL SYSTEMS BASES STANDBY LIQUID CONTROL SYSTEM (Continued)

Surveillance requirements are established on a frequency that assures a high reliability of the system. Once the solution is established, boron con-centration will not vary unless more boron or water is added, thus a check on the temperature and volume once each 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> assures that the solution is available for use.

Replacement of the explosive charges in the valves at regular intervals will assure that these valves will not fail because of deterioration of the charges.

1. C. J. Paone, R. C. Stirn and J. A. Woolley, "Rod Drop Accident Analysis for Large BWR's," G. E. Topical Report NEDO-10527, March 1972.

2. C. J. Paone, R. C. Stirn, and R. M. Young, Supplement 1 to NEDO-10527, July 1972.

3. J. M. Haun, C. J. Paone, and R. C. Stirn, Addendum 2, 'Exposed Cores."

Supplement 2 to NEDO-10527, January 1973.

4. Amendment 17 to General Electric Licensing Topical Report NEDE-24011-P-A, "General Electric Standard Application for Reactor Fuel'.

5. "Maximum Extended Load Line Limit and ARTS Improvement Program Analyses for Limerick Generating Station Units 1 and 2," NEDC-32193P, Revision 2, October 1993. I LIMERICK - UNIT 1 B 3/4 1-5 Amendment No. Zi, 66 FEB 10 1994

REACTOR COOLANT SYSTEM BASES 3/4.4.7 MAIN STEAM LINE ISOLATION VALVES Double isolation valves are provided on each of the main steam lines to minimize the potential leakage paths from the containment in case of a line break.

Only one valve in each line is required to maintain the integrity of the containment, however, single failure considerations require that two valves be OPERABLE. The surveillance requirements are based on the operating history of this type valve. The maximum closure time has been selected to contain fission products and to following line breaks. The minimum closure flimeiis cons in the safety analyses to prevent pressure surges.

3/4.4.8 STRUCTURAL INTEGRITY The inspection programs for ASME Code Class 1, 2, and 3 components ensure that the structural integrity of these components will be maintained at an acceptable level throughout the life of the plant.

Components of the reactor coolant system were designed to provide access to permit inservice inspections in accordance with Section XI of the ASME Boiler and Pressure Vessel Code 1971 Edition and Addenda through Winter 1972.

The inservice inspection program for ASME'Code Class 1, 2, and.3 components will be performedin accordance with:Section XI of the ASME Boiler and-Pressure Vessel'Code'and.applicable addenda as required by 10 CFR 50.55a. Additionally, the Inservice Inspection Program conforms to.-the NRC staff positions identified in NRC Generic Letter'88-01, "NRC Position on .IGSCC in BWR Austenitic Stainless Steel Piping,"

as approved in NRC Safety Evaluations dated March 6, 1990 and-October 22, 1990.

3/4.4.9 RESIDUAL HEAT REMOVAL The'RHR system is required to remove decay heat and sensible heat in order to maintain the temperature of the reactor coolant. RHR shutdown cooling is comprised of four (4)subsystems which make two (2) loops. Each loop consists of two (2) motor driven pumps, a heat exchanger, and associated piping and valves. Both lops have a common suction' from the same recirculation loop. Two (2) redundant, manually controlled shutdown'cooling subsystems of the HR System can provide the required decay heat removal capability. Each pump discharges the reactor coolant, after it has been cooled by circulation through the respective heat exchangers, to the reactor via the associated recirculation loop or to the reactor via the low pressure coolant injection pathway. The RHR heat exchangers transfer heat to the RHR Service Water System. The RHR shutdown cooling mode is manually controlled.

An OPERABLE RHR shutdown cooling subsystem consists of an RHR pump, a heat exchanger, valves, piping, instruments, and controls to ensure an OPERABLE flow path.

In HOT SHUTDOWN condition, the requirement to maintain OPERABLE two (2) independent RHR shutdown cooling subsystems means that each subsystem considered OPERABLE must be associated with a different heat exhanger loop, i.e., the "A" RHR heat exchanger with the 'Al RHR pump or the "C" RHR pump, and the "B" RHR heat exchanger with the "B" RHR' pump or the "DO RHR pump are two (2) independent RHR shutdown cooing sub systems. Only one (1) of the two (2) RHR pumps associated with each RHR heat exchanger loop is APR 2 8 1998

-.. IA RI'l - UNIT 1 B 3/4 4-6 -Amendmp~nt No. 4M,97,119',125

COWTAIWFNT SYSTEMS WAES 314J .5 SECONEARY CONAIM T Secondary containment is designed to minimize any ground level release of radioactive material which may result from an accident. The Reactor Enclosure and associated structures provide secondary containment during normal operation when the drywell is sealed and in service. At other times the drywell may be open and, when required, secondary containment integrity is specified.

Establishing and maintaining a vaccuum in the reactor enclosure secondary containment with the standby gas treatment system once per 24 months, along with the surveillance of the doors, hatches, dampers and valves, is adequate to ensure that there are no violations of the integrity of the secondary containment-The OPERABILITY of the reactor enclosure recirculation system and the standby gas treatment systems ensures that sufficient iodine removal capability will be available in the event of a LOCA or refueling accident(SGTS only). The reduction in containment iodine inventory reduces the resulting SITE BOUNDA _

radiation doses associated with containment leakage. The operation of this e3 system and resultant iodine removal capacity are consistent with the assumptions used in the LOCA and refueling accident analyses. Provisions have been made to continuously purge the filter plenums with instrument air when the filters are not in use to prevent buildup of moisture on the adsorbers and the HEPA filters.

Although the safety analyses assumes that the reactor enclosure secondary containment draw down time will take 930 seconds, these surveillance require-ments specify a draw down time of 916 seconds. This 14 second difference is due to the diesel generator starting and sequence loadiing delays which is not part of this surveillance requirement.

The reactor enclosure secondary containment draw down time analyses assumes a starting point of 0.zs inch of vacuum water gauge and worst case SGTS dirty filter flow rate of 28UO cft. The surveillance requirements satisfy this as-sumption by starting the drawdown from ambient conditions and connecting the adjacent reactor enclosure and refueling area to the SGTS to split the exhaust flow between the three Zones and verifying a minimum flow rate of 2800 cfM from the test zone- This simulates the worst case flow alignment and verifies ade-quate flaw is available to drawdown the test zone within the required time.

The Technical Specification Surveillance Requirement 4.6.5.3.b.3 is intended abe a multi-zone air balance verification without isolating any test zone.

The STTS fans are sized for three zones and therefore, when aligned to a single zone or two zones, will have excess capacity to more quickly drawdown the affected Zones_ There is no maximum flow limit to individual zones or pairs of zones and the air balance and drawdown time are verified when.all three zones are connected to the SGTS.

The three zone air balance verification and drawdown test will be done after. any major system alteration, which is any modification which will have an effect on the SGTS flowrate such that the ability of the SGTS to drawdown the reactor enclosure to greater than or equal to 0.25 inch of vacuum water gage in less than or equal to 916 seconds could be affected.

LIMERICK - UNIT 1 B 3/4 6-5 krsndat No. 9, 4d, 1Y, 10'. L22 FEB 18 1991

3901038720 PLANT SYSTEMS BASES 3/4.7.2 CONTROL ROOM EMERGENCY FRESH AIR SUPPLY SYSTEM - COMMON SYSTEM The OPERABILITY of the control room emergency fresh air supply system ensures that the control room will remain habitable for operations personnel gr t\

during and following all design basis accident conditions. Constant purge of l the system at 1 cfm is sufficient to reduce the buildup of moisture on the adsorbers and'HEPA filters. The OPERABILITY of this system in conjunction with D control room design provisions is based on limiting the radiation exposure to prsonnel occupying the control room to 5 rem or less limitation is consistent with the iThs of Q ~~~~~~10 CFR Part W_,, ,aAA 3/4.7.3 REACTOR CORE ISOLATION COOLING SYSTEM The reactor core isolation cooling (RCIC) system is provided to assure adequate core cooling in the event of reactor isolation from its primary heat sink and the loss of feedwater flow to the reactor vessel without requiring 't actuation of any of the emergency core cooling system equipment. The RCIC system is conservatively required to be OPERABLE whenever reactor pressure ex--

ceeds .150 psig. This pressure is substantially below that for which low pressure core cooling systems can provide adequate core cooling.

The RCIC system specifications are applicable during OPERATIONAL CONDITIONS 1, 2, and 3 when reactor vessel pressure exceeds 150 psig because RCIC is the primary non-ECCS source of emergency core cooling when the reactor is pressurized.

With the RCIC system inoperable, adequate core cooling is assured by the OPERABILITY of the HPCI'system and justifies the specified 14 day out-of-service period.

The surveillance'requirements provide adequate assurance that RCIC will be OPERABLE when required. Although all active components are testable and full flow can be demonstrated by recirculation during reactor operation, a complete functional test requires reactor shutdown. The pump discharge piping is maintained full to prevent water hammer damage and to start cooling at the earliest possible moment.

LIMERICK - UNIT 1 B 3/4 7-la Amendient- No.:27, 40 MAY 30 1990

Unit 1 Limerick AST LAR Bases Inserts INSERT 1 (Page B3/4 1-4)

The above quantities calculated at 29% Boron-1 0 enrichment have been demonstrated by analysis to provide a Boron-1 0 weight equivalent of 185 lbs in the sodium pentaborate solution. Maintaining this Boron-10 weight in the net tank contents ensures a sufficient quantity of boron to bring the reactor to a cold, Xenon-free shutdown.

INSERT 2 (Page B3/4 1-5)

The Standby Liquid Control System also has a post-DBA LOCA safety function to buffer Suppression Pool pH in order to maintain bulk pH above 7.0. The buffering of Suppression Pool pH is necessary to prevent iodine re-evolution to satisfy the methodology for Alternative Source Term. Manual initiation is used, and the minimum amount of total boron required for Suppression Pool pH buffering is 240 lbs. Given that at least 185 lbs of Boron-10 is maintained in the tank, the total boron in the tank will be greater than 240 lbs for the range of enrichments from 29% to 62%.

ACTION Statement (a) applies only to OPERATIONAL CONDITIONS 1 and 2 because a single pump can satisfy both the reactor control function and the post-DBA LOCA function to control Suppression Pool pH since boron injection is not required until 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br /> post-LOCA. ACTION Statement (b) applies to OPERATIONAL CONDITIONS 1, 2 and 3 to address the post-LOCA safety function of the SLC system.

INSERT 3 (Page B3/4 6-5)

Based on implementation of the AST methodology, it has been demonstrated through analysis that operating the subsystem with a minimum flow rate of 30,000 cfm satisfies the dose requirements. However, based on the applicable testing standards and guidance provided in Regulatory Guide 1.52, Revision 2, and ANSI N510, 1980, the acceptance criteria must be satisfied based on the system rated flow, which provides the most conservative results. Satisfying the acceptance criteria at rated flow also demonstrates system operability at lower measured flows because the residence time is longer. Therefore, verification of subsystem flow rates between 30,000 cfm and the maximum rated flow of 66,000 cfm (includes a 10% factor) satisfies the surveillance requirements for the HEPA filters and charcoal adsorber housings based on 70%

efficiencies.

INSERT 4 (Page B3/4 7-1 a)

Since the Control Room Emergency Fresh Air Supply System is not credited for filtration in OPERATIONAL CONDITIONS 4 and 5, applicability to 4 and 5 is only required to support the Chlorine and Toxic Gas design basis isolation requirements.

Additionally, based on implementation of the Alternative Source Term methodology, it has been demonstrated through analysis that manual initiation of the CREFAS radiation mode within 30 minutes of the start of gap release for the limiting design basis LOCA is sufficient to assure that control room operator dose limits in 10 CFR Part 50.67 are met.

.REACTIVITY CONTROL SYSTEMS BASES 3/4.1.5 STANDBY LIQUID CONTROL SYSTEM The standby liquid control system provides a backup capability for bringing the reactor from full power to a cold, Xenon-free shutdown, assuming that the withdrawn control rods remain fixed in the rated power pattern. To meet this objective it is necessary to inject a quantity of boron which produces a concen-tration of 660 ppm in the reactor core and other piping systems connected to the.

reactor vessel. To allow for potential leakage and improper mixing, this con-centration is increased by 25%. The required concentration is achieved by having available a minimum quantity of 3,160 gallons of sodium pentaborate solution coitzannn a minimum of 3,754 lbs of sodium pentaborate having the requis tom % enrichment of 29% as determined from Reference 5.

This quantity of solution is a net amount which is above the pump suction shutoff level setpoint thus allowing for the portion which cannot be injected.

The pumping rate of 41.2 gpm provides a negative reactivity insertion rate over the permissible solution volume range, which adequately compensates for the positive reactivity effects due to elimination of steam voids, increased water density from hot to cold, reduced doppler effect in uranium, reduced neutron leakage from boiling to cold, decreased control rod worth as the moderator cools, and xenon decay. The temperature requirement ensures that the sodium pentaborate always remains in solution.

With redundant pumps and explosive injection valves and with a highly reliable control rod scram system, operation of the reactor is permitted to continue for short periods of time with the system inoperable or for longer periods of time with one of the redundant components inoperable.

The SLCS system consists of three separate and independent pumps and explosive valves. Two of the separate and independent pumps and explosive valves are required to meet the minimum requirements of this technical specification and, where applicable, satisfy the single failure criterion.

The SLCS must have an equivalent control capacity of 86 gpm of 13% weight sodium pentaborate in order to satisfy 10 CFR 50.62 (Requirements for reduction of risk from anticipated transients without scram (ATWS) events for light-water-cooled nuclear power plants). As part of the ARTS/MELLL program the ATWS analysis was updated to reflect the new rod line. As a result of this it was determined that the Boron 10 enrichment was required to be increased to 29% to prevent exceeding a suppression pool temperature of 190 0F. This equivalency requirement is fulfilled by having a system which satisfies the equation given in 4.1.5.b.2.

The upper limit concentration of 13.8% has been established as a reasonable limit to prevent precipitation of sodium pentaborate in the event of a loss of tank heating, which allow the solution to cool.

LIMERICK - UNIT 2 B 3/4 1-4 Amendment No. 48 JAN 3i 1995

REACTIVITY CONTROL SYSTEMS BASES STANDBY LIQUID CONTROL SYSTEM (Continued)

Surveillance requirements are established on a frequency that assures a high reliability of the system. Once the solution is established, boron concentration will not vary unless more boron or water is added, thus'a check on the temperature and volume once each 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> assures that the solution is available for use.

Replacement of the explosive charges in the valves at regular intervals will assure that these valves will not fail because of deterioration of the charges.

1. C. J. Paone, R. C. Stirn and J. A. Woolley, "Rod Drop Accident Analysis for Large BWR's," G. E. Topical Report NEDO-10527, March 1972.

2. C. J. Paone, R. C. Stirn, and R. M. Young, Supplement I to NEDO-10527, July 1972.

3. J. M. Haun, C. J. Paone, and R. C. Stirn, Addendum 2, "Exposed Cores,"

Supplement 2 to NEDO-10527, January 1973.

4. Amendment 17 to General Electric Licensing Topical Report NEDE-24011-P-A, "General Electric Standard Application for Reactor Fuel".

5. "Maximum Extended Load .Line..Limit and-ARTS Improvement PrDgram Analyses for Limerick Generating Station Units I and 2," NEDC-32193P, Revision 2, October 1993.

LIMERICK - UNIT 2 B 3/4 1-5 AMiendnmiet No. 48 JAN 3 1 1995

REACTOR COOLANT SYSTEM BASES 3/4.4.7 MAIN STEAM LINE ISOLATION VALVES Double isolation valves are provided on each of the main steam lines to minimize the potential leakage paths from the containment in case of a line break.

Only one valve in each line is required to maintain the integrity of the containment however, single failure considerations require that two valves be OPERABLE. the surveillance requirements are based on the operating history of this type valve. xi 1su i ha been selected to contain fission products and to following line breaks. The minimum closure ime is consistent with te asumptions in the safety analyses to prevent pressure surges.

3/4.4.8 STRUCTURAL INTEGRITY The inspection programs for ASME Code Class 1, 2, and 3 components ensure that the structural integrity of these components will be maintained at an acceptable level throughout the life of the plant.

Components of the reactor coolant system were desi ned to provide access to permit inservice inspections in accordance with Section RI of the ASME Boiler and Pressure Vessel Code 1971 Edition and Addenda through Winter 1972.

The inservice inspection program for ASME Code Class 1, 2, and 3 components will be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable addenda as required by 10 CFR 50.55a. Additionally, the Inservice Inspection Pro ram conforms to the NRC staff positions identified in NRC Generic Letter 88-01, "NRC Position on IGSCC in BWR Austenitic Stainless Steel Piping,"

as approved in NRC Safety Evaluations dated March 6, 1990 and October 22, 1990.

3/4.4.9 RESIDUAL HEAT REMOVAL The RHR system is required to remove decay heat and sensible heat in order to maintain the temperature of the reactor coolant. RHR shutdown cooling is comprised of four (4) subsystems which make two (2) loops. Each loop consists of two (2) motor driven pumps, a heat exchanger, and associated piping and valves. Both loops have a common suction from the same recirculation loop. Two (2) redundant, manually controlled shutdown cooling subsystems of the RHR System can provide the required decay heat removal capability; Each pump discharges the reactor coolant, after it has been cooled by circulation through the respective heat exchangers, to the reactor via the associated recirculation loop or to the reactor via the low pressure coolant injection pathway. The RHR heat exchangers transfer heat to the RHR Service Water System. The RHR shutdown cooling mode is manually controlled.

An OPERABLE RHR shutdown cooling subsystem consists of an RHR pum , a heat exchanger, valves, piping, instruments, and controls to ensure an OPERABLE flow path.

In HOT SHUTDOWN condition, the requirement to maintain OPERABLE two 2 independent RHR shutdown cooling subsystems means that each subsystem,considered OPERABLE must be associated with a different heat exhanger loop, i.e., the "A" RHR heat exchanger with the "A" RHR pump or the "C' RHR pump, and the "B" RHR heat exchanger with the "B" RHR pump or the 'D"RHR pump are two (2) inidpendent RHR shutdown cooling subsystems. Only one (1) of the two_(2) RHRpumps associated with each RHR_heat exchanger loop is APR 2 8 1998 LIMERICK - UNIT 2 B 3/4 4-6 Ar~ndment No. 1Z,F1,Fi,89

• . (CTAIMENSYStEM 3J4.6.S SECONDARY COWTAINMENT Secondary containment is designed to minimize any ground level release of radioactive material which may result from an accident. The Reactor Enclosure and associated structures provide secondary containment during normal operation when the drywell is sealed and in service. At other times the drywell may be open and, when required. secondary containment integrity is specified.

Establishing and maintaining a vacuuM in the reactor enclosure secondary containment with the standby gas treatment system once per 24 months, along with the surveillance of the doors. hatches, dampers and valves, is adequate to ensure that there are no violations of the integrity of the secondary containment.

The OPERAILITY of the reactor enclosure recirculation system and the standby gas treatment systems ensures that sufficient iodine removal capability will be available in the event of a LOCA or refueling accldentv(SGTS only). The I.VJ/,n reduction in containment iodine inventory reduces the resulting SITE BOUNDY radiation doses associated with containment leakage. The operation of this ^tENre4 1IRAI system and resultant iodine removal capacity are consistent with the assumption TEEj) used it the LOCA and refueling accident analyses. Provisions have been made to continuously purge the filter plenums with instrument air when the filters are not in use to prevent buildup of moisture on the adsorbers and the HEPA filters.

Although the safety analyses assumes that the reactor enclosure secondary containment draw down time will take 930 seconds, these surveillance require-ments specify a draw dowin time of 916 seconds. This 14 second difference is due to the diesel generator starting and sequence loading delays which is not part of this surveillance requirement.

The reactor enclosure secondary containment draw down time analyses assumes a starting point of 0.Z5 inch of vacuum water gauge and worst case SGTS dirty filter flow rate of 2800 cft. The surveillance requirements satisfy this as-sumption by starting the drawdown from ambient conditions and connecting the adjacent reactor enclosure and refueling area to the SGTS to split the exhaust flow between the three zones and verifying a minimum flow rate of 2800 cfa from the test zone. This simulates the worst case flow alignment and verifies ade-quate flow is available to drawdown the test zone within the required time.

The Technical Specification Surveillance Requirement 4.6.S.3.b.3 is intended to be a multi-zone air balance verification without isolating any test zone.

The SETS is comnon to Unit I and 2 and consists of two independent subsystems. The power supplies for the corn portions of the subsystems are from Unit I safeguard busses, therefore the inoperabIlity of these Unit 1 supplies are addressed in the SGTS ACTION statements in order to ensure adequate onsite powersources to SGTS for its Unit 2 function during a loss of offsite power event. The allowable out of service times are consistent with those in the Unit 1 Technical Specifications for SGTS and At electrical power supply out of service condition combinations.

LIMERICK - UNIT 2 B 3/4 6-S Am fa N 34, 5r, 86

,E I 8 1997i

3/4.7 PLANT SYSTEMS BASES 3/4.7.1 SERVICE WATER SYSTEMS - COMMON SYSTEMS The OPERABILITY of the service water systems ensures that sufficient cooling capacity is available for continued operation of safety-related equipment during normal and accident conditions. The redundant cooling capacity of these systems, assuming a single failure, is consistent with the assumptions used in the accident conditions within acceptable limits.

The RHRSW and ESW systems are common to Units 1 and 2 and consist of two independent subsystems each with two pumps. One pump per subsystem (loop) is powered from a Unit 1 safeguard bus and the other pump is powered from a Unit 2 safeguard bus. In order to ensure adequate onsite power sources to the systems during a loss of offsite power event, the inoperability of these supplies are restricted in system ACTION statements.

RHRSW is a manually operated system used for core and containment heat removal. Each of two RHRSW subsystems has one heat exchanger per unit. Each RHRSW pump provides adequate cooling for one RHR heat exchanger. By limiting operation with less than three OPERABLE RHRSW pumps with OPERABLE Diesel Generators, each unit is ensured adequate heat removal capability for the design scenario of LOCA/LOOP on one unit and simultaneous safe shutdown of the other unit.

Each ESW 'pump provides adequate flow to the cooling loads in its associated loop. With only two divisions of power required for LOCA mitigation of one unit and one division of power required for safe shutdown of the other unit, one ESW pump provides sufficient capacity to fulfill design requirements. ESW pumps are automatically started upon start of the associated Diesel Generators.

Therefore, the allowable out of service times for OPERABLE ESW pumps and their associated Diesel Generators is limited to ensure adequate cooling during a loss of offsite power event.

3/4.7.2 CONTROL ROOM EMERGENCY FRESH AIR SUPPLY SYSTEM - COMMON SYSTEM The OPERABILITY of the control room emergency fresh air supply system ensures that the control room will remain habitable for operations personnel during and following all design basis accident conditions. Constant purge of the system at 1 cfm is sufficient to reduce the buildup of moisture on the adsorbers and HEPA filters. The OPERABILITY of this system in conjunction with control room design provisions is based on limiting the radiation exposure to Personnel occupying the control room to 5 rem or less Tis consistent with the requirements o e 10 CFR Part"?-'

\5. 67 ALo deA J ob.i A The CREFAS is common to Units 1 and 2 an o two in subsystems. The power supplies for the system are from Unit 1 Safeguard busses, therefore, the inoperability of these Unit 1 supplies are addressed in the CREEAS_ACTION statements .. in order. to ensure adequate -ons-ite power -sources to CREFAS during a loss of offsite power event. The allowable out of service LIMERICK - UNIT 2 B 3/4 7-1 AUG 2 5 1989

Unit 2 Limerick AST LAR Bases Inserts INSERT 1 (Page B3/4 1-4)

The above quantities calculated at 29% Boron-10 enrichment have been demonstrated by analysis to provide a Boron-10 weight equivalent of 185 Ibs in the sodium pentaborate solution. Maintaining this Boron-10 weight in the net tank contents ensures a sufficient quantity of boron to bring the reactor to a cold, Xenon-free shutdown.

INSERT 2 (Page B3/4 1-5)

The Standby Liquid Control System also has a post-DBA LOCA safety function to buffer Suppression Pool pH in order to maintain bulk pH above 7.0. The buffering of Suppression Pool pH is necessary to prevent iodine re-evolution to satisfy the methodology for Alternative Source Term. Manual initiation is used, and the minimum amount of total boron required for Suppression Pool pH buffering is 240 Ibs. Given that at least 185 Ibs of Boron-10 is maintained in the tank, the total boron in the tank will be greater than 240 Ibs for the range of enrichments from 29% to 62%.

ACTION Statement (a) applies only to OPERATIONAL CONDITIONS 1 and 2 because a single pump can satisfy both the reactor control function and the post-DBA LOCA function to control Suppression Pool pH since boron injection is not required until 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br /> post-LOCA. ACTION Statement (b) applies to OPERATIONAL CONDITIONS 1, 2 and 3 to address the post-LOCA safety function of the SLC system.

INSERT 3 (Page B3/4 6-5)

Based on implementation of the AST methodology, it has been demonstrated through analysis that operating the subsystem with a minimum flow rate of 30,000 cfm satisfies the dose requirements. However, based on the applicable testing standards and guidance provided in Regulatory Guide 1.52, Revision 2, and ANSI N510, 1980, the acceptance criteria must be satisfied based on the system rated flow, which provides the most conservative results. Satisfying the acceptance criteria at rated flow also demonstrates system operability at lower measured flows because the residence time is longer. Therefore, verification of subsystem flow rates between 30,000 cfm and the maximum rated flow of 66,000 cfm (includes a 10% factor) satisfies the surveillance requirements for the HEPA filters and charcoal adsorber housings based on 70%

efficiencies.

INSERT 4 (Paae B3/4 7-1)

Since the Control Room Emergency Fresh Air Supply System is not credited for filtration in OPERATIONAL CONDITIONS 4 and 5, applicability to 4 and 5 is only required to support the Chlorine and Toxic Gas design basis isolation requirements.

Additionally, based on implementation of the Alternative Source Term methodology, it has been demonstrated through analysis that manual initiation of the CREFAS radiation mode within 30 minutes of the start of gap release for the limiting design basis LOCA is sufficient to assure that control room operator dose limits in 10 CFR Part 50.67 are met.

ATTACHMENT 4 LIMERICK GENERATING STATION UNITS 1AND 2 Docket Nos. 50-352 & 50-353 License Nos. NPF-39 & NPF-85 License Amendment Request

'LGS Alternative Source Term Implementation' Retyped Technical Specification Pages UNIT 1 UNIT 2 1-2 1-2 1-6 1-6 1-7 1-7 3/4 1-19 3/4 1-19 3/4 1-20 3/4 1-20 3/4 3-16 3/4 3-16 3/4 3-31 3/4 3-31 3/4 3-64 3/4 3-64 3/4 3-65 3/4 3-65 3/4 3-66 3/4 3-66 3/4 3-67 3/4 3-67 3/4 4-23 3/4 4-23 3/4 6-3 3/4 6-3 3/4 6-47 3/4 6-47 3/4 6-50 3/4 6-50 3/4 6-52 3/4 6-52 3/4 6-53 3/4 6-53 3/4 6-55 3/4 6-55 3/4 6-56 3/4 6-56 3/4 7-3 3/4 7-3 3/4 7-4 3/4 7-4 3/4 7-5 3/4 7-5 3/4 7-6 3/4 7-6 3/4 7-7 3/4 7-7 3/4 8-9 3/4 8-9 3/4 8-14 3/4 8-14 3/4 8-14A 3/4 8-14A 3/4 8-20 3/4 8-19 3/4 8-20

DEFINITIONS CORE ALTERATION 1.7 CORE ALTERATION shall be the movement of any fuel, sources, or reactivity control components within the reactor vessel with the vessel head removed and fuel in the vessel. The following exceptions are not considered to be CORE ALTERATIONS:

a) Movement of source range monitors, local power range monitors, intermediate range monitors, traversing incore probes, or special moveable detectors (including undervessel replacement); and b) Control rod movement, provided there are no fuel assemblies in the associated core cell.

Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe position.

CORE OPERATING LIMITS REPORT 1.7a The CORE OPERATING LIMITS REPORT (COLR) is the unit-specific document that provides the core operating limits for the current operating reload cycle. These cycle-specific core operating limits shall be determined for each reload cycle in accordance with Specifications 6.9.1.9 thru 6.9.1.12. Plant operation within these limits is addressed in individual specifications.

CRITICAL POWER RATIO 1.8 The CRITICAL POWER RATIO (CPR) shall be the ratio of that power in the assembly which is calculated by application of the (GEXL) correlation to cause some point in the-assembly to experience boiling transition, divided by the actual assembly operating power.

DOSE EQUIVALENT 1-131 1.9 DOSE EQUIVALENT I-131 shall be that concentration of I-131, microcuries per gram, which alone would produce the same inhalation committed effective dose equivalent (CEDE) as the quantity and isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present. The inhalation committed effective dose equivalent (CEDE) conversion factors used for this calculation shall be those listed in Table 2.1 of Federal Guidelines Report 11, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion," ORNL, 1989, as described in Regulatory Guide 1.183. The factors in the column headed "effective" yield doses corresponding to the CEDE.

DOWNSCALE TRIP SETPOINT (DTSP) 1.9a The downscale trip setpoint associated with the Rod Block Monitor (RBM) rod block trip setting.

E-AVERAGE DISINTEGRATION ENERGY 1.10 E shall be the average, weighted in proportion to the concentration of each radionuclide in the reactor coolant at the time of sampling, of the sum of the average beta and gamma energies per disintegration, in MeV, for isotopes, with half lives greater than 15 minutes, making up at least 95% of the total noniodine activity in the coolant.

EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME 1.11 The EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ECCS actuation set-point at the channel sensor until the ECCS equipment is capable of performing its safety function, i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc. Times shall include diesel generator starting and sequence loading delays where applicable. The response time may be measured by any series of sequential, overlapping or total steps such that the entire response time is measured.

LIMERICK - UNIT 1 1-2 Amendment No. 3X, 6&, 8X,

DEFINITIONS PURGE - PURGING 1.31 PURGE or PURGING shall be the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is required to purify the confinement.

RATED THERMAL POWER 1.32 RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 3458 MWt.

REACTOR ENCLOSURE SECONDARY CONTAINMENT INTEGRITY 1.33 REACTOR ENCLOSURE SECONDARY CONTAINMENT INTEGRITY shall exist when:

a. All reactor enclosure secondary containment penetrations required to be closed during accident conditions are either:

1. Capable of being closed by an OPERABLE secondary containment automatic isolation system, or

2. Closed by at least one manual valve, blind flange, slide gate damper, or deactivated automatic valve secured in its closed position, except as provided by Specification 3.6.5.2.1.

b. All reactor enclosure secondary containment hatches and blowout panels are closed and sealed.

c. The standby gas treatment system is in compliance with the requirements of Specification 3.6.5.3.

d. The reactor enclosure recirculation system is in compliance with the requirements of Specification 3.6.5.4.

e. At least one door in each access to the reactor enclosure secondary containment is closed.

f. The sealing mechanism associated with each reactor enclosure secondary containment penetration, e.g., welds, bellows, or O-rings, is OPERABLE.

9. The pressure within the reactor enclosure secondary containment is less than or equal to the value required by Specification 4.6.5.1.1a.

REACTOR PROTECTION SYSTEM RESPONSE TIME 1.34 REACTOR PROTECTION SYSTEM RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its trip setpoint at the channel sensor until de-energization of the scram pilot valve solenoids. The response time may be measured by any series of sequential, overlapping or total steps such that the entire response time is measured.

RECENTLY IRRADIATED FUEL 1.35 RECENTLY IRRADIATED FUEL is fuel that, has occupied part of a critical reactor core within the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

REFUELING FLOOR SECONDARY CONTAINMENT INTEGRITY 1.36 REFUELING FLOOR SECONDARY CONTAINMENT INTEGRITY shall exist when:

a. All refueling floor secondary containment penetrations required to be closed during accident conditions are either:

LIMERICK - UNIT I 1-6 Amendment No. 33, 6X, 1-0-5, 4Ad6,

DEFINITIONS REFUELING FLOOR SECONDARY CONTAINMENT INTEGRITY (Continued)

1. Capable of being closed by an OPERABLE secondary containment automatic isolation system, or

2. Closed by at least one manual valve, blind flange, slide gate damper, or deactivated automatic valve secured in its closed position, except as provided by Specification 3.6.5.2.2.

b. All refueling floor secondary containment hatches and blowout panels are closed and sealed.

c. The standby gas treatment system is in compliance with the requirements of specification 3.6.5.3.

d. At least one door in each access to the refueling floor secondary containment is closed.

e. The sealing mechanism associated with each refueling floor secondary containment penetration, e.g., welds, bellows, or O-rings, is OPERABLE.

f. The pressure within the refueling floor secondary containment is less than or equal to the value required by Specification 4.6.5.1.2a.

REPORTABLE EVENT 1.37 A REPORTABLE EVENT shall be any of those conditions specified in Section 50.73 to 10 CFR Part 50.

ROD DENSITY 1.38 ROD DENSITY shall be the number of control rod notches inserted as a fraction of the total number of control rod notches. All rods fully inserted is equivalent to 100% ROD DENSITY.

SHUTDOWN MARGIN 1.39 SHUTDOWN MARGIN shall be the amount of reactivity by which the reactor is subcritical or would be subcritical assuming all control rods are fully inserted except for the single control rod of highest reactivity worth which is assumed to be fully withdrawn and the reactor is in the shutdown condition; cold, i.e. 68'F; and xenon free.

SITE BOUNDARY 1.40 The SITE BOUNDARY shall be that line as defined in Figure 5.1.3-la.

SOURCE CHECK 1.41 A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a radioactive source.

LIMERICK - UNIT 1 1-7 Amendment No. 48, 66, 4105,

REACTIVITY CONTROL SYSTEMS 3/4.1.5 STANDBY LIQUID CONTROL SYSTEM LIMITING CONDITION FOR OPERATION 3.1.5 The standby liquid control system shall be OPERABLE and consist of a minimum of the following:

a. In OPERATIONAL CONDITIONS 1 and 2, two pumps and corresponding flow paths,

b. In OPERATIONAL CONDITION 3, one pump and corresponding flow path.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3 ACTION:

a. With only one pump and corresponding explosive valve OPERABLE, in OPERATIONAL CONDITION 1 or 2, restore one inoperable pump and corresponding explosive valve to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

b. With standby liquid control system otherwise inoperable, in OPERATIONAL CONDITION 1, 2, or 3, restore the system to OPERABLE status within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.1.5 The standby liquid control system shall be demonstrated OPERABLE:

a. At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by verifying that:

1. The temperature of the sodium pentaborate solution is within the limits of Figure 3.1.5-1.

2. The available volume of sodium pentaborate solution is at least 3160 gallons.

3. The temperature of the pump suction piping is within the limits of Figure 3.1.5-1 for the most recent concentration analysis.

LIMERICK - UNIT 1 3/4 1-19 Amendment No. .59, &6, 8X,

REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

b. At least once per 31 days by:

1. Verifying the continuity of the explosive charge.

2. Determining by chemical analysis and calculation* that the available weight of Boron-10 is greater than or equal to 185 lbs; the concentration of sodium pentaborate in solution is less than or equal to 13.8% and within the limits of Figure 3.1.5-1 and; the following equation is satisfied:

C x E x Q 2 1 13% wt. 29 atom % 86 gpm where C = Sodium pentaborate solution (% by weight)

Q = Two pump flowrate, as determined per surveillance requirement 4.1.5.c.

E = Boron 10 enrichment (atom % Boron 10)

3. Verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.

c. Demonstrating that, when tested pursuant to Specification 4.0.5, the minimum flow requirement of 41.2 gpm per pump at a pressure of greater than or equal to 1230 +/- 25 psig is met.

d. At least once per 24 months during shutdown by:

1. Initiating at least one of the standby liquid control system loops, including an explosive valve, and verifying that a flow path from the pumps to the reactor pressure vessel is available by pumping demineralized water into the reactor vessel. The replacement charge for the explosive valve shall be from the same manufactured batch as the one fired or from another batch which has been certified by having one of the batch success-fully fired. All injection loops shall be tested in 3 operating cycles.

2. Verify all heat-treated piping between storage tank and pump suction is unblocked.**

e. Prior to addition of Boron to storage tank verify sodium pentaborate enrichment to be added is 2 29 atom % Boron 10.

• This test shall also be performed anytime water or boron is added to the solu-tion or when the solution temperature drops below the limits of Figure 3.1.5-1 for the most recent concentration analysis, within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after water or boron addition or solution temperature is restored.

• • This test shall also be performed whenever suction piping temperature drops below the limits of Figure 3.1.5-1 for the most recent concentration analysis, within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after solution temperature is restored.

LIMERICK - UNIT 1 3/4 1-20 Amendment No. -59, f4, 66, 9-, 4F6,

TABLE 3.3.2-1 (Continued)

ISOLATION ACTUATION INSTRUMENTATION ACTION STATEMENTS ACTION 20 - Be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the next 24-hours.

ACTION 21 - Be in at least STARTUP with the associated isolation valves closed within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> or be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

ACTION 22 - Be in at least STARTUP within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

ACTION 23 - In OPERATIONAL CONDITION 1 or 2, verify the affected system isolation valves are closed within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and declare the affected system inoperable. In OPERATIONAL CONDITION 3, be in at least COLD SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

ACTION 24 - Restore the manual initiation function to OPERABLE status within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or close the affected system isolation valves within the next hour and declare the affected system inoperable or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

ACTION 25 - Establish SECONDARY CONTAINMENT INTEGRITY with the standby gas treatment system operating within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

ACTION 26 - Close the affected system isolation valves within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

TABLE NOTATIONS

• Required when (1) handling RECENTLY IRRADIATED FUEL in the secondary containment, or (2) during operations with a potential for draining the reactor vessel with the vessel head removed and fuel in the vessel.

• • May be bypassed under administrative control, with all turbine stop valves closed.

1. During operation of the associated Unit 1 or Unit 2 ventilation exhaust system.

(a) DELETED (b) A channel may be placed in an inoperable status for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for required surveillance without placing the trip system in the tripped condition provided at least one OPERABLE channel in the same trip system is monitoring that parameter. Trip functions common to RPS Actuation Instrumentation are shown in Table 4.3.2.1-1. In addition, for the HPCI system and RCIC system isolation, provided that the redundant isolation valve, inboard or outboard, as applicable, in each line is OPERABLE and all required actuation instrumentation for that valve is OPERABLE, one channel may be placed in an inoperable status for up to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> for required surveillance without placing the channel or trip system in the tripped condition.

LIMERICK - UNIT 1 3/4 3-16 Amendment No. 2-3, 40, 53, 6g, 44X,

TABLE 4.3.2.1-1 (Continued)

ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL OPERATIONAL CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH TRIP FUNCTION CHECK TEST CALIBRATION SURVEILLANCE REOUIRED

7. SECONDARY CONTAINMENT ISOLATION

a. Reactor Vessel Water Level Low, Low - Level 2 S Q R 1, 2, 3

b. Drywell Pressure## - High S 0 R 1, 2, 3 c.1. Refueling Area Unit 1 Ventilation Exhaust Duct Radiation - High S Q R

2. Refueling Area Unit 2 Ventilation Exhaust Duct Radiation - High S Q R

d. Reactor Enclosure Ventilation Exhaust Duct Radiation - High S Q R 1, 2, 3

e. Deleted

f. Deleted

g. Reactor Enclosure Manual Initiation N.A. R N.A. 1, 2, 3

h. Refueling Area Manual Initiation N.A. R N.A. *

• Required when (1) handling RECENTLY IRRADIATED FUEL in the secondary containment, or (2) during operations I with a potential for draining the reactor vessel with the vessel head removed and fuel in the vessel.

• • When not administratively bypassed and/or when any turbine stop valve is open.

1. During operation of the associated Unit 1 or Unit 2 ventilation exhaust system.

1. 1. These trip functions (2a, 6b, and 7b) are common to the RPS actuation trip function.

LIMERICK - UNIT 1 3/4 3-31 Amendment No. 3, 40, 63, 49, 89, 44.1,

TABLE 3.3.7.1-1 RADIATION MONITORING INSTRUMENTATION MINIMUM CHANNELS APPLICABLE ALARM/TRIP INSTRUMENTATION OPERABLE CONDITIONS SETPOINT ACTION

1. Main Control Room Normal 4 1,2,3, 1 x 101 4Ci/cc"b) 70 Fresh Air Supply Radiation and

• Monitor

2. Area Monitors

a. Criticality Monitors

1) Spent Fuel 2 (a) > 5 mR/h and *2OmR/h(b) 71 Storage Pool

b. Control Room Direct 1 At All Times N.A.(b) 73 Radiation Monitor

3. Reactor Enclosure Cooling Water Radiation Monitor 1 At All Times

• 3 x Background(b) 72 LIMERICK - UNIT 1 3/4 3 -64 Amendment No.

TABLE 3.3.7.1-1 (Continued)

RADIATION MONITORING INSTRUMENTATION TABLE NOTATIONS

• When RECENTLY IRRADIATED FUEL is being handled in the secondary containment or during operations with a potential for draining the reactor vessel with the vessel head'removed and fuel in the'vessel./

(a) With fuel in the spent fuel storage pool.

(b) Alarm only.

ACTION STATEMENTS ACTION 70 With.one monitor inoperable, restore the inoperable monitor to the OPERABLE status within 7 days or, within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, initiate and maintain operation of the control room emergency filtration system in the radiation isolation mode of operation.

With two or more of the monitors inoperable, within one hour, initiate and maintain operation of the control room emergency filtration system in the radiation mode of operation.

ACTION 71 With one of the required monitor inoperable, assure a portable continuous monitor with the same alarm setpoint is OPERABLE in the vicinity of the installed monitor during any fuel movement.

If no fuel movement is being made, perform area surveys of the monitored area with portable monitoring instrumentation at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

ACTION 72 With the required monitor inoperable, obtain and analyze at least one grab sample of the monitored parameter at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

ACTION 73 With the required monitor inoperable, assure a portable alarming monitor is OPERABLE in the vicinity of the installed monitor or perform area surveys of the monitored area with portable monitor-ing instrumentation at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

LIMERICK - UNIT 1 3/4 3-65 Amendment No.

l TABLE 4.3.7.1-1 RADIATION MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS OPERATIONAL CHANNEL CONDITIONS FOR CHANNEL FUNCTIONAL CHANNEL WHICH SURVEILLANCE INSTRUMENTATION CHECK TEST CALIBRATION IS REOUIRED

1. Main Control Room Normal Fresh Air Supply Radiation Monitor S Q R 1, 2, 3, and

• I

2. Area Monitors

a. Criticality Monitors

1) Spent Fuel Storage S M R (a)

Pool

b. Control Room Direct S M R At All Times Radiation Monitor

3. Reactor Enclosure Cooling Water Radiation Monitor S M R(b) At All Times LIMERICK - UNIT 1 3/4 3-66 Amendment No. ;4,

TABLE 4.3.7.1-1 (Continued)

RADIATION MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS TABLE NOTATIONS

• When RECENTLY IRRADIATED FUEL is being handled in the secondary containment or during operations with a potential for draining the reactor vessel with the vessel head removed and fuel in the vessel.

(a) With fuel in the spent fuel storage pool.

(b) The initial CHANNEL CALIBRATION shall be performed using one or more of the reference standards certified by the National Bureau of Standards (NBS) or using standards that have been obtained from suppliers that participate in measurement assurance activities with NBS. These standards shall permit calibrating the system over its intended range of energy and measurement range. For subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration shall be used.

LIMERICK - UNIT 1 3/4 3-67 Amendment No.

REACTOR COOLANT SYSTEM 3/4.4.7 MAIN STEAM LINE ISOLATION VALVES LIMITING CONDITION FOR OPERATION 3.4.7 Two main steam line isolation valves (MSIVs) per main steam line shall be OPERABLE with closing times greater than or equal to 3 and less than or equal to 10 seconds. I APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3.

ACTION:

a. With one or more MSIVs inoperable:

1. Maintain at least one MSIV OPERABLE in each affected main steam line that is open and within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, either:

a) Restore the inoperable valve(s) to OPERABLE status, or b) Isolate the affected main steam line by use of a deactivated MSIV in the closed position.

2. Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

b. The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE REOUIREMENTS 4.4.7 Each of the above required MSIVs shall be demonstrated OPERABLE by verifying full closure between 3 and 10 seconds when tested pursuant to I Specification 4.0.5.

LIMERICK - UNIT 1 3/4 4-23 Amendment No.

CONTAINMENT SYSTEMS LIMITING CONDITION FOR OPERATION (Continued)

ACTION: (Continued)

b. The combined leakage rate to be in accordance with the Primary Containment Leakage Rate Testing Program for all primary containment penetrations and all primary containment isolation valves that are subject to Type B and C tests, except for: main steam line isolation valves*, valves which are hydrostatically tested, and those valves where an exemption to Appendix J of 10 CFR 50 has been granted, and

c. The leakage rate to *100 scf per hour for any main steam isolation valve I that exceeds 100 scf per hour, and restore the combined maximum pathway leakage to *200 scf per hour, and

d. The combined leakage rate for all containment isolation valves in hydrostatically tested lines which penetrate the primary containment to less than or equal to 1 gpm times the total number of such valves, prior to increasing the reactor coolant system temperature above 200'F.

SURVEILLANCE REQUIREMENTS 4.6.1.2 The primary containment leakage rates shall be demonstrated to be in accordance with the Primary Containment Leakage Rate Testing Program, or approved exemptions, for the following:

a. Type A Test

b. Type B and C Tests (including air locks)

c. Main Steam Line Isolation Valves

d. Hydrostatically tested Containment Isolation Valves

• Exemption to Appendix "J" to 10 CFR Part 50.

LIMERICK - UNIT 1 3/4 6-3 Amendment No. f7, 407, 14-8, 446,

CONTAINMENT SYSTEMS 3/4.6.5 SECONDARY CONTAINMENT REFUELING AREA SECONDARY CONTAINMENT INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.5.1.2 REFUELING AREA SECONDARY CONTAINMENT INTEGRITY shall be maintained.

APPLICABILITY: When RECENTLY IRRADIATED FUEL is being handled in the secondary containment, or during operations with a potential for draining the reactor vessel, with the vessel head removed and fuel in the vessel.

ACTION:

Without REFUELING AREA SECONDARY CONTAINMENT INTEGRITY, suspend handling of RECENTLY IRRADIATED FUEL in the secondary containment and operations with a potential for draining the reactor vessel. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.6.5.1.2 REFUELING AREA SECONDARY CONTAINMENT INTEGRITY shall be demonstrated by:

a. Verifying at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> that the pressure within the refueling area secondary containment is greater than or equal to 0.25 inch of vacuum water gauge.

b. Verifying at least once per 31 days that:

1. All refueling area secondary containment equipment hatches and blowout panels are closed and sealed.

2. At least one door in each access to the refueling area secondary containment is closed.

3. All refueling area secondary containment penetrations not capable of being closed by OPERABLE secondary containment automatic iso-lation dampers/valves and required to be closed during accident conditions are closed by valves, blind flanges, slide gate dampers or deactivated automatic dampers/valves secured in position.

c. At least once per 24 months:

Operating one standby gas treatment subsystem for one hour and main-taining greater than or equal to 0.25 inch of vacuum water gauge in the refueling area secondary containment at a flow rate not exceeding 764 cfm.

LIMERICK - UNIT 1 3/4 6-47 Amendment No. 29, ,

CONTAINMENT SYSTEMS REFUELING AREA SECONDARY CONTAINMENT AUTOMATIC ISOLATION VALVES LIMITING CONDITION FOR OPERATION 3.6.5.2.2 The refueling area secondary containment ventilation system automatic isolation valves shall be OPERABLE.

APPLICABILITY: When RECENTLY IRRADIATED FUEL is being handled in the secondary containment, or during operations with a potential for draining the reactor vessel, with the vessel head removed and fuel in the vessel.

ACTION:

With one or more of the refueling area secondary containment ventilation system automatic isolation valves inoperable, maintain at least one isolation valve OPERABLE in each affected penetration that is open and within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> either:

a. Restore the inoperable valves to OPERABLE status, or

b. Isolate each affected penetration by use of at least one deactivated valve secured in the isolation position, or

c. Isolate each affected penetration by use of at least one closed manual valve, blind flange or slide gate damper.

Otherwise, suspend handling of RECENTLY IRRADIATED FUEL in the refueling area secondary containment, and, operations with a potential for draining the reactor vessel. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REOUIREMENTS 4.6.5.2.2 Each refueling area secondary containment ventilation system auto-matic isolation valve shall be demonstrated OPERABLE:

a. Prior to returning the valve to service after maintenance, repair or replacement work is performed on the valve or its associated actuator, control or power circuit by cycling the valve through at least one complete cycle of full travel and verifying the specified isolation time.

b. At least once per 24 months by verifying that on a containment isolation test signal each isolation valve actuates to its isolation position.

c. By verifying the isolation time to be within its limit at least once per 92 days.

LIMERICK - UNIT 1 3/4 6-50 Amendment No. 6, 40, 74,40-0,

CONTAINMENT SYSTEMS STANDBY GAS TREATMENT SYSTEM - COMMON SYSTEM LIMITING CONDITION FOR OPERATION 3.6.5.3 Two independent standby gas treatment subsystems shall be OPERABLE.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3, and when RECENTLY IRRADIATED FUEL is being handled in the secondary containment, or during operations with a potential for draining the reactor vessel, with the vessel head removed and fuel in the vessel.

ACTION:

a. With one standby gas treatment subsystem inoperable, restore the inoperable subsystem to OPERABLE status within 7 days, or:

1. In OPERATIONAL CONDITION 1, 2, or 3, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

2. When RECENTLY IRRADIATED FUEL is being handled in the secondary containment, or during operations with a potential for draining the reactor vessel, with the vessel head removed and fuel in the vessel, suspend handling of RECENTLY IRRADIATED FUEL in the secondary containment and operations with a potential for draining the reactor vessel. The provisions of Specification 3.0.3 are not applicable.

b. With both standby gas treatment subsystems inoperable, if in progress, suspend handling of RECENTLY IRRADIATED FUEL'in the secondary containment or operations with a potential for draining the reactor vessel. The provisions of Specification 3.0.3. are not applicable.

SURVEILLANCE REQUIREMENTS 4.6.5.3 Each standby gas treatment subsystem shall be demonstrated OPERABLE:

a. At least once per 31 days by initiating, from the control room, flow through the HEPA filters and charcoal adsorbers and verifying that the subsystem operates with the heaters OPERABLE.

LIMERICK - UNIT 1 3/4 6-52 Amendment No. Z9, 49,

CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

b. At least once per 24* months or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire, or chemical release in any ventilation zone communicating with the subsystem by:

1. Verifying that the subsystem satisfies the in-place penetration and bypass leakage testing acceptance criteria of less than 0.05%

and uses the test procedure guidance in Regulatory Positions C.5.a, C.5.c and C.5.d of Regulatory Guide 1.52, Revision 2, March 1978, and the system flow rate is 5764 cfm + 10%.

2. Verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, shows the methyl iodide penetration of less than 1.25%

when tested in accordance with ASTM D3803-1989 at a temperature of 30'C (860 F), at a relative humidity of 70% and at a face velocity of 66 fpm.

3. Verify that when the fan is running the subsystem flowrate is 2800 cfm minimum from each reactor enclosure (Zones I and II) and 2200 cfm minimum from the refueling area (Zone III) when tested in accordance with ANSI N510-1980.

4. Verify that the pressure drop across the refueling area to SGTS prefilter is less than 0.25 inches water gage while operating at a flow rate of 2400 cfm +/- 10%.

c. After every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of charcoal adsorber operation by verifying within 31 days after removal that a laboratory analysis of a repre-sentative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, shows the methyl iodide penetration of less than 1.25% when tested in accordance with ASTM D3803-1989 at a temperature of 30'C (860 F),

at a relative humidity of 70% and at a face velocity of 66 fpm.

d. At least once per 24 months by:

1. Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than 9.1 inches water gauge while operating the filter train at a flow rate of 8400 cfm +/- 10%.

• Surveillance interval is an exception to the guidance provided in Regulatory Guide 1.52, Revision 2, March 1978.

LIMERICK - UNIT 1 3/4 6-53 Amendment No. 6, 33, A4, 49, 444,

CONTAINMENT SYSTEMS REACTOR ENCLOSURE RECIRCULATION SYSTEM LIMITING CONDITION FOR OPERATION 3.6.5.4 Two independent reactor enclosure recirculation subsystems shall be OPERABLE.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3.

ACTION:

a. With one reactor enclosure recirculation subsystem inoperable, restore the inoperable subsystem to OPERABLE status within 7 days, or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

b. With both reactor enclosure recirculation subsystems inoperable, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hors.

SURVEILLANCE REQUIREMENTS 4.6.5.4 Each reactor enclosure recirculation subsystem shall be demonstrated OPERABLE:

a. At least once per 31 days by initiating, from the control room, flow through the HEPA filters and charcoal adsorbers and verifying that the subsystem operates properly (flow at a minimum of 30,000 cfm).

b. At least once per 24* months or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire, or chemical release in any ventilation zone communi-cating with the subsystem by:

1. Verifying that the subsystem satisfies the in-place penetration and bypass leakage testing acceptance criteria of less than 0.05%

and uses the test procedure guidance in Regulatory Positions C.5.a, C.5.c, and C.5.d of Regulatory Guide 1.52, Revision 2, March 1978, at rated flow (60,000 cfm +/- 10%).

2. Verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, shows the methyl iodide penetration of less than 15%

when tested in accordance with ATM D3803-1989 at a temperature of 30'C (860F) and a relative humidity of 70%.

3. Verifying a subsystem flow rate within a range of 30,000 cfm to 66,000 cfm during system operation when tested in accordance with ANSI N510-1980.

• Surveillance interval is an exception to the guidance provided in Regulatory Guide 1.52, Revision 2, March 1978.

LIMERICK - UNIT 1 3/4 6-55 Amendment No. Z., 444,

CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

c. After every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of charcoal adsorber operation by verifying within 31 days after removal that a laboratory analysis of a repre-sentative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, shows, the methyl iodide penetration of less than 15% when tested in accordance with ASTM D3803-1989 at a temperature of 30'C (860 F) and a relative humidity of 70%.

d. At least once per 24 months by:

1. Verifying that the pressure drop across the combined prefilter, upstream and downstream HEPA filters, and charcoal adsorber banks is less than 6 inches water gauge while operating the filter train at rated flow (60,000 cfm +/- 10%), verifying that the prefilter pressure drop is less than 0.8 inch water gauge and that the pressure drop across each HEPA is less than 2 inches water gauge.

2. Verifying that the filter train starts and the isolation valves which take suction on and return to the reactor enclosure open on each of the following test signals:

a. Manual initiation from the control room, and

b. Simulated automatic initiation signal.

e. After each complete or partial replacement of a HEPA filter bank by verifying that the HEPA filter bank satisfies the inplace penetration and leakage testing acceptance criteria of less than 0.05% in accordance with ANSI N510-1980 while operating the system at rated flow (60,000 cfm +/- 10%).

f. After each complete or partial replacement of a charcoal adsorber bank by verifying that the charcoal adsorber bank satisfies the inplace penetration and leakage testing acceptance criteria of less than 0.05% in accordance with ANSI N510-1980 for a halogenated hydro-carbon refrigerant test gas while operating the system at rated flow (60,000 cfm +/- 10%).

LIMERICK - UNIT 1 3/4 6-56 Amendment No. .42, B4, 444,

PLANT SYSTEMS EMERGENCY SERVICE WATER SYSTEM - COMMON SYSTEM LIMITING CONDITION FOR OPERATION 3.7.1.2 At least the following independent emergency service water system loops, with each loop comprised of:

a. Two OPERABLE emergency service water pumps, and

b. An OPERABLE flow path capable of taking suction from the emergency service water pumps wet pits which are supplied from the spray pond or the cooling tower basin and transferring the water to the associated Unit 1 and common safety-related equipment, shall be OPERABLE:

a. Two loops, in OPERATIONAL CONDITIONS 1, 2, and 3.

b. One loop, in OPERATIONAL CONDITIONS 4, 5, and when RECENTLY IRRADIATED FUEL is being handled in the secondary containment, or during operations with a potential for draining the reactor vessel.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3, 4, 5, and when RECENTLY IRRADIATED FUEL is being handled in the secondary containment, or during operations with a potential for draining the reactor vessel.

ACTION:

a. In OPERATION CONDITION 1, 2, or 3:

1. With one emergency service water pump inoperable, restore the inoperable pump to OPERABLE status within 45 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

2. With one emergency service water pump in each loop inoperable, restore at least one inoperable pump to OPERABLE status within 30 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

3. With one emergency service water system loop otherwise inoperable, declare all equipment aligned to the inoperable loop inoperable**, restore the inoperable loop to OPERABLE status with at least one OPERABLE pump within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

• • The diesel generators may be aligned to the OPERABLE emergency service water system loop provided confirmatory flow testing has been performed. Those diesel generators no aligned to the OPERABLE emergency service water system loop shall be declared inoperable and the actions of 3.8.1.1 taken.

LIMERICK - UNIT 1 3/4 7-3 Amendment No. 2-7, 40, 86, 1431,

PLANT SYSTEMS LIMITING CONDITION FOR OPERATION (Continued)

ACTION: (Continued)

4. With three ESW pump/diesel generator pairs** inoperable, restore at least one inoperable ESW pump/diesel generator pair** to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

5. With four ESW pump/diesel generator pairs** inoperable, restore at least one inoperable ESW pump/diesel generator pair** to OPERABLE status within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

b. In OPERATIONAL CONDITION 4 or 5:

1. With only one emergency service water pump and its associated flowpath OPERABLE, restore at least two pumps with at least one flow path to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or declare the associated safety related equipment inoperable and take the ACTION required by Specifications 3.5.2 and 3.8.1.2.

c. When RECENTLY IRRADIATED FUEL is being handled in the secondary containment, or during operations with a potential for draining the reactor vessel:

1. With only one emergency service water pump and its associated flow path OPERABLE, restore at least two pumps with at least one flow path to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or verify adequate cooling remains available for the diesel generators required to be OPERABLE or declare the associated diesel generator(s) inoperable and take the ACTION required by Specification 3.8.1.2. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REOUIREMENT 4.7.1.2 At least the above required emergency service water system loop(s) shall be demonstrated OPERABLE:

a. At least once per 31 days by verifying that each valve (manual, power-operated, or automatic) that is not locked, sealed, or otherwise secured in position, is in its correct position.

b. At least once per 24 months by verifying that:

1. Each automatic valve actuates to its correct position on its appropriate ESW pump start signal.

2. Each pump starts automatically when its associated diesel generator starts.

• • An ESW pump/diesel generator pair consists of an ESW pump and its associated diesel generator. If either an ESW pump or its associated diesel generator becomes inoperable, then the ESW pump/diesel generator pair is inoperable.

L.IMERICK - UNIT 1 3/4 7-4 Amendment No. 24, 44, ;4,

PLANT SYSTEMS ULTIMATE HEAT SINK LIMITING CONDITION FOR OPERATION 3.7.1.3 The spray pond shall be OPERABLE with:

a. A minimum pond water level at or above elevation 250' 10" Mean Sea Level, and

b. A pond water temperature of less than or equal to 88'F.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3, 4,. 5, and when RECENTLY IRRADIATED FUEL is being handled in the secondary containment, or during operations with a potential for draining the reactor vessel.

ACTION:

With the requirements of the above specification not satisfied:

a. In OPERATIONAL CONDITION 1, 2, or 3, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

b. In OPERATIONAL CONDITION 4 or 5, declare the RHRSW system and the emergency service water system inoperable and take the ACTION required by Specifications 3.7.1.1 and 3.7.1.2.

c. When RECENTLY IRRADIATED FUEL is being handled in the secondary containment, or during operations with a potential for draining the reactor vessel, declare the emergency service water system inoperable and take the ACTION required by Specification 3.7.1.2. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REOUIREMENTS 4.7.1.3 The spray pond shall be determined OPERABLE:

a. By verifying the pond water level to be greater than its limit at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

b. By verifying the water surface temperature (within the upper two feet of the surface) to be less than or equal to 880 F:

1. at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> when the spray pond temperature is greater than or equal to 80'F; and

2. at least once per 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> when the spray pond temperature is greater than or equal to 850 F; and

3. at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when the spray pond temperature is greater than 320 F.

c. By verifying all piping above the frost line is drained:

1. within one (1) hour after being used when ambient air temperature is below 40'F; or

2. when ambient air temperature falls below 40'F if the piping has not been previously drained.

LIMERICK - UNIT I 3/4 7-5 Amendment No. 2-5, 44, 90,

PLANT SYSTEMS 3/4.7.2 CONTROL ROOM EMERGENCY FRESH AIR SUPPLY SYSTEM - COMMON SYSTEM LIMITING CONDITION FOR OPERATION 3.7.2 Two independent control room emergency fresh air supply system subsystems shall be OPERABLE.

APPLICABILITY: All OPERATIONAL CONDITIONS and when RECENTLY IRRADIATED FUEL is being handled in the secondary containment, or during operations with a potential for draining the reactor vessel.

ACTION:

a. In OPERATIONAL CONDITION 1, 2, or 3 with one control room emergency fresh air supply subsystem inoperable, restore the inoperable subsystem to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

b. In OPERATIONAL CONDITION 4, 5, or when RECENTLY IRRADIATED FUEL is being handled in the secondary containment, or during operations with a potential for draining the reactor vessel:

1. With one control room emergency fresh air supply subsystems inoperable, restore the inoperable subsystem to OPERABLE status within 7 days or initiate and maintain operation of the OPERABLE subsystem in the radiation isolation mode of operation.

2. With both control room emergency fresh air supply subsystems inoperable, suspend handling of RECENTLY IRRADIATED FUEL in the secondary containment and operations with a potential for draining the reactor vessel. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.7.2 Each control room emergency fresh air supply subsystem shall be demonstrated OPERABLE:

a. At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by verifying the control room air tempera-ture to be less than or equal to 85'F effective temperature.

b. At least once per 31 days on a STAGGERED TEST BASIS by initiating, from the control room, flow through the HEPA filters and charcoal adsorbers and verifying that the subsystem operates with the heaters OPERABLE.

c. At least once per 24** months or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire, or chemical release in any ventilation zone communicating with the subsystem by:

1. Verifying that the subsystem satisfies the in-place penetration and bypass leakage testing acceptance criteria of less than 0.05% and uses the test procedure guidance in Regulatory Positions C.5.a, C.5.c, and C.5.d of Regulatory Guide 1.52, Revision 2, March 1978, and the system flow rate is 3000 cfm +/- 10%.

• • Surveillance interval is an exception to the guidance provided in Regulatory Guide 1.52, Revision 2, March 1978.

LIMERICK - UNIT 1 3/4 7 -6 Amendment No. 40, 74,

PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

2. Verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, shows the methyl iodide penetration of less than 10%

when tested in accordance with ASTM D3803-1989 at a temperature of 30'C (860F) and a relative humidity of 70%.

3. Verifying a subsystem flow rate of 3000 cfm +/- 10% during subsystem operation when tested in accordance with ANSI N510-1980.

d. After every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of charcoal adsorber operation by verifying within 31 days after removal that a laboratory analysis of a repre-sentative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, shows the methyl iodide penetration of less than 10%

when tested in accordance with ASTM D3803-1989 at a temperature of 30'C (860F) and a relative humidity of 70%.

e. At least once per 24 months by:

1. Verifying that the pressure drop across the combined prefilter, upstream and downstream HEPA filters, and charcoal adsorber banks is less than 6 inches water gauge while operating the subsystem at a flow rate of 3000 cfm +/- 10%; verifying that the prefilter pressure drop is less than 0.8 inch water gauge and that the pressure drop across each HEPA is less than 2 inches water gauge.

2. Verifying that on each of the below chlorine isolation mode actuation test signals, the subsystem automatically switches to the chlorine isolation mode of operation and the isolation valves close within 5 seconds:

a) Outside air intake high chlorine, and b) Manual initiation from the control room.

3. Verifying that on manual initiation from the control room, the subsystem switches to the radiation isolation mode of operation and the control room is maintained at a positive pressure of at least 1/8 inch water gauge relative to the turbine enclosure and auxiliary equipment room and outside atmosphere during subsystem operation with an outdoor air flow rate less than or equal to 525 cfm.

LIMERICK - UNIT 1 3/4 7-7 Amendment No. -5, 40, W4, 4-44,

ELECTRICAL POWER SYSTEMS A.C. SOURCES - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.8.1.2 As a minimum, the following A.C. electrical power sources shall be OPERABLE:

a. One circuit between the offsite transmission network and the onsite Class 1E distribution system, and

b. Two diesel generators each with:

1. A day fuel tank containing a minimum of 200 gallons of fuel.

2. A fuel storage system containing a minimum of 33,500 gallons of fuel.

3. A fuel transfer pump.

APPLICABILITY: OPERATIONAL CONDITIONS 4, 5, and when RECENTLY IRRADIATED FUEL is being handled in the secondary containment, or during operations with a potential for draining the reactor vessel.

ACTION:

a. With less than the above required A.C. electrical power sources OPERABLE, suspend CORE ALTERATIONS, handling of RECENTLY IRRADIATED FUEL in the secondary containment, operations with a potential for draining the reactor vessel and crane operations over the spent fuel storage pool when fuel assemblies are stored therein. In addition, when in OPERATIONAL CONDITION 5 with the water level less than 22 feet above the reactor pressure vessel flange, immediately initiate corrective action to restore the required power sources to OPERABLE status as soon as practical.

b. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.8.1.2 At least the above required A.C. electrical power sources shall be demonstrated OPERABLE per Surveillance Requirements 4.8.1.1.1, 4.8.1.1.2, and 4.8.1.1.3.

LIMERICK - UNIT 1 3/4 8-9 Amendment No. 32-,

ELECTRICAL POWER SYSTEMS D.C. SOURCES - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.8.2.2 As a minimum, two of the following four divisions of the D.C.

electrical power sources system shall be OPERABLE with:

a. Division 1, Consisting of:

1. 125-Volt Battery lAl (lAlD101).

2. 125-Volt Battery 1A2 C1A2D101).

3. 125-Volt Battery Charger 1BCA1 C1A1D103).

4. 125-Volt Battery Charger 1BCA2 (1A2D103).

b. Division 2, Consisting of:

1. 125-Volt Battery iB1 (1BID101)

2. 125-Volt Battery 1B2 (1B2D1O1)

3. 125-Volt Battery Charger 1BCB3I (IBlD103).

4. 125-Volt Battery Charger 1BCB27 CB2D103).

c. Division 3, Consisting of:

1. 125-Volt Battery 1C (1CD101).

2. 125-Volt Battery Charger 1BCC (CCD103).

d. Division 4, Consisting of:

1. 125-Volt Battery iD (1DD101).

2. 125-Volt Battery Charger 1BCD (IDD103).

APPLICABILITY: OPERATIONAL CONDITIONS 4, 5, and when RECENTLY IRRADIATED FUEL is being handled in the secondary containment, or during operations with a potential for draining the reactor vessel.

ACTION:

a. With one or two required battery chargers on one required division inoperable:

1. Restore battery terminal voltage to greater than or equal to the minimum established float voltage within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />,

2. Verify associated Division 1 or 2 float current < 2 amps, or Division 3 or 4 float current < 1 amp within 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br /> and once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter, and

3. Restore battery charger(s) to OPERABLE status within 7 days.

b. With one or more required batteries inoperable due to:

1. One or two batteries on one division with one or more battery cells float voltage < 2.07 volts, perform 4.8.2.1.a.1 and 4.8.2.1.a.2 within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for affected battery(s) and restore affected cell(s) voltage > 2.07 volts within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

I LIMERICK - UNIT I 3/4 8-14 Amendment No. 1-64,

ELECTRICAL POWER SYSTEMS LIMITING CONDITION FOR OPERATION (Continued)

ACTION: (Continued)

2. Division 1 or 2 with float current > 2 amps, or with Division 3 or 4 with float current > 1 amp, perform 4.8.2.1.a.2 within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for affected battery(s) and restore battery float current to within limits within 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br />.

3. One or two batteries on one division with one or more cells electrolyte level less than minimum established design limits, if electrolyte level was below the top of the plates restore electrolyte level to above top of plates within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and verify no evidence of leakage(*) within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. In all cases, restore electrolyte level to greater than or equal to minimum established design limits within 31 days.

4. One or two batteries on one division with pilot cell electrolyte temperature less than minimum established design limits, restore battery pilot cell temperature to greater than or equal to minimum established design limits within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

5. Batteries in more than one division affected, restore battery parameters for all batteries in one division to within limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

6. (i) Any battery having both (Action b.1) one or more battery cells float voltage < 2.07 volts and (Action b.2) float current not within limits, and/or (ii)Any battery not meeting any Action b.1 through b.5, Restore the battery parameters to within limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

c. 1. With the requirements of Action a. and/or Action b. not met, or

2. With less than two divisions of the above required D.C. electrical power sources OPERABLE for reasons other than Actions a. and/or b.,

Suspend CORE ALTERATIONS, handling of RECENTLY IRRADIATED FUEL in the secondary containment and operations with a potential for draining the reactor vessel.

d. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REOUIREMENTS 4.8.2.2 At least the above required battery and charger shall be demonstrated OPERABLE per Surveillance Requirement 4.8.2.1.

(*) Contrary to the provisions of Specification 3.0.2, if electrolyte level was below the top of the plates, the verification that there is no evidence of leakage is required to be completed regardless of when electrolyte level is restored.

LIMERICK - UNIT 1 3/4 8-14a Amendment No. 4-64,

ELECTRICAL POWER SYSTEMS LIMITING CONDITION FOR OPERATION (Continued)

APPLICABILITY: OPERATIONAL CONDITIONS 4, 5, and when RECENTLY IRRADIATED FUEL is being handled in the secondary containment, or during operations with a potential for draining the reactor vessel.

ACTION:

a. With less than two divisions of the above required Unit 1 A.C. dis-tribution systems energized, suspend CORE ALTERATIONS, handling of RECENTLY IRRADIATED FUEL in the secondary containment and operations I with a potential for draining the reactor vessel.

b. With less than two divisions of the above required Unit 1 D.C. dis-tribution systems energized, suspend CORE ALTERATIONS, handling of RECENTLY IRRADIATED FUEL in the secondary containment and operations with a potential for draining the reactor vessel.

c. With any of the above required Unit 2 and common AC and/or DC distribution system divisions not energized, declare the associated common equipment inoperable, and take the appropriate ACTION for that system.

d. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REOUIREMENTS 4.8.3.2 At least the above required power distribution system divisions shall be determined energized at least once per 7 days by verifying correct breaker alignment and voltage on the busses/MCCs/panels.

I LIMERICK - UNIT 1 3/4 8-20 Amendment No. 2-4,

DEFINITIONS CORE ALTERATION 1.7 CORE ALTERATION shall be the movement of any fuel, sources, or reactivity control components within the reactor vessel with the vessel head removed and fuel in the vessel. The following exceptions are not considered to be CORE ALTERATIONS:

a) Movement of source range monitors, local power range monitors, intermediate range monitors, traversing incore probes, or special moveable detectors (including undervessel replacement); and b) Control rod movement, provided there are no fuel assemblies in the associated core cell.

Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe position.

CORE OPERATING LIMITS REPORT 1.7a The CORE OPERATING LIMITS REPORT (COLR) is the unit-specific document that provides the core operating limits for the current operating reload cycle. These cycle-specific core operating limits shall be determined for each reload cycle in accordance with Specifications 6.9.1.9 thru 6.9.12. Plant operation within these limits is addressed in individual specifications.

CRITICAL POWER RATIO 1.8 The CRITICAL POWER RATIO (CPR) shall be the ratio of that power in the assembly which is calculated by application of the (GEXL) correlation to cause some point in the assembly to experience boiling transition, divided by the actual assembly operating power.

DOSE EQUIVALENT I-131 1.9 DOSE EQUIVALENT I-131 shall be that concentration of I-131, microcuries per gram, which alone would produce the same inhalation committed effective dose equivalent (CEDE) as the quantity and isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present. The inhalation committed effective dose equivalent (CEDE) conversion factors used for this calculation shall be those listed in Table 2.1 of Federal Guidelines Report 11, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion," ORNL, 1989, as described in Regulatory Guide 1.183. The factors in the column headed "effective" yield doses corresponding to the CEDE.

DOWNSCALE TRIP SETPOINT (DTSP) 1.9a The downscale trip setpoint associated with the Rod Block Monitor (RBM) rod block trip setting.

E-AVERAGE DISINTEGRATION ENERGY 1.10 E shall be the average, weighted in proportion to the concentration of each radionuclide in the reactor coolant at the time of sampling, of the sum of the average beta and gamma energies per disintegration, in MeV, for isotopes, with half lives greater than 15 minutes, making up at least 95% of the total noniodine activity in the coolant.

EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME 1.11 The EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ECCS actuation set-point at the channel sensor until the ECCS equipment is capable of performing its safety function, i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc. Times shall include diesel generator starting and sequence loading delays where applicable. The response time may be measured by any series of sequential, overlapping or total steps such that the entire response time is measured.

LIMERICK - UNIT 2 1-2 Amendment No. 4, 48, 49,

DEFINITIONS PURGE - PURGING 1.31 PURGE or PURGING shall be the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is required to purify the confinement.

RATED THERMAL POWER 1.32 RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 3458 MWt.

REACTOR ENCLOSURE SECONDARY CONTAINMENT INTEGRITY 1.33 REACTOR ENCLOSURE SECONDARY CONTAINMENT INTEGRITY shall exist when:

a. All reactor enclosure secondary containment penetrations required to be closed during accident conditions are either:

1. Capable of being closed by an OPERABLE secondary containment automatic isolation system, or

2. Closed by at least one manual valve, blind flange, slide gate damper or deactivated automatic valve secured in its closed position, except as provided by Specification 3.6.5.2.1.

b. All reactor enclosure secondary containment hatches and blowout panels are closed and sealed.

c. The standby gas treatment system is in compliance with the requirements of Specification 3.6.5.3.

d. The reactor enclosure recirculation system is in compliance with the requirements of Specification 3.6.5.4.

e. At least one door in each access to the reactor enclosure secondary containment is closed.

f. The sealing mechanism associated with each reactor enclosure secondary containment penetration, e.g., welds, bellows, or O-rings, is OPERABLE.

g. The pressure within the reactor enclosure secondary containment is less than or equal to the value required by Specification 4.6.5.1.1a.

REACTOR PROTECTION SYSTEM RESPONSE TIME 1.34 REACTOR PROTECTION SYSTEM RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its trip setpoint at the channel sensor until de-energization of the scram pilot valve solenoids. The response time may be measured by any series of sequential, overlapping or total steps such that the entire response time is measured.

RECENTLY IRRADIATED FUEL 1.35 RECENTLY IRRADIATED FUEL is fuel that has occupied part of a critical reactor core within the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

REFUELING FLOOR SECONDARY CONTAINMENT INTEGRITY 1.36 REFUELING FLOOR SECONDARY CONTAINMENT INTEGRITY shall exist when:

a. All refueling floor secondary containment penetrations required to be closed during accident conditions are either:

LIMERICK - UNIT 2 1-6 Amendment No. 48, A4, 69,

DEFINITIONS REFUELING FLOOR SECONDARY CONTAINMENT INTEGRITY (Continued)

1. Capable of being closed by an OPERABLE secondary containment automatic isolation system, or

2. Closed by at least one manual valve, blind flange, slide gate damper or deactivated automatic valve secured in its closed position, except as provided by Specification 3.6.5.2.2.

b. All refueling floor secondary containment hatches and blowout panels are closed and sealed.

c. The standby gas treatment system is in compliance with the requirements of Specification 3.6.5.3.

d. At least one door in each access to the refueling floor secondary containment is closed.

e. The sealing mechanism associated with each refueling floor secondary containment penetration, e.g., welds, bellows, or 0-rings, is OPERABLE.

f. The pressure within the refueling floor secondary containment is less than or equal to the value required by Specification 4.6.5.1.2a.

REPORTABLE EVENT 1.37 A REPORTABLE EVENT shall be any of those conditions specified in Section 50.73 to 10 CFR Part 50.

ROD DENSITY 1.38 ROD DENSITY shall be the number of control rod notches inserted as a fraction of the total number of control rod notches. All rods fully inserted is equivalent to 100% ROD DENSITY.

SHUTDOWN MARGIN 1.39 SHUTDOWN MARGIN shall be the amount of reactivity by which the reactor is subcritical or would be subcritical assuming all control rods are fully inserted except for the single control rod of highest reactivity worth which is assumed to be fully withdrawn and the reactor is in the shutdown condition; cold, i.e. 680F; and xenon free.

SITE BOUNDARY 1.40 The SITE BOUNDARY shall be that line as defined in Figure 5.1.3-la.

SOURCE CHECK 1.41 A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a radioactive source.

LIMERICK - UNIT 2 1-7 Amendment No. 44, 48, fi.9,

REACTIVITY CONTROL SYSTEMS 3/4.1.5 STANDBY LIQUID CONTROL SYSTEM LIMITING CONDITION FOR OPERATION 3.1.5 The standby liquid control system shall be OPERABLE and consist of a minimum of the following:

a. In OPERATIONAL CONDITIONS 1 and 2, two pumps and corresponding flow paths,

b. In OPERATIONAL CONDITION 3, one pump and corresponding flow path.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2 and 3 ACTION:

a. With only one pump and corresponding explosive valve OPERABLE, in OPERATIONAL CONDITION 1 or 2, restore one inoperable pump and corresponding explosive valve to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

b. With standby liquid control system otherwise inoperable, in OPERATIONAL CONDITION 1, 2, or 3, restore the system to OPERABLE status within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

SURVEILLANCE REOUIREMENTS 4.1.5 The standby liquid control system shall be demonstrated OPERABLE:

a. At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by verifying that:

1. The temperature of the sodium pentaborate solution is within the limits of Figure 3.1.5-1.

2. The available volume of sodium pentaborate solution is at least 3160 gallons.

3. The temperature of the pump suction piping is within the limits of Figure 3.1.5-1 for the most recent concentration analysis.

LIMERICK - UNIT 2 3/4 1-19 4, 48, 49, Amendment No. .-

REACTIVITY CONTROL SYSTEMS SURVEILLANCE REOUIREMENTS (Continued)

b. At least once per 31 days by:

1. Verifying the continuity of the explosive charge.

2. Determining by chemical analysis and calculation* that the available weight of Boron-10 is greater than or equal to 185 lbs; the concentration of sodium pentaborate in solution is less than or equal to 13.8% and within the limits of Figure 3.1.5-1 and; the following equation is satisfied:

C x E x 0 21 13% wt. 29 atom % 86 gpm where C = Sodium pentaborate solution (% by weight)

Q = Two pump flowrate, as determined per surveillance requirement 4.1.5.c.

E = Boron 10 enrichment (atom % Boron 10)

3. Verifying that each valve (manual , power-operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.

c. Demonstrating that, when tested pursuant to Specification 4.0.5, the minimum flow requirement of 41.2 gpm per pump at a pressure of greater than or equal to 1230+/-25 psig is met.

d. At least once per 24 months during shutdown by:

1. Initiating at least one of the standby liquid control system loops, including an explosive valve, and verifying that a flow path from the pumps to the reactor pressure vessel is available by pumping demineralized water into the reactor vessel. The replacement charge for the explosive valve shall be from the same manufactured batch as the one fired or from another batch which has been certified by having one of the batch success-fully fired. All injection loops shall be tested in 3 operating cycles.

2. Verify all heat-treated piping between storage tank and pump suction is unblocked.**

e. Prior to addition of Boron to storage tank verify sodium pentaborate enrichment to be added is 2 29 atom % Boron 10.

• This test shall also be performed anytime water or boron is added to the solu-tion or when the solution temperature drops below the limits of Figure 3.1.5-1 for the most recent concentration analysis, within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after water or boron addition or solution temperature is restored.

• • This test shall also be performed whenever suction piping temperature drops below the limits of Figure 3.1.5-1 for the most recent concentration analysis, within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after solution temperature is restored.

LIMERICK - UNIT 2 3/4 1-20 Amendment No. ?4, 26, 34, 48, At,

TABLE 3.3.2-1 (Continued)

ISOLATION ACTUATION INSTRUMENTATION ACTION STATEMENTS ACTION 20 - Be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

ACTION 21 - Be in at least STARTUP with the associated isolation valves closed within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> or be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

ACTION 22 - Be in at least STARTUP within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

ACTION 23 - In OPERATIONAL CONDITION 1 or 2, verify the affected system isolation valves are closed within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and declare the affected system inoperable. In OPERATIONAL CONDITION 3, be in at least COLD SHUTDOWN within'12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

ACTION 24 - Restore the manual initiation function to OPERABLE status within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or close the affected system isolation valves within the next hour and declare the affected system inoperable or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

ACTION 25 - Establish SECONDARY CONTAINMENT INTEGRITY with the standby gas treatment system operating within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

ACTION 26 - Close the affected system isolation valves within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

TABLE NOTATIONS

• Required when (1) handling RECENTLY IRRADIATED FUEL in the secondary containment, or (2) during operations with a potential for draining the reactor vessel with the vessel head removed and fuel in the vessel.

• • May be bypassed under administrative control, with all turbine stop valves closed.

1. During operation of the associated Unit 1 or Unit 2 ventilation exhaust system.

(a) DELETED (b) A channel may be placed in an inoperable status for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for required surveillance without placing the trip system in the tripped condition provided at least one OPERABLE channel in the same trip system is monitoring that parameter. Trip functions common to RPS Actuation Instrumentation are shown in Table 4.3.2.1-1. In addition, for the HPCI system and RCIC system isolation, provided that the redundant isolation valve, inboard or outboard, as applicable, in each line is OPERABLE and all required actuation instrumentation for that valve is OPERABLE, one channel may be placed in an inoperable status for up to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> for required surveillance without placing the channel or trip system in the tripped condition.

LIMERICK - UNIT 2 3/4 3-16 Amendment No. 4-7, 3, 407,

TABLE 4.3.2.1-1 (Continued)

ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL OPERATIONAL CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH TRIP FUNCTION CHECK TEST CALIBRATION SURVEILLANCE REOUIRED

7. SECONDARY CONTAINMENT ISOLATION

a. Reactor Vessel Water Level Low, Low - Level 2 S Q R 1, 2, 3

b. Drywell Pressure## - High S Q R 1, 2, 3 c.1. Refueling Area Unit 1 Ventilation Exhaust Duct Radiation - High S 0 R *#

2. Refueling Area Unit 2 Ventilation Exhaust Duct Radiation - High S Q R

d. Reactor Enclosure Ventilation Exhaust Duct Radiation - High S Q R 1, 2, 3

e. Deleted

f. Deleted

g. Reactor Enclosure Manual Initiation N.A. R N.A. 1, 2, 3

h. Refueling Area Manual Initiation N.A. R N.A. *

• Required when (1) handling RECENTLY IRRADIATED FUEL in the secondary containment, or (2) during operations with a I potential for draining the reactor vessel with the vessel head removed and fuel in the vessel.

• • When not administratively bypassed and/or when any turbine stop valve is open.

1. During operation of the associated Unit 1 or Unit 2 ventilation exhaust system.

1. 1. These trip functions (2a, 6b, and 7b) are common to the RPS actuation trip function.

LIMERICK - UNIT 2 3/4 3-31 Amendment No. t, 3X, Em, 74,

TABLE 3.3.7.1-1 RADIATION MONITORING INSTRUMENTATION MINIMUM CHANNELS APPLICABLE ALARM/TRIP INSTRUMENTATION OPERABLE CONDITIONS SETPOINT ACTION

1. Main Control Room Normal 4 1,2,3, 1 x lo gCi/cc(b) 70 Fresh Air Supply Radiation and

• Monitor

2. Area Monitors

a. Criticality Monitors

1) Spent Fuel 2 (a) 2 5 mR/h and *2OmR/h(b) 71 Storage Pool

b. Control Room Direct 1 At All Times N.A.(b) 73 Radiation Monitor

3. Reactor Enclosure Cooling Water Radiation Monitor 1 At All Times

• 3 x Background(b) 72 LIMERICK - UNIT 2 3/4 3-64 Amendment No.

TABLE 3.3.7.1-1 (Continued)

RADIATION MONITORING INSTRUMENTATION TABLE NOTATIONS

• When RECENTLY IRRADIATED FUEL is being handled in the secondary containment or during operations with a potential for draining the reactor vessel with the vessel head removed and fuel in the vessel.

(a) With fuel in the spent fuel storage pool.

(b) Alarm only.

ACTION STATEMENTS ACTION 70 - With one monitor inoperable, restore the inoperable monitor to the OPERABLE status within 7 days or, within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, initiate and maintain operation of the control room emergency filtration system in the radiation isolation mode of operation.

With two or more of the monitors inoperable, within one hour, initiate and maintain operation of the control room emergency filtration system in the radiation mode of operation.

ACTION 71 With one of the required monitor inoperable, assure a portable continuous monitor with the same alarm setpoint is OPERABLE in the vicinity of the installed monitor during any fuel movement.

If no fuel movement is being made, perform area surveys of the monitored area with portable monitoring instrumentation at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

ACTION 72 With the required monitor inoperable, obtain and analyze at least one grab sample of the monitored parameter at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

ACTION 73 With the required monitor inoperable, assure a portable alarming monitor is OPERABLE in the vicinity of the installed monitor or perform area surveys of the monitored area with portable monitor-ing instrumentation at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

LIMERICK - UNIT 2 3/4 3-65 Amendment No.

TABLE 4.3.7.1-1 RADTATION MONITORING INSTRUMENTATION SURVEIL LANCE REOUIREMENTS OPERATIONAL CHANNEL CONDITIONS FOR CHANNEL FUNCTIONAL CHANNEL WHICH SURVEILLANCE INSTRUMENTATION . CHECK TEST CALIBRATION IS REOUIRED

1. Main Control Room Normal Fresh Air Supply Radiation Monitor S 0 R 1, 2, 3, and

• I

2. Area Monitors

a. Criticality Monitors

1) Spent Fuel Storage S M R (a)

Pool

b. Control Room Direct S M R At All Times Radiation Monitor

3. Reactor Enclosure Cooling Water Radiation Monitor S M R(b) At All Times LIMERICK UNIT 2 3/4 3-66 Amendment No. 33,

TABLE 4.3.7.1-1 (Continued)

RADIATION MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS TABLE NOTATIONS

• When RECENTLY IRRADIATED FUEL is being handled in the secondary containment or during operations with a potential for draining the reactor vessel with the vessel head removed and fuel in the vessel.

(a) With fuel in the spent fuel storage pool.

(b) The initial CHANNEL CALIBRATION shall be performed using one or more of the reference standards certified by the National Bureau of Standards (NBS) or using standards that have been obtained from suppliers that participate in measurement assurance activities with NBS. These standards shall permit calibrating the system over its intended range of energy and measurement range. For subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration shall be used.

LIMERICK - UNIT 2 3/4 3-67 Amendment No.

REACTOR COOLANT SYSTEM 3/4.4.7 MAIN STEAM LINE ISOLATION VALVES LIMITING CONDITION FOR OPERATION 3.4.7 Two main steam line isolation valves (MSIVs) per main steam line shall be OPERABLE with closing times greater than or equal to 3 and less than or equal to 10 seconds. I APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3.

ACTION:

a. With one or more MSIVs inoperable:

1. Maintain at least one MSIV OPERABLE in each affected main steam line that is open and within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, either:

a) Restore the inoperable valve(s) to OPERABLE status, or b) Isolate the affected main steam line by use of a deactivated MSIV in the closed position.

2. Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

b. The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE REOUIREMENTS 4.4.7 Each of the above required MSIVs shall be demonstrated OPERABLE by verifying full closure between 3 and 10 seconds when tested pursuant to I Specification 4.0.5.

LIMERICK - UNIT 2 3/4 4-23 Amendment No.

CONTAINMENT SYSTEMS LIMITING CONDITION FOR OPERATION (Continued)

ACTION: (Continued)

b. The combined leakage rate to be in accordance with the Primary Containment Leakage Rate Testing Program for all primary containment penetrations and all primary containment isolation valves that are subject to Type B and C tests, except for: main steam line isolation valves*, valves which are hydrostatically tested, and those valves where an exemption to Appendix J of 10 CFR 50 has been granted, and

c. The leakage rate to <100 scf per hour for any main steam isolation valve I that exceeds 100 scf per hour, and restore the combined maximum pathway leakage to <200 scf per hour, and

d. The combined leakage rate for all containment isolation valves in hydrostatically tested lines which penetrate the primary containment to less than or equal to 1 gpm times the total number of such valves, prior to increasing reactor coolant system temperature above 200'F.

SURVEILLANCE REQUIREMENTS 4.6.1.2 The primary containment leakage rates shall be demonstrated to be in accordance with the Primary Containment Leakage Rate Testing Program, or approved exemptions, for the following:

a. Type A Test

b. Type B and C Tests (including air locks)

c. Main Steam Line Isolation Valves

d. Hydrostatically tested Containment Isolation Valves

• Exemption to Appendix "J" to 10 CFR Part 50.

LIMERICK - UNIT 2 3/4 6-3 Amendment No. -53, , 8., 4-7,

CONTAINMENT SYSTEMS 3/4.6.5 SECONDARY CONTAINMENT REFUELING AREA SECONDARY CONTAINMENT INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.5.1.2 REFUELING AREA SECONDARY CONTAINMENT INTEGRITY shall be maintained.

APPLICABILITY: When RECENTLY IRRADIATED FUEL is being handled in the secondary containment, or during operations with a potential for draining the reactor vessel, with the vessel head removed and fuel in the vessel.

ACTION:

Without REFUELING AREA SECONDARY CONTAINMENT INTEGRITY, suspend handling of RECENTLY IRRADIATED FUEL in the secondary containment, and operations with a potential for draining the reactor vessel. The provisions of Specifica-tion 3.0.3 are not applicable.

SURVEILLANCE REOUIREMENTS 4.6.5.1.2 REFUELING AREA SECONDARY CONTAINMENT INTEGRITY shall be demonstrated by:

a. Verifying at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> that the pressure within the refueling area secondary containment is greater than or equal to 0.25 inch of vacuum water gauge.

b. Verifying at least once per 31 days that:

1. All refueling area secondary containment equipment hatches and blowout panels are closed and sealed.

2. At least one door in each access to the refueling area secondary containment is closed.

3. All refueling area secondary containment penetrations not capable of being closed by OPERABLE secondary containment automatic iso-lation dampers/valves and required to be closed during accident conditions are closed by valves, blind flanges, slide gate dampers or deactivated automatic dampers/valves secured in position.

c. At least once per 24 months:

Operating one standby gas treatment subsystem for one hour and main-taining greater than or equal to 0.25 inch of vacuum water gauge in the refueling area secondary containment at a flow rate not exceeding 764 cfm.

LIMERICK - UNIT 2 3/4 6-47 Amendment No. .24,

CONTAINMENT SYSTEMS REFUELING AREA SECONDARY CONTAINMENT AUTOMATIC ISOLATION VALVES LIMITING CONDITION FOR OPERATION 3.6.5.2.2 The refueling area secondary containment ventilation system automatic isolation valves shall be OPERABLE.

APPLICABILITY: When RECENTLY IRRADIATED FUEL is being handled in the secondary containment, or during operations with a potential for draining the reactor vessel, with the vessel head removed and fuel in the vessel.

ACTION:

With one or more of the refueling area secondary containment ventilation system automatic isolation valves inoperable, maintain at least one isolation valve OPERABLE in each affected penetration that is open and within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> either:

a. Restore the inoperable valves to OPERABLE status, or

b. Isolate each affected penetration by use of at least one deactivated valve secured in the isolation position, or

c. Isolate each affected penetration by use of at least one closed manual valve, blind flange or slide gate damper.

Otherwise, suspend handling of RECENTLY IRRADIATED FUEL in the refueling area secondary containment, and operations with a potential for draining the reactor vessel. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REOUIREMENTS 4.6.5.2.2 Each refueling area secondary containment ventilation system auto-matic isolation valve shall be demonstrated OPERABLE:

a. Prior to returning the valve to service after maintenance, repair or replacement work is performed on the valve or its associated actuator, control or power circuit by cycling the valve through at least one complete cycle of full travel and verifying the specified isolation time.

b. At least once per 24 months by verifying that on a containment isolation test signal each isolation valve actuates to its isolation position.

c. By verifying the isolation time to be within its limit at least once per 92 days.

LIMERICK - UNIT 2 3/4 6 -50 Amendment No. 34, .9,

CONTAINMENT SYSTEMS STANDBY GAS TREATMENT SYSTEM - COMMON SYSTEM LIMITING CONDITION FOR OPERATION 3.6.5.3 Two independent standby gas treatment subsystems shall be OPERABLE.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3, and when RECENTLY IRRADIATED FUEL is being handled in the secondary containment, or during operations with a potential for draining the reactor vessel , with the vessel head removed and fuel in the vessel.

ACTION:

a. In OPERATIONAL CONDITION 1, 2, or 3:

1. With the Unit 1 diesel generator for one standby gas treatment subsystem inoperable for more than 30 days, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The provisions of Specification 3.0.4 are not applicable.

2. With one standby gas treatment subsystem inoperable, restore the inoperable subsystem to OPERABLE status within 7 days, or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

3. With one standby gas treatment subsystem inoperable and the other standby gas treatment subsystem with an inoperable Unit 1 diesel generator, restore the inoperable subsystem to OPERABLE status or restore the inoperable Unit 1 diesel generator to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

4. With the Unit 1 diesel generators for both standby gas treatment system subsystems inoperable for more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

b. When RECENTLY IRRADIATED FUEL is being handled in the secondary containment, or during operations with a potential for draining the reactor vessel, with the vessel head removed and fuel in the vessel:

1. With one standby gas treatment subsystem inoperable, restore the inoperable subsystem to OPERABLE status within 7 days, or suspend handling of RECENTLY IRRADIATED FUEL in the secondary containment, and operations with a potential for draining the reactor vessel. The provisions of Specification 3.0.3 are not applicable.

2. With both standby gas treatment subsystems inoperable, suspend handling of RECENTLY IRRADIATED FUEL in the secondary containment, and operations with a potential for draining the reactor vessel. The provisions of Specification 3.0.3 are not applicable.

LIMERICK - UNIT 2 3/4 6 -52 Amendment No.

CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

b. At least once per 24* months or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire, or chemical release in any ventilation zone communicating with the subsystem by:

1. Verifying that the subsystem satisfies the in-place penetration and bypass leakage testing acceptance criteria of less than 0.05%

and uses the test procedure guidance in Regulatory Positions C.5.a, C.5.c and C.5.d of Regulatory Guide 1.52, Revision 2, March 1978, and the system flow rate is 5764 cfm +/- 10%.

2. Verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, shows the methyl iodide penetration of less than 1.25%

when tested in accordance with ASTM D3803-1989 at a temperature of 300C (860F), at a relative humidity of 70% and at a face velocity of 66 fpm.

3. Verify that when the fan is running the subsystem flowrate is 2800 cfm minimum from each reactor enclosure (Zones I and II) and 2200 cfm minimum from the refueling area (Zone III) when tested in accordance with ANSI N510-1980.

4. Verify that the pressure drop across the refueling area to SGTS prefilter is less than 0.25 inches water gage while operating at a flow rate of 2400 cfm +/- 10%.

c. After every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of charcoal adsorber operation by verifying within 31 days after removal that a laboratory analysis of a repre-sentative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, shows the methyl iodide penetration of less than 1.25% when tested in accordance with ASTM D3803-1989 at a temperature of 30'C (860 F),

at a relative humidity of 70% and at a face velocity of 66 fpm.

d. At least once per 24 months by:

1. Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than 9.1 inches water gauge while operating the filter train at a flow rate of 8400 cfm +/- 10%.

• Surveillance interval is an exception to the guidance provided in Regulatory Guide 1.52, Revision 2, March 1978.

LIMERICK - UNIT 2 .3/4 6-53 Amendment No. 34, 86, 4Ad6,

CONTAINMENT SYSTEMS REACTOR ENCLOSURE RECIRCULATION SYSTEM LIMITING CONDITION FOR OPERATION 3.6.5.4 Two independent reactor enclosure recirculation subsystems shall be OPERABLE.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3.

ACTION:

a. With one reactor enclosure recirculation subsystem inoperable, restore the inoperable subsystem to OPERABLE status within 7 days, or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

b. With both reactor enclosure recirculation subsystems inoperable, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.6.5.4 Each reactor enclosure recirculation subsystem shall be demonstrated OPERABLE:

a. At least once per 31 days by initiating, from the control room, flow through the HEPA filters and charcoal adsorbers and verifying that the subsystem operates properly (flow at a minimum of 30,000 cfm).

b. At least once per 24* months or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire, or chemical release in any ventilation zone communi-cating with the subsystem by:

1. Verifying that the subsystem satisfies the in-place penetration and bypass leakage testing acceptance criteria of less than 0.05%

and uses the test procedure guidance in Regulatory Positions C.5.a, C.5.c, and C.5.d of Regulatory Guide 1.52, Revision 2, March 1978, at rated flow (60,000 cfm +/- 10%).

2. Verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, shows the methyl iodide penetration of less than 15%

when tested in accordance with ASTM D3803-1989 at a temperature of 30'C (860F) and a relative humidity of 70%.

3. Verifying a subsystem flow rate within a range of 30,000 cfm to 66,000 cfm during system operation when tested in accordance with ANSI N510-1980.

• Surveillance interval is an exception to the guidance provided in Regulatory Guide 1.52 Revision 2, March 1978.

LIMERICK - UNIT 2 3/4 6-55 Amendment No. 34, 4d6,

CONTAINMENT SYSTEMS SURVEILLANCE REOUIREMENTS (Continued)

c. After every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of charcoal adsorber operation by verifying within 31 days after removal that a laboratory analysis of a repre-sentative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, shows the methyl iodide penetration of less than 15% when tested in accordance with ASTM D3803-1989 at a temperature of 30'C (860 F) an a relative humidity of 70%.

d. At least once per 24 months by:

1. Verifying that the pressure drop across the combined prefilter, upstream and downstream HEPA filters, and charcoal adsorber banks is less than 6 inches water gauge while operating the filter train at rated flow (60,000 cfm +/- 10%), verifying that the prefilter pressure drop is less than 0.8 inch water gauge and that the pressure drop across each HEPA is less than 2 inches water gauge.

2. Verifying that the filter train starts and the isolation valves which take suction on and return to the reactor enclosure open on each of the following test signals:

a. Manual initiation from the control room, and

b. Simulated automatic initiation signal.

e. After each complete or partial replacement of a HEPA filter bank by verifying that the HEPA filter bank satisfies the inplace penetration and leakage testing acceptance criteria of less than 0.05% in accordance with ANSI N510-1980 while operating the system at rated flow (60,000 cfm +/- 10%).

f. After each complete or partial replacement of a charcoal adsorber bank by verifying that the charcoal adsorber bank satisfies the inplace penetration and leakage testing acceptance criteria of less than 0.05% in accordance with ANSI N510-1980 for a halogenated hydro-carbon refrigerant test gas while operating the system at rated flow (60,000 cfm +/- 10%).

LIMERICK - UNIT 2 3/4 6-56 Amendment No. 34, 10-6,

PLANT SYSTEMS EMERGENCY SERVICE WATER SYSTEM - COMMON SYSTEM LIMITING CONDITION FOR OPERATION 3.7.1.2 At least the following independent emergency service water system loops, with each loop comprised of:

a. Two OPERABLE emergency service water pumps, and

b. An OPERABLE flow path capable of taking suction from the emergency service water pumps wet pits which are supplied from the spray pond or the cooling tower basin and transferring the water to the associated Unit 2 and common safety-related equipment, shall be OPERABLE:

a. Two loops, in OPERATIONAL CONDITIONS 1, 2, and 3

b. One Loop, in OPERATIONAL CONDITIONS 4, 5, and when RECENTLY IRRADIATED FUEL is being handled in the secondary containment, or during operations with a potential for draining the reactor vessel APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3, 4, 5, and when RECENTLY IRRADIATED FUEL is being handled in the secondary containment, or during operations with a potential for draining the reactor vessel.

ACTION:

a. In OPERATION CONDITION 1, 2, or 3:

1. With one emergency service water pump inoperable, restore the inoperable pump to OPERABLE status within 45 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

2. With one emergency service water pump in each loop inoperable, restore at least one inoperable pump to OPERABLE status within 30 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

3. With one emergency service water system loop otherwise inoperable, declare all equipment aligned to the inoperable loop inoperable**, restore the inoperable loop to OPERABLE status with at least one OPERABLE pump within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

• • The diesel generators may be aligned to the OPERABLE emergency service water system loop provided confirmatory flow testing has been performed. Those diesel generators not aligned to the OPERABLE emergency service water system loop shall be declared inoperable and the actions of 3.8.1.1 taken.

LIMERICK - UNIT 2 3/4 7-3 Amendment No. -8, 0, 9X,

PLANT SYSTEMS LIMITING CONDITION FOR OPERATION (Continued)

ACTION: (Continued)

4. With three ESW pump/diesel generator pairs** inoperable, restore at least one inoperable ESW pump/diesel generator pair** to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

5. With four ESW pump/diesel generator pairs** inoperable, restore at least one inoperable ESW pump/diesel generator pair** to OPERABLE status within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

b. In OPERATIONAL CONDITION 4 or 5:

1. With only one emergency service water pump and its associated flow path OPERABLE, restore at least two pumps with at least one flow path to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or declare the associated safety related equipment inoperable and take the ACTION required by Specifications 3.5.2 and 3.8.1.2.

c. When RECENTLY IRRADIATED FUEL is being handled in the secondary containment, or during operations with a potential for draining the reactor vessel:

1. With only one emergency service water pump and its associated flow path OPERABLE, restore at least two pumps with at least one flow path to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or verify adequate cooling remains available for the diesel generators required to be OPERABLE or declare the associated diesel genera-tor(s) inoperable and take the ACTION required by Specifica-tion 3.8.1.2. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REOUIREMENT 4.7.1.2 At least the above required emergency service water system loop(s) shall be demonstrated OPERABLE:

a. At least once per 31 days by verifying that each valve (manual, power-operated, or automatic) that is not locked, sealed, or otherwise secured in position, is in its correct position.

b. At least once per 24 months by verifying that:

1. Each automatic valve actuates to its correct position on its appropriate ESW pump start signal.

2. Each pump starts automatically when its associated diesel generator starts.

• • An ESW pump/diesel generator pair consists of an ESW pump and its associated diesel generator. If either an ESW pump or its associated diesel generator becomes inoperable, than the ESW pump/diesel generator pair is inoperable.

LIMERICK - UNIT 2 3/4 7-4 Amendment No. 34,

PLANT SYSTEMS ULTIMATE HEAT SINK LIMIIING CONDITION FOR OPERATION 3.7.1.3 The spray pond shall be OPERABLE with:

a. A minimum pond water level at or above elevation 250'-10" Mean Sea Level, and

b. A pond water temperature of less than or equal to 880 F.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3, 4, 5, and when RECENTLY IRRADIATED FUEL is being handled in the secondary containment, or during operations with a potential for draining the reactor vessel.

ACTION:

With the requirements of the above specification not satisfied:

a. In OPERATIONAL CONDITION 1, 2, or 3, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

b. In OPERATIONAL CONDITION 4 or 5, declare the RHRSW system and the emergency service water system inoperable and take the ACTION required by Specifications 3.7.1.1 and 3.7.1.2.

c. When RECENTLY IRRADIATED FUEL is being handled in the secondary containment, or during operations with a potential for draining the reactor vessel, declare the emergency service water system inoperable and take the ACTION required by Specification 3.7.1.2.

The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.7.1.3 The spray pond shall be determined OPERABLE:

a. By verifying the pond water level to be greater than its limit at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

b. By verifying the water surface temperature (within the upper two feet of the surface) to be less than or equal to 880F:

1. at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> when the spray pond temperature is greater than or equal to 80'F; and

2. at least once per 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> when the spray pond temperature is greater than or equal to 850F; and

3. at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when the spray pond temperature is greater than 320F.

c. By verifying all piping above the frost line is drained:

1. within one (1) hour after being used when ambient air temperature is below 40'F; or

2. when ambient air temperature falls below 40'F if the piping has not been previously drained.

LIMERICK - UNIT 2 3/4 7-5 Amendment No. -54,

PLANT SYSTEMS 3/4.7.2 CONTROL ROOM EMERGENCY FRESH AIR SUPPLY SYSTEM - COMMON SYSTEM LIMITING CONDITION FOR OPERATION 3.7.2 Two independent control room emergency fresh air supply system subsystems shall be OPERABLE.

APPLICABILITY: All OPERATIONAL CONDITIONS and when RECENTLY IRRADIATED FUEL is being handled in the secondary containment, or during operations with a potential for draining the reactor vessel.

ACTION:

a. In OPERATIONAL CONDITION 1, 2, or 3:

1. With the Unit 1 diesel generator for one control room emergency fresh air supply subsystem inoperable for more than 30 days, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The provisions of Specification 3.0.4 are not applicable.

2. With one control room emergency fresh air supply subsystem inoperable, restore the inoperable subsystem to OPERABLE status within 7 days, or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

3. With one control room emergency fresh air supply subsystem inoperable and the other control room emergency fresh air supply subsystem with an inoperable Unit 1 diesel generator, restore the inoperable subsystem to OPERABLE status or restore the Unit 1 diesel generator to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

4. With the Unit 1 diesel generators for both control room emergency fresh air supply subsystems inoperable for more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

b. In OPERATIONAL CONDITION 4, 5 or when RECENTLY IRRADIATED FUEL is being handled in the secondary containment, or during operations with a potential for draining the reactor vessel:

1. With one control room emergency fresh air supply subsystem inoperable, restore the inoperable subsystem to OPERABLE status within 7 days, or initiate and maintain operation of the OPERABLE subsystem in the radiation isolation mode of operation.

2. With both both control room emergency fresh air supply subsystem inoperable, suspend handling of RECENTLY IRRADIATED FUEL in the secondary containment and operations with a potential for draining the reactor vessel. The provisions of Specification 3.0.3 are not applicable.

LIMERICK - UNIT 2 3/4 7-6 Amendment No.

PLANT SYSTEMS SURVEILLANCE REOUIREMENTS (Continued)

2. Verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, shows the methyl iodide penetration of less than 10%

when tested in accordance with ASTM D3803-1989 at a temperature of 30'C (860F) and a relative humidity of 70%.

3. Verifying a subsystem flow rate of 3000 cfm +/- 10% during subsystem operation when tested in accordance with ANSI N510-1980.

d. After every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of charcoal adsorber operation by verifying within 31 days after removal that a laboratory analysis of a repre-sentative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, shows the methyl iodide penetration of less than 10% when tested in accordance with ASTM D3803-1989 at a temperature of 300C (860F) and a relative humidity of 70%.

e. At least once per 24 months by:

1. Verifying that the pressure drop across the combined prefilter, upstream and downstream HEPA filters, and charcoal adsorber banks is less than 6 inches water gauge while operating the subsystem at a flow rate of 3000 cfm +/- 10%; verifying that the prefilter pressure drop is less than 0.8 inch water gauge and that the pressure drop across each HEPA is less than 2 inches water gauge.

2. Verifying that on each of the below chlorine isolation mode actuation test signals, the subsystem automatically switches to the chlorine isolation mode of operation and the isolation valves close within 5 seconds:

a) Outside air intake high chlorine, and b) Manual initiation from the control room.

3. Verifying that on manual initiation from the control room, the subsystem switches to the radiation isolation mode of operation and the control room is maintained at a positive pressure of at least 1/8 inch water gauge relative to the turbine enclosure and auxiliary equipment room and outside atmosphere during subsystem operation with an outdoor air flow rate less than or equal to 525 cfm.

LIMERICK - UNIT 2 3/4 7-7 Amendment No. 34, 416,

ELECTRICAL POWER SYSTEMS A.C. SOURCES - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.8.1.2 As a minimum, the following A.C. electrical power sources shall be OPERABLE:

a. One circuit between the offsite transmission network and the onsite Class lE distribution system, and

b. Two diesel generators each with:

1. A day fuel tank containing a minimum of 200 gallons of fuel.

2. A fuel storage system containing a minimum of 33,500 gallons of fuel.

3. A fuel transfer pump.

APPLICABILITY: OPERATIONAL CONDITIONS 4, 5, and when RECENTLY IRRADIATED FUEL is being handled in the secondary containment, or during operations with a potential for draining the reactor vessel.

ACTION:

a. With less than the above required A.C. electrical power sources OPERABLE, suspend CORE ALTERATIONS, handling of RECENTLY IRRADIATED FUEL in the secondary containment, operations with a potential for draining the reactor vessel and crane operations over the spent fuel storage pool when fuel assemblies are stored therein. In addition, when in OPERATIONAL CONDITION 5 with the water level less than 22 feet above the reactor pressure vessel flange, immediately initiate corrective action to restore the required power sources to OPERABLE status as soon as practical.

b. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REOUIREMENTS 4.8.1.2 At least the above required A.C. electrical power sources shall be demonstrated OPERABLE per Surveillance Requirements 4.8.1.1.1, 4.8.1.1.2, and 4.8.1.1.3.

LIMERICK - UNIT 2 3.4 8 -9 Amendment No.

ELECTRICAL POWER SYSTEMS DnC. SOuRCES - SH[UTDOWN LIMITING CONDITION FOR OPERATION 3.8.2.2 As a minimum, two of the following four divisions of the D.C.

electrical power sources system shall be OPERABLE with:

a. Division 1, Consisting of:

1. 125-Volt Battery 2A1 (2A1D101).

2. 125-Volt Battery 2A2 (2A2D101).

3. 125-Volt Battery Charger 2BCA1 (2AID103).

4. 125-Volt Battery Charger 2BCA2 (2A2D103).

b. Division 2, Consisting of:

1. 125-Volt Battery 2B1 (2B1D101).

2. 125-Volt Battery 2B2 (2B2D101).

3. 125-Volt Battery Charger 2BCB1 (2B1D103).

4. 125-Volt Battery Charger 2BCB2 (2B2D103).

c. Division 3, Consisting of:

1. 125-Volt Battery 2C (2CD101).

2. 125-Volt Battery Charger 2BCC (2CD103).

d. Division 4, Consisting of:

1. 125-Volt Battery 2D (2DD101).

2. 125-Volt Battery Charger 2BCD (2DD103).

APPLICABILITY: OPERATIONAL CONDITIONS 4, 5, and when RECENTLY IRRADIATED FUEL is being handled in the secondary containment, or during operations with a potential for draining the reactor vessel.

ACTION:

a. With one or two required battery chargers on one required division inoperable:

1. Restore battery terminal voltage to greater than or equal to the minimum established float voltage within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />,

2. Verify associated Division 1 or 2 float current < 2 amps, or Division 3 or 4 float current < 1 amp within 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br /> and once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter, and

3. Restore battery charger(s) to OPERABLE status within 7 days.

b. With one or more required batteries inoperable due to:

1. One or two batteries on one division with one or more battery cells float voltage < 2.07 volts, perform 4.8.2.1.a.1 and 4.8.2.1.a.2 within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for affected battery(s) and restore affected cell(s) voltage > 2.07 volts within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

LIMERICK - UNIT 2 3/4 8-14 Amendment No. 1-2-6,

ELECTRICAL POWER SYSTEMS LIMITING CONDITION FOR OPERATION ACTION: (Continued)

2. Division 1 or 2 with float current > 2 amps, or with Division 3 or 4 with float current > 1 amp, perform 4.8.2.1.a.2 within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for affected battery(s) and restore battery float current to within limits within 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br />.

3. One or two batteries on one division with one or more cells electrolyte level less than minimum established design limits, if electrolyte level was below the top of the plates restore electrolyte level to above top of plates within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and verify no evidence of leakage(*) within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. In all cases, restore electrolyte level to greater than or equal to minimum established design limits within 31 days.

4. One or two batteries on one division with pilot cell electrolyte temperature less than minimum established design limits, restore battery pilot cell temperature to greater than or equal to minimum established design limits within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

5. Batteries in more than one division affected, restore battery parameters for all batteries in one division to within limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

6. (i) Any battery having both (Action b.1) one or more battery cells float voltage < 2.07 volts and (Action b.2) float current not within limits, and/or (ii) Any battery not meeting any Action b.1 through b.5, Restore the battery parameters to within limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

c. 1. With the requirements of Action a. and/or Action b. not met, or

2. With less than two divisions of the above required D.C. electrical power sources OPERABLE for reasons other than Actions a. and/or b.,

Suspend CORE ALTERATIONS, handling of RECENTLY IRRADIATED FUEL in the secondary containment and operations with a potential for draining the reactor vessel.

d. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REOUIREMENTS 4.8.2.2 At least the above required batteries and chargers shall be demonstrated OPERABLE per Surveillance Requirement 4.8.2.1.

(*) Contrary to the provisions of Specification 3.0.2, if electrolyte level was below the top of the plates, the verification that there is no evidence of leakage is required to be completed regardless of when electrolyte level is restored.

LIMERICK - UNIT 2 3/4 8-14a LTAmendment No. 4-1-6,

ELECTRICAL POWER SYSTEMS LIMITING CONDITION FOR OPERATION (Continued) c) 125-V DC Distribution Panels: 2PPA1 (2AD102) 2PPA2 (2AD501) 2PPA3 (2AD162)

2. Unit 2 Division 2, Consisting of:

a) 250-V DC Fuse Box: 2FB (2BD105) b) 250-V DC Motor Control Centers: 2DB-1 (20D202) 2DB-2 (20D203) c) 125-V DC Distribution Panels: 2PPB1 (2BD102) 2PPB2 (2BD501) 2PPB3 (2BD162)

3. Unit 2 Division 3, Consisting of:

a) 125-V DC Fuse Box: 2FC (2CD105) b) 125-V DC Distribution Panels: 2PPC1 (2CD102) 2PPC2 (2CD501) 2PPC3 (2CD162)

4. Unit 2 Division 4, Consisting of:

a) 125-V DC Fuse Box: 2FD (2DD105) b) 125-V DC Distribution Panels: 2PPD1 (2DD102) 2PPD2 (2DD501) 2PPD3 (2DD162)

5. Unit 1 and Common Division 1, Consisting of:

a) 250-V DC Fuse Box: 1FA (lAD105) b) 125-V DC Distribution Panels: 1PPA1 (lAD102) 1PPA2 (1AD501)

6. Unit 1 and Common Division 2, Consisting of:

a) 250-V DC Fuse Box: 1FB (IBD105) b) 125-V DC Distribution Panels: lPPB1 (lBD102) lPPB2 (1BD501)

7. Unit 1 and Common Division 3, Consisting of:

a) 125-V DC Fuse Box: 1FC (ClC105) b) 125-V DC Distribution Panels: 1PPC1 (lCD102) lPPC2 (lCD501)

8. Unit 1 and Common Division 4, Consisting of:

a) 125-V DC Fuse Box: 1FD (CDD105) b) 125-V DC Distribution Panels: 1PPD1 (1DD1O2) 1PPD2 (1DD501)

APPLICABILITY: OPERATIONAL CONDITIONS 4, 5, and when RECENTLY IRRADIATED FUEL is being handled in the secondary containment, or during operations with a potential for draining the reactor vessel.

ACTION:

a. With less than two divisions of the above required Unit 2 A.C.

distribution systems energized, suspend CORE ALTERATIONS, handling of RECENTLY IRRADIATED FUEL in the secondary containment and operations with a potential for draining the reactor vessel.

LIMERICK - UNIT 2 3/4 8-19 Amendment No. 4-210,

ELECTRICAL POWER SYSTEMS LIMITING CONDITION FOR OPERATION (Continued)

ACTION: (Continued)

b. With less than two divisions of the above required Unit 2 D.C.

distribution systems energized, suspend CORE ALTERATIONS, handling of RECENTLY IRRADIATED FUEL in the secondary containment and operations with a potential for draining the reactor vessel.

c. With any of the above required Unit 1 and common AC and/or DC distribution system divisions not energized, declare the associated common equipment inoperable, and take the appropriate ACTION for that system.

d. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.8.3.2 At least the above required power distribution system divisions shall be determined energized at least once per 7 days by verifying correct breaker alignment and voltage on the busses/MCCs/panels.

LIMERICK - UNIT 2 3/4 8-20 Amendment No.

ATTACHMENT 5 LIMERICK GENERATING STATION UNITS 1 AND 2 Docket Nos. 50-352 & 50-353 License Nos. NPF-39 & NPF-85 License Amendment Request

'LGS Alternative Source Term Implementation" Retyped Technical Specification Bases Pages (For information only)

UNITS 1 B 3/4 1-4 B 3/4 1-5 B 3/4 4-6 B 3/4 6-5 B 3/4 7-la UNITS 2 B 3/4 1-4 B 3/4 1-5 B 3/4 4-6 B 3/4 6-5 B 3/4 7-1

RFACTTVTTY CONTROL SYSTEMS BASES 3/4.1.5 STANDBY LIQUID CONTROL SYSTEM The standby liquid control system provides a backup capability for bringing the reactor from full power to a cold, Xenon-free shutdown, assuming that the withdrawn control rods remain fixed in the rated power pattern. To meet this objective it is necessary to inject a quantity of boron which produces a concen-tration of 660 ppm in the reactor core and other piping systems connected to the reactor vessel. To allow for potential leakage and improper mixing, this concentration is increased by 25%. The required concentration is achieved by having available a minimum quantity of 3,160 gallons of sodium pentaborate solution containng a minimum of 3,754 lbs of sodium pentaborate having the requisite Boron-10 atom % enrichment of 29% as determined from Reference 5.

This quantity of solution is a net amount which is above the pump suction shutoff level setpoint thus allowing for the portion which cannot be injected.

The above quantities calculated at 29% Boron-10 enrichment have been demonstrated by analysis to provide a Boron-10 weight equivalent of 185 lbs in the sodium pentaborate solution. Maintaining this Boron-10 weight in the net tank contents ensures a sufficient quantity of boron to bring the reactor to a cold, Xenon-free shutdown.

The pumping rate of 41.2 gpm provides a negative reactivity insertion rate over the permissible solution volume range, which adequately compensates for the positive reactivity effects due to elimination of steam voids, increased water density from hot to cold, reduced doppler effect in uranium, reduced neutron leakage from boiling to cold, decreased control rod worth as the moderator cools, and xenon decay. The temperature requirement ensures that the sodium pentaborate always remains in solution.

With redundant pumps and explosive injection valves and with a highly reliable control rod scram system, operation of the reactor is permitted to continue for short periods of time with the system inoperable or for longer periods of time with one of the redundant components inoperable.

The SLCS system consists of three separate and independent pumps and explosive valves. Two of the separate and independent pumps and explosive valves are required to meet the minimum requirements of this technical specification and, where applicable, satisfy the single failure criterion.

The SLCS must have an equivalent control capacity of 86 gpm of 13% weight sodium pentaborate in order to satisfy 10 CFR 50.62 (Requirements for reduction of risk from anticipated transients without scram (ATWS) events for light-water-cooled nuclear power plants). As part of the ARTS/MELLL program the ATWS analysis was updated to reflect the new rod line. As a result of this it was determined that the Boron 10 enrichment was required to be increased to 29% to prevent exceeding a suppression pool temperature of 1901F. This equivalency requirement is fulfilled by having a system which satisfies the equation given in 4.1.5.b.2.

The upper limit concentration of 13.8% has been established as a reasonable limit to prevent precipitation of sodium pentaborate in the event of a loss of tank heating, which allow the solution to cool.

LIMERICK - UNIT 1 B 3/4 1-4 Amendment No. 21, 66,

REACTIVITY CONTROL SYSTEMS BASES STANDBY LIGIJID CONTROL SYSTEM (Continued)

Surveillance requirements are established on a frequency that assures a high reliability of the system. Once the solution is established, boron con-centration will not vary unless more boron or water is added, thus a check on the temperature and volume once each 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> assures that the solution is available for use.

Replacement of the explosive charges in the valves at regular intervals will assure that these valves will not fail because of deterioration of the charges.

The Standby Liquid Control System also has a post-DBA LOCA safety function to buffer Suppression Pool pH in order to maintain bulk pH above 7.0. The buffering of Suppression Pool pH is necessary to prevent iodine re-evolution to satisfy the methodology for Alternative Source Term. Manual initiation is used, and the minimum amount of total boron required for Suppression Pool pH buffering is 240 lbs. Given that at least 185 lbs of Boron-10 is maintained in the tank, the total boron in the tank will be greater than 240 lbs for the range of enrichments from 29% to 62%.

ACTION Statement (a) applies only to OPERATIONAL CONDITIONS 1 and 2 because a single pump can satisfy both the reactor control function and the post-DBA LOCA function to control Suppression Pool pH since boron injection is not required until 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br /> post-LOCA. ACTION Statement (b) applies to OPERATIONAL CONDITIONS 1, 2 and 3 to address the post-LOCA safety function of the SLC system.

1. C. J. Paone, R. C. Stirn and J. A. Woolley, "Rod Drop Accident Analysis for Large BWR's," G. E. Topical Report NEDO-10527, March 1972.

2. C. J. Paone, R. C. Stirn, and R. M. Young, Supplement 1 to NEDO-10527, July 1972.

3. J. M. Haun, C. J. Paone, and R. C. Stirn, Addendum 2, "Exposed Cores."

Supplement 2 to NEDO-10527, January 1973.

4. Amendment 17 to General Electric Licensing Topical Report NEDE-24011-P-A, "General Electric Standard Application for Reactor Fuel".

5. "Maximum Extended Load Line Limit and ARTS Improvement Program Analyses for Limerick Generating Station Units 1 and 2," NEDC-32193P, Revision 2, October 1993.

LIMERICK - UNIT 1 B 3/4 J-5 Amendment No. 24, 66,

REACTOR COOLANT SYSTEM BASES 3/4.4.7 MAIN STEAM LINE ISOLATION VALVES Double isolation valves are provided on each of the main steam lines to minimize the potential leakage paths from the containment in case of a line break. Only one valve in each line is required to maintain the integrity of the containment, however, single failure considerations require that two valves be OPERABLE. The surveillance requirements are based on the operating history of this type valve. The maximum closure time has been selected to contain fission products and to prevent core damage following line breaks. The minimum closure time is consistent with the assumptions in the safety analyses to prevent pressure surges.

3/4.4.8 STRUCTURAL INTEGRITY The inspection programs for ASME Code Class 1, 2, and 3 components ensure that the structural integrity of these components will be maintained at an acceptable level throughout the life of the plant.

Components of the reactor coolant system were designed to provide access to permit inservice inspections in accordance with Section XI of the ASME Boiler and Pressure Vessel Code 1971 Edition and Addenda through Winter 1972.

The inservice inspection program for ASME Code Class 1, 2, and 3 components will be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable addenda as required by 10 CFR 50.55a. Additionally, the Inservice Inspection Program conforms to the NRC staff positions identified in NRC Generic Letter 88-01, "NRC Position on IGSCC in BWR Austenitic Stainless Steel Piping,"

as approved in NRC Safety Evaluations dated March 6, 1990 and October 22, 1990.

3/4.4.9 RESIDUAL HEAT REMOVAL The RHR system is required to remove decay heat and sensible heat in order to maintain the temperature of the reactor coolant. RHR shutdown cooling is comprised of four (4) subsystems which make two (2) loops. Each loop consists of two C2) motor driven pumps, a heat exchanger, and associated piping and valves. Both loops have a common suction from the same recirculation loop. Two (2) redundant, manually controlled shutdown cooling subsystems of the RHR System can provide the required decay heat removal capability. Each pump discharges the reactor coolant, after it has been cooled by circulation through the respective heat exchangers, to the reactor via the associated recirculation loop or to the reactor via the low pressure coolant injection pathway. The RHR heat exchangers transfer heat to the RHR Service Water System. The RHR shutdown cooling mode is manually controlled.

An OPERABLE RHR shutdown cooling subsystem consists of an RHR pump, a heat exchanger, valves, piping, instruments, and controls to ensure an OPERABLE flow path.

In HOT SHUTDOWN condition, the requirement to maintain OPERABLE two (2) independent RHR shutdown cooling subsystems means that each subsystem considered OPERABLE must be associated with a different heat exhanger loop, i.e., the "A" RHR heat exchanger with the "A" RHR pump or the "C" RHR pump, and the "B" RHR heat exchanger with the "B" RHR pump or the "D" RHR pump are two (2) independent RHR shutdown cooling subsystems. Only one (1) of the two (2) RHR pumps associated with each RHR heat exchanger loop is LIMERICK - UNIT 1 B 3/4 4-6 Amendment No. 49, 9-7, 4-do, 42-5,

CONTAINMENT SYSTEMS BASES 3/4.6.5 SECONDARY CONTAINMENT Secondary containment is designed to minimize any ground level release of radioactive material which may result from an accident. The Reactor Enclosure and associated structures provide secondary containment during normal operation when the drywell is sealed and in service. At other times the drywell may be open and, when required, secondary containment integrity is specified.

Establishing and maintaining a vacuum in the reactor enclosure secondary containment with the standby gas treatment system once per 24 months, along with the surveillance of the doors, hatches, dampers and valves, is adequate to ensure that there are no violations of the integrity of the secondary containment.

The OPERABILITY of the reactor enclosure recirculation system and the standby gas treatment systems ensures that sufficient iodine removal capability will be available in the event of a LOCA or refueling accident involving RECENTLY IRRADIATED FUEL (SGTS only). The reduction in containment iodine inventory reduces the resulting SITE BOUNDARY radiation doses associated with containment leakage.

The operation of this system and resultant iodine removal capacity are consistent with the assumptions used in the LOCA and refueling accident analyses. Provisions have been made to continuously purge the filter plenums with instrument air when the filters are not in use to prevent buildup of moisture on the adsorbers and the HEPA filters.

Although the safety analyses assumes that the reactor enclosure secondary containment draw down time will take 930 seconds, these surveillance require-ments specify a draw down time of 916 seconds. This 14 second difference is due to the diesel generator starting and sequence loading delays which is not part of this surveillance requirement.

The reactor enclosure secondary containment draw down time analyses assumes a starting point of 0.25 inch of vacuum water gauge and worst case SGTS dirty filter flow rate of 2800 cfm. The surveillance requirements satisfy this as-sumption by starting the drawdown from ambient conditions and connecting the adjacent reactor enclosure and refueling area to the SGTS to split the exhaust flow between the three zones and verifying a minimum flow rate of 2800 cfm from the test zone. This simulates the worst case flow alignment and verifies ade-quate flow is available to drawdown the test zone within the required time.

The Technical Specification Surveillance Requirement 4.6.5.3.b.3 is intended to be a multi-zone air balance verification without isolating any test zone.

Based on implementation of the AST methodology, it has been demonstrated through analysis that operating the subsystem with a minimum flow rate of 30,000 cfm satisfies the dose requirements. However, based on the applicable testing standards and guidance provided in Regulatory Guide 1.52, Revision 2, and ANSI N510, 1980, the acceptance criteria must be satisfied based on the system rated flow, which provides the most conservative results. Satisfying the acceptance criteria at rated flow also demonstrates system operability at lower measured flows because the residence time is longer. Therefore, verification of subsystem flow rates between 30,000 cfm and the maximum rated flow of 66,000 cfm (includes a 10% factor) satisfies the surveillance requirements for the HEPA filters and charcoal adsorber housings based on 70% efficiencies.

The SGTS fans are sized for three zones and therefore, when aligned to a single zone or two zones, will have excess capacity to more quickly drawdown the affected zones. There is no maximum flow limit to individual zones or pairs of zones and the air balance and drawdown time are verified when all three zones are connected to the SGTS.

The three zone air balance verification and drawdown test will be done after any major system alteration, which is any modification which will have an effect on the SGTS flowrate such that the ability of the SGTS to drawdown the reactor enclosure to greater than or equal to 0.25 inch of vacuum water gage in less than or equal to 916 seconds could be affected.

LIMERICK - UNIT 1 B 3/4 6-5 Amendment No. 6, 40, X4, 4a6, 4-22,

PLANT SYSTEMS BASES 3/4.7.2 CONTROL ROOM EMERGENCY FRESH AIR SUPPLY SYSTEM - COMMON SYSTEM The OPERABILITY of the control room emergency fresh air supply system ensures that the control room will remain habitable for operations personnel during and following all design basis accident conditions. Constant purge of the system at 1 cfm is sufficient to reduce the buildup of moisture on the adsorbers and HEPA filters. The OPERABILITY of this system in conjunction with control room design provisions is based on limiting the radiation exposure to personnel occupying the control room to 5 rem or less Total Effective Dose Equivalent. This limitation is consistent with the requirements of 10 CFR Part 50.67, Accident Source Terms.

Since the Control Room Emergency Fresh Air Supply System is not credited for filtration in OPERATIONAL CONDITIONS 4 and 5, applicability to 4 and 5 is only required to support the Chlorine and Toxic Gas design basis isolation requirements.

Additionally, based on implementation of the Alternative Source Term methodology, it has been demonstrated through analysis that manual initiation of the CREFAS radiation mode within 30 minutes of the start of gap release for the limiting design basis LOCA is sufficient to assure that control room operator dose limits in 10 CFR Part 50.67 are met.

3/4.7.3 REACTOR CORE ISOLATION COOLING SYSTEM The reactor core isolation cooling (RCIC) system is provided to assure adequate core cooling in the event of reactor isolation from its primary heat sink and the loss of feedwater flow to the reactor vessel without requiring actuation of any of the emergency core cooling system equipment. The RCIC system is conservatively required to be OPERABLE whenever reactor pressure ex-ceeds 150 psig. This pressure is substantially below that for which low pressure core cooling systems can provide adequate core cooling.

The RCIC system specifications are applicable during OPERATIONAL CONDITIONS 1, 2, and 3 when reactor vessel pressure exceeds 150 psig because RCIC is the primary non-ECCS source of emergency core cooling when the reactor is pressurized.

With the RCIC system inoperable, adequate core cooling is assured by the OPERABILITY of the HPCI system and justifies the specified 14 day out-of-service period.

The surveillance requirements provide adequate assurance that RCIC will be OPERABLE when required. Although all active components are testable and full flow can be demonstrated by recirculation during reactor operation, a complete functional test requires reactor shutdown. The pump discharge piping is maintained full to prevent water hammer damage and to start cooling at the earliest possible moment.

LIMERICK - UNIT 1 B 3/4 7-la Amendment No. 2-7, 40,

REACTIVITY CONTROL SYSTEMS BASES 3/4.1.5 STANDBY LIQUID CONTROL SYSTEM The standby liquid control system provides a backup capability for bringing the reactor from full power to a cold, Xenon-free shutdown, assuming that the withdrawn control rods remain fixed in the rated power pattern. To meet this objective it is necessary to inject a quantity of boron which produces a concen-tration of 660 ppm in the reactor core and other piping systems connected to the reactor vessel. To allow for potential leakage and improper mixing, this con-centration is increased by 25%. The required concentration is achieved by having available a minimum quantity of 3,160 gallons of sodium pentaborate solution containng a minimum of 3,754 lbs of sodium pentaborate having the requisite Boron-10 atom % enrichment of 29% as determined from Reference 5.

This quantity of solution is a net amount which is above the pump suction shutoff level setpoint thus allowing for the portion which cannot be injected.

The above quantities calculated at 29% Boron-10 enrichment have been demonstrated by analysis to provide a Boron-10 weight equivalent of 185 lbs in the sodium pentaborate solution. Maintaining this Boron-10 weight in the net tank contents ensures a sufficient quantity of boron to bring the reactor to a cold, Xenon-free shutdown.

The pumping rate of 41.2 gpm provides a negative reactivity insertion rate over the permissible solution volume range, which adequately compensates for the positive reactivity effects due to elimination of steam voids, increased water density from hot to cold, reduced doppler effect in uranium, reduced neutron leakage from boiling to cold, decreased control rod worth as the moderator cools, and xenon decay. The temperature requirement ensures that the sodium pentaborate always remains in solution.

With redundant pumps and explosive injection valves and with a highly reliable control rod scram system, operation of the reactor is permitted to continue for short periods of time with the system inoperable or for longer periods of time with one of the redundant components inoperable.

The SLCS system consists of three separate and independent pumps and explosive valves. Two of the separate and independent pumps and explosive valves are required to meet the minimum requirements of this technical specification and, where applicable, satisfy the single failure criterion.

The SLCS must have an equivalent control capacity of 86 gpm of 13% weight sodium pentaborate in order to satisfy 10 CFR 50.62 (Requirements for reduction of risk from anticipated transients without scram (ATWS) events for light-water-cooled nuclear power plants). As part of the ARTS/MELLL program the ATWS analysis was updated to reflect the new rod line. As a result of this it was determined that the Boron 10 enrichment was required to be increased to 29% to prevent exceeding a suppression pool temperature of 190'F. This equivalency requirement is fulfilled by having a system which satisfies the equation given in 4.1.5.b.2.

The upper limit concentration of 13.8% has been established as a reasonable limit to prevent precipitation of sodium pentaborate in the event of a loss of tank heating, which allow the solution to cool.

LIMERICK - UNIT 2 B 3/4 1-4 Amendment No. 489,

REACTIVITY CONTROL SYSTEMS BASES STANDBY LIQUID CONTROL SYSTEM (Continued)

.Surveillance requirements are established on a frequency that assures a high reliability of the system. Once the solution is established, boron concentration will not vary unless more boron or water is added, thus a check on the temperature and volume once each 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> assures that the solution is available for use.

Replacement of the explosive charges in the valves at regular intervals will assure that these valves will not fail because of deterioration of the charges.

The Standby Liquid Control System also has a post-DBA LOCA safety function to buffer Suppression Pool pH in order to maintain bulk pH above 7.0. The buffering of Suppression Pool pH is necessary to prevent iodine re-evolution to satisfy the methodology for Alternative Source Term. Manual initiation is used, and the minimum amount of total boron required for Suppression Pool pH buffering is 240 lbs. Given that at least 185 lbs of Boron-10 is maintained in the tank, the total boron in the tank will be greater than 240 lbs for the range of enrichments from 29% to 62%.

ACTION Statement (a) applies only to OPERATIONAL CONDITIONS 1 and 2 because a single pump can satisfy both the reactor control function and the post-DBA LOCA function to control Suppression Pool pH since boron injection is not required until 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br /> post-LOCA. ACTION Statement (b) applies to OPERATIONAL CONDITIONS 1, 2 and 3 to address the post-LOCA safety function of the SLC system.

1. C. J. Paone, R. C. Stirn and J. A. Woolley, "Rod Drop Accident Analysis for Large BWR's," G. E. Topical Report NEDO-10527, March 1972.

2. C. J. Paone, R. C. Stirn, and R. M. Young, Supplement 1 to NEDO-10527, July 1972.

3. J. M. Haun, C. J. Paone, and R. C. Stirn, Addendum 2, "Exposed Cores,"

Supplement 2 to NEDO-10527, January 1973.

4. Amendment 17 to General Electric Licensing Topical Report NEDE-24011-P-A, "General Electric Standard Application for Reactor Fuel".

5. "Maximum Extended Load Line Limit and ARTS Improvement Program Analyses for Limerick Generating Station Units 1 and 2," NEDC-32193P, Revision 2, October 1993.

LIMERICK - UNIT 2 B 3/4 1-5 Amendment No. 4.9,

REACTOR COOLANT SYSTEM BASES 3/4.4.7 MAIN STEAM LINE ISOLATION VALVES Double isolation valves are provided on each of the main steam lines to minimize the potential leakage paths from the containment in case of a line break.

Only one valve in each line is required to maintain the integrity of the containment, however, single failure considerations require that two valves be OPERABLE. The surveillance requirements are based on the operating history of this type valve. The maximum closure time has been selected to contain fission products and to prevent core damage following line breaks. The minimum closure time is consistent with the assumptions in the safety analyses to prevent pressure surges.

3/4.4.8 STRUCTURAL INTEGRITY The inspection programs for ASME Code Class 1, 2, and 3 components ensure that the structural integrity of these components will be maintained at an acceptable level throughout the life of the plant.

Components of the reactor coolant system were designed to provide access to permit inservice inspections in accordance with Section XI of the ASME Boiler and Pressure Vessel Code 1971 Edition and Addenda through Winter 1972.

The inservice inspection program for ASME Code Class 1, 2, and 3 components will be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable addenda as required by 10 CFR 50.55a. Additionally, the Inservice Inspection Program conforms to the NRC staff positions identified in NRC Generic Letter 88-01, "NRC Position on IGSCC in BWR Austenitic Stainless Steel Piping,"

as approved in NRC Safety Evaluations dated March 6, 1990 and October 22, 1990.

3/4.4.9 RESIDUAL HEAT REMOVAL The RHR system is required to remove decay heat and sensible heat in order to maintain the temperature of the reactor coolant. RHR shutdown cooling is comprised of four (4) subsystems which make two (2) loops. Each loop consists of two (2) motor driven pumps, a heat exchanger, and associated piping and valves. Both loops have a common suction from the same recirculation loop. Two (2) redundant, manually controlled shutdown cooling subsystems of the RHR System can provide the required decay heat removal capability. Each pump discharges the reactor coolant, after it has been cooled by circulation through the respective heat exchangers, to the reactor via the associated. recirculation loop or to the reactor via the low pressure coolant injection pathway. The RHR heat exchangers transfer heat to the RHR Service Water System. The RHR shutdown cooling mode is manually controlled.

An OPERABLE RHR shutdown cooling subsystem consists of an RHR pump, a heat exchanger, valves, piping, instruments, and controls to ensure an OPERABLE flow path.

In HOT SHUTDOWN condition, the requirement to maintain OPERABLE two (2) independent RHR shutdown cooling subsystems means that each subsystem considered OPERABLE must be associated with a different heat exhanger loop, i.e., the "A" RHR heat exchanger with the "A" RHR pump or the "C" RHR pump, and the "B" RHR heat exchanger with the "B" RHR pump or the "D" RHR pump are two (2) independent RHR shutdown cooling subsystems. Only one (1) of the two (2) RHR pumps associated with each RHR heat exchanger loop is LIMERICK - UNIT 2 B 3/4 4-6 Amendment No. A-t, &I, 821, 8X,

CONTAINMENT SYSTEMS BASES 3/4.6.5 SECONDARY CONTAINMENT Secondary containment is designed to minimize any ground level release of radioactive material which may result from an accident. The Reactor Enclosure and associated structures provide secondary containment during normal operation when the drywell is sealed and in service. At other times the drywell may be open and, when required, secondary containment integrity is specified.

Establishing and maintaining a vacuum in the reactor enclosure secondary containment with the standby gas treatment system once per 24 months, along with the surveillance of the doors, hatches, dampers and valves, is adequate to ensure that there are no violations of the integrity of the secondary containment.

The OPERABILITY of the reactor enclosure recirculation system and the standby gas treatment systems ensures that sufficient iodine removal capability will be available in the event of a LOCA or refueling accident involving RECENTLY IRRADIATED FUEL (SGTS only). The reduction in containment iodine inventory reduces the resulting SITE BOUNDARY radiation doses associated with containment leakage. The operation of this system and resultant iodine removal capacity are consistent with the assumptions used in the LOCA and refueling accident analyses. Provisions have been made to continuously purge the filter plenums with instrument air when the filters are not in use to prevent buildup of moisture on the adsorbers and the HEPA filters.

Although the safety analyses assumes that the reactor enclosure secondary containment draw down time will take 930 seconds, these surveillance require-ments specify a draw down time of 916 seconds. This 14 second difference is due to the diesel generator starting and sequence loading delays which is not part of this surveillance requirement.

The reactor enclosure secondary containment draw down time analyses assumes a starting point of 0.25 inch of vacuum water gauge and worst case SGTS dirty filter flow rate of 2800 cfm. The surveillance requirements satisfy this as-sumption by starting the drawdown from ambient conditions and connecting the adjacent reactor enclosure and refueling area to the SGTS to split the exhaust flow between the three zones and verifying a minimum flow rate of 2800 cfm from the test zone. This simulates the worst case flow alignment and verifies ade-quate flow is available to drawdown the test zone within the required time.

The Technical Specification Surveillance Requirement 4.6.5.3.b.3 is intended to be a multi-zone air balance verification without isolating any test zone.

Based on implementation of the AST methodology, it has been demonstrated through analysis that operating the subsystem with a minimum flow rate of 30,000 cfm satisfies the dose requirements. However, based on the applicable testing standards and guidance provided in Regulatory Guide 1.52, Revision 2, and ANSI N510, 1980, the acceptance criteria must be satisfied based on the system rated flow, which provides the most conservative results. Satisfying the acceptance criteria at rated flow also demonstrates system operability at lower measured flows because the residence time is longer. Therefore, verification of subsystem flow rates between 30,000 cfm and the maximum rated flow of 66,000 cfm (includes a 10% factor) satisfies the surveillance requirements for the HEPA filters and charcoal adsorber housings based on 70% efficiencies.

The SGTS is common to Unit 1 and 2 and consists of two independent subsystems. The power supplies for the common portions of the subsystems are from Unit 1 safeguard busses, therefore the inoperability of these Unit 1 supplies are addressed in the SGTS ACTION statements in order to ensure adequate onsite power sources to SGTS for its Unit 2 function during a loss of offsite power event. The allowable out of service times are consistent with those in the Unit 1 Technical Specifications for SGTS and AC electrical power supply out of service condition combinations.

LIMERICK - UNIT 2 B 3/4 6-5 Amendment No. 34, X51, 86,

3/4.7 PLANT SYSTEMS BASES 3/4.7.1 SERVICE WATER SYSTEMS - COMMON SYSTEMS The OPERABILITY of the service water systems ensures that sufficient cooling capacity is available for continued operation of safety-related equipment during normal and accident conditions. The redundant cooling capacity of these systems, assuming a single failure, is consistent with the assumptions used in the accident conditions within acceptable limits.

The RHRSW and ESW systems are common to Units 1 and 2 and consist of two independent subsystems each with two pumps. One pump per subsystem (loop) is powered from a Unit 1 safeguard bus and the other pump is powered from a Unit 2 safeguard bus. In order to ensure adequate onsite power sources to the systems during a loss of offsite power event, the inoperability of these supplies are restricted in system ACTION statements.

RHRSW is a manually operated system used for core and containment heat removal. Each of two RHRSW subsystems has. one heat exchanger per unit. Each RHRSW pump provides adequate cooling for one RHR heat exchanger. By limiting operation with less than three OPERABLE RHRSW pumps with OPERABLE Diesel Generators, each unit is ensured adequate heat removal capability for the design scenario of LOCA/LOOP on one unit and simultaneous safe shutdown of the other unit.

Each ESW pump provides adequate flow to the cooling loads in its associated loop. With only two divisions of power required for LOCA mitigation of one unit and one division of power required for safe shutdown of the other unit, one ESW pump provides sufficient capacity to fulfill design requirements. ESW pumps are automatically started upon start of the associated Diesel Generators.

Therefore, the allowable out of service times for OPERABLE ESW pumps and their associated Diesel Generators is limited to ensure adequate cooling during a loss of offsite power event.

3/4.7.2 CONTROL ROOM EMERGENCY FRESH AIR SUPPLY SYSTEM - COMMON SYSTEM The OPERABILITY of the control room emergency fresh air supply system ensures that the control room will remain habitable for operations personnel during and following all design basis accident conditions. Constant purge of the system at 1 cfm is sufficient to reduce the buildup of moisture on the adsorbers and HEPA filters. The OPERABILITY of this system in conjunction with control room design provisions is based on limiting the radiation exposure to personnel occupying the control room to 5 rem or less Total Effective Dose Equivalent. This limitation is consistent with the requirements of 10 CFR Part 50.67, Accident Source Terms.

Since the Control Room Emergency Fresh Air Supply System is not credited for filtration in OPERATIONAL CONDITIONS 4 and 5, applicability to 4 and 5 is only required to support the Chlorine and Toxic Gas design basis isolation requirements.

Additionally, based on implementation of the Alternative Source Term methodology, it has been demonstrated through analysis that manual initiation of the CREFAS radiation mode within 30 minutes of the start of gap release for the limiting design basis LOCA is sufficient to assure that control room operator dose limits in 10 CFR Part 50.67 are met.

The CREFAS is common to Units 1 and 2 and consists of two independent subsystems. The power supplies for the system are from Unit 1 Safeguard busses, therefore, the inoperability of these Unit 1 supplies are addressed in the CREFAS ACTION statements in order to ensure adequate onsite power sources to CREFAS during a loss of offsite power event. The allowable out of service LIMERICK - UNIT 2 B 3/4 7-1 Amendment No.

ATTACHMENT 6 LIMERICK GENERATING STATION UNITS 1 AND 2 Docket Nos. 50-352 & 50-353 License Nos. NPF-39 & NPF-85 License Amendment Request "LGS Alternative Source Term Implementation" List Of Commitments LGS AST LAR Page 1 of 1 February 27, 2004 The following table identifies those actions committed to by Exelon in this document. Any other statements in this submittal are provided for information purposes and are not considered to be regulatory commitments.

Commitment Continuing Scheduled Compliance Completion Date Per TSTF-51, licensees adding the term 'recently" must X Upon make the following commitment which is consistent with Implementation.

NUMARC 93-01, Revision 3, Section 11.3.6.5, "Safety Removal for Removal of Equipment from Service During Shutdown Conditions," subheading "Containment - Primary (PWR)/Secondary (BWR)". Exelon makes a commitment to the following NUMARC 93-01 section:

"In addition to the guidance in NUMARC 91-06, for plants which obtain license amendments to utilize shutdown safety administrative controls in lieu of Technical Specification requirements on primary or secondary containment operability or ventilation system operability, during fuel handling or core alterations, the following guidelines should be included in the assessment of systems removed from service:

-During Fuel Handling/Core Alterations, ventilation system and radiation monitor availability (as defined in NUMARC 91-06) should be assessed, with respect to filtration and monitoring of releases from the fuel.

Following shutdown, radioactivity in the RCS decays away fairly rapidly. The basis of the Technical Specification operability amendment is the reduction in doses due to such decay. The goal of maintaining ventilation system and radiation monitoring availability is to reduce doses even further below that provided by the natural decay and to avoid unmonitored releases.

-A single normal or contingency method to promptly close primary or secondary containment penetrations should be developed. Such prompt methods need not completely block the penetration or be capable of resisting pressure. The purpose of this is to enable ventilation systems to draw the release from a postulated fuel handling accident in the proper direction such that it can be treated and monitored."

Limerick is defining the definition of prompt in this context to mean being accomplished within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

ATTACHMENT 7 LIMERICK GENERATING STATION UNITS 2 AND 3 Docket Nos. 50-352 50-353 License Nos. NPF-39 NPF-85 License Amendment Request ULGS Alternative Source Term Implementation" Compact Disk Containina LGS Meteoroloaical Data

ATTACHMENT 8 LIMERICK GENERATING STATION UNITS 2 AND 3 Docket Nos. 50-352 50-353 License Nos. NPF-39 NPF-85 License Amendment Request "LGS Alternative Source Term Implementation" Supportina Input Parameters for AST Calculations

Attachment 8 Technical Parameters for AST Calculations ATTACHMENT 8 Technical Parameters for AST Calculations This attachment serves as a supplement to provide additional technical information needed to fully understand the technical analyses that were performed.

Table of Contents

'* ' iscussion Sp ,, Topage.' " _ a' ¢! -' * . ' '. . '.g .

Equilibrium Two-Year Cycle Isotopic Core Inventory 2 Maintaining Operator Doses ALARA in the Vicinity of the ECCS Pipe at the Control Room North 6 Wall Treatment of Alternate Drain Pathway MSIV Leakage 9 Atmospheric Dispersion 15

' Tables P age Table 1, Bounding Core Isotopic Activities For RADTRAD Input 3 Table 2, Bounding Core Isotopic Concentrations in Grams 5 Table 3, MicroShield Calculated Integrated Operator Direct Shine Doses Due to Core Spray Pipe 7 Table 4, Determination of MSL Decontamination Factors Due to Iodine Deposition 12 Table 5, LGS Condenser Characterization 13 Table 6, Main Steam Piping Summary 14 Table 7, ARCON96 X/Q (sec/m 3 ) Results for Control Room 17 Table 8, PAVAN Options Invoked for the EAB and LPZ X/Q Calculations 18 Table 9, PAVAN X/Q (sec/m 3 ) Results, North And South Stacks to EAB and LPZ 21 Table 10, X/Q (sec/m 3 ) Results Summary 21 Table I Ia - I Ie, RADTRAD Input Information (Primary Containment Leakage Pathway) 22 Table 12a - 12g, RADTRAD Input Information (MSIV Leakage Pathway) 26 Table 13a - 13e, RADTRAD Input Information (ECCS Leakage Pathway) 32 Table 14a - 14b, RADTRAD Control Room Results Summary (by Mode) 36 Table 15, RADTRAD Bounding LOCA Results By Pathway 37

- .. - . .: ' -; g; F' . ;es - - - - Page. ,

Figure 1, Control Room Isodose Diagram 8 Figure 2, Assessment of Steam Line Temperatures for Piping Deposition Credit Analysis II Attachment 8, Page 1of 37

Attachment 8 Technical Parameters for AST Calculations Equilibrium Two-Year Cycle Isotopic Core Inventory The LGS core isotopic activities are based on those developed for Peach Bottom.

The Peach Bottom Source Term fuel cycle assumptions are considered bounding relative to Limerick Generating Station. The Peach Bottom batch sizes, batch average bumups, and reload bundle average enrichments are similar to those at Limerick. For example, the bundle average enrichment used in the Peach Bottom analysis was approximately 4.11 wt. % U-235. The expected bundle average enrichment for LGS is approximately 4.16 wt. % U-235. Additionally, the rated thermal power level for the Peach Bottom analyses was 3514 MWt (uprated for Appendix K Thermal Power Optimization). Source terms are developed on a curie/MWt basis, and ultimately multiplied by the maximum rated power, including the applicable instrument uncertainty. The two plants are essentially identical for the accident analysis power.

LGS has a similar core loading as PB-3 (in terms of mtU of uranium). Design cycle lengths for LGS are bounded by the 71 1 EFPD assumed for the PB-3 source term; shorter cycle lengths would result in reduced activities from long-lived isotopes. Since the batch sizes, enrichments, and burnup distributions are similar between PB-3 and LGS (as noted below), the PB-3 AST source term would be bounding for current LGS cycle designs.

Overall, minor LGS and PB differences such as these in initial enrichment, power levels, loading, and ultimate bumup are not significant in development of source terms from ORIGEN simulations. In general, these source terms are considered representative for two year fuel cycles for plants of this size and are not intended to be adjusted for future reloads with these general characteristics.

The ORIGEN2.1 core inventory activity and composition results for the equilibrium two-year cycle at 100 EFPD (BOC) and EOC are shown below in Table 1. The maximum of the 100 EFPD and EOC values for each isotope are selected to generate the bounding isotopic core inventory activity and composition results as shown in Table I below.

Attachment 8, Page 2 of 37

Attachment 8 Technical Parameters for AST Calculations Table I Bounding Core Isotopic Activities For RADTRAD Input Isotopic Isotopic Bounding Bounding Isotope Activity at' Activity at Isotopic Isotopic 100 EFPD  : EOC Activity Activity

,.) ('). (Ci) (Ci/MWt)

KR 83M 1.324E+07 1.1 58E+07 1.324E+07 3.767E+0O BR 84 2.373E+07 2.002E+07 2.373E+07 6.751 E+O' BR 85 2.888E+07 2.404E+07 2.888E+07 8.216E+0O KR 85 8.806E+05 1.387E+06 1.387E+06 3.946E+o0 KR 85M 2.922E+07 2.436E+07 2.922E+07 8.313E+O' RB 86 1.118E+05 2.291 E+05 2.291 E+05 6.518E+01 KR 87 5.739E+07 4.672E+07 5.739E+07 1.633E+04 KR 88 8.096E+07 6.570E+07 8.096E+07 2.303E+04 RB 88 8.1 97E+07 6.678E+07 8.197E+07 2.332E+04 SR 89 9.836E+07 8.846E+07 9.836E+07 2.798E+04 SR 90 6.982E+06 1.117E+07 1.117E+07 3.178E+O'-

- Y 90 7.142E+06 1.150E+07 1.150E+07 3.272E+0O SR 91 1.336E+08 1.11OE+08 1.336E+08 3.801 E+04 Y 91 1.212E+08 1.143E+08 1.212E+08 3.448E+04 SR 92 1.412E+08 1.205E+08 1.412E+08 4.017E+04 Y 92 1.416E+08 1.21 OE+08 1.416E+08 4.029E+04 Y 93 1.591 E+08 1.404E+08 1.591 E+08 4.526E+04 ZR 95 1.521 E+08 1.578E+08 1.578E+08 4.489E+04 NB 95 1.399E+08 1.586E+08 1.586E+08 4.512E+04 ZR 97 1.637E+08 1.578E+08 1.637E+08 4.657E+04 MO 99 1.771 E+08 1.785E+08 1.785E+08 5.078E+04 TC 99M 1.551 E+08 1.563E+08 1.563E+08 4.447E+04 RU103 1.234E+08 1.477E+08 1.477E+08 4.202E+04 RU105 7.642E+07 1.022E+08 1.022E+08 2.908E+04 RH105 7.31 OE+07 9.673E+07 9.673E+07 2.752E+04 RU106 3.856E+07 6.081 E+07 6.081 E+07 1.730E+04 SB127 8.572E+06 1.01 8E+07 1.01 8E+07 2.896E+02 TE127 8.402E+06 1.01 OE+07 1.01 OE+07 2.873E+02 TE127M 1.039E+06 1.355E+06 1.355E+06 3.855E+02 SB129 2.732E+07 3.036E+07 3.036E+07 8.638E+0O TE129 2.679E+07 2.988E+07 2.988E+07 8.501 E+03 TE129M 3.875E+06 4.453E+06 4.453E+06 1.267E+0O 1129 2.774E+00 4.816E+00 4.816E+00 1.370E-0O TE131M 1.269E+07 1.360E+07 I 1.360E+07 3.869E+0O 1131 j 9.139E+07 9.444E+07 9.444E+07 _ 2.687E+04 XE131M I 1.015E+06 1.056E+06 I 1.056E+06 I 3.004E+02 TE1 32 1.322E+08 1.343E+08 1.343E+08 3.821 E+04 1132 1.338E+08 1.364E+08 1.364E+08 3.881 E+04 Attachment 8, Page 3 of 37

Attachment 8 Technical Parameters for AST Calculations Table 1 Bounding Core Isotopic Activities For RADTRAD Input (Continued)

. Isotopic ,  ;'.oIsotopic B Bounding Bounding Activity at Activity at Isotopic Isotopic Isotope 100 EFPD 10 EFEOC Activity' Activity (Ci) - (Ci) (Ci) (Cl)) (Ci/MWt) 1133 1.953E+08 1.925E+08 1.953E+08 I I 5.556E+04 XE133 1.904E+08 1.930E+08 1.930E+08 J 5.491E+04 XE133M 5.956E+06 6.007E+06 6.007E+06 I 1.709E+03 1134 2.167E+08 2.11 8E+08 2.1 67E+08 6.165E+04 CS134 1.335E+07 2.559E+07 2.559E+07 7.280E+03 1135 1.825E+08 1.806E+08 1.825E+08 5.192E+04 XE135 7.832E+07 7.086E+07 7.832E+07 2.228E+04 XE1 35M 3.636E+07 3.773E+07 3.773E+07 1.073E+04 CS136 3.568E+06 7.1 23E+06 7.123E+06 2.027E+03

_CS137 9.460E+06 1.595E+07 1.595E+07 4.538E+03 BA1 37M 8.965E+06 1.51 OE+07 1.51 OE+07 4.296E+03 XE138 1.679E+08 1.589E+08 1.679E+08 4.777E+04 CS138 1.841 E+08 1.760E+08 1.841 E+08 5.238E+04 BA139 1.787E+08 1.719E+08 1.787E+08 5.084E+04 BA140 1.721 E+08 1.661 E+08 1.721 E+08 4.896E+04 LA1 40 1.764E+08 1.722E+08 1.764E+08 5.019E+04 LA1 41 1.631 E+08 1.563E+08 1.631 E+08 4.640E+04 CE141 1.579E+08 1.575E+08 1.579E+08 4.492E+04 LA142 1.593E+08 1.51 OE+08 1.593E+08 4.532E+04 CE143 1.556E+08 1.449E+08 1.556E+08 4.427E+04 PR143 1.509E+08 1.416E+08 1.509E+08 4.293E+04 CE144 1.01 2E+08 1.264E+08 1.264E+08 3.596E+04 ND147 6.459E+07 6.321 E+07 6.459E+07 1.838E+04 NP239 1.616E+09 1.897E+09 1.897E+09 5.397E+05 PU238 2.639E+05 6.31 2E+05 6.312E+05 1.796E+02 PU239 3.1 OOE+04 4.218E+04 4.218E+04 1 .200E+01 PU240 2.871 E+04 4.526E+04 4.526E+04 1.288E+01 PU241 1.365E+07 2.173E+07 2.173E+07 6.1 82E+03 AM241 1.634E+04 3.349E+04 3.349E+04 9.528E+00 CM242 3.645E+06 8.393E+06 8.393E+06 2.388E+03 CM244 2.654E+05 9.147E+05 9.147E+05 2.602E+02 Attachment 8, Page 4 of 37

Attachment 8 Technical Parameters for AST Calculations Table 2 Bounding Core Isotopic Concentrations (Grams by Isotope)

.1sotIsotopic - Concentration Isotope' (grams)

KR 83M 6.41 4E-01 XE131 M 1 .260E+01 BR 84 3.370E-01 TE132 4.423E+02 BR 85 3.741 E-02 1132 1.321 E+01 KR 85 3.534E+03 1133 1 .723E+02 KR 85M 3.550E+00 XE1 33 1.031 E+03 RB 86 2.81 4E+00 XE1 33M 1 .339E+01 KR 87 2.025E+00 CS1 33 1 .678E+05 KR 88 6.451 E+00 1134 8.11 8E+00 RB 88 6.826E-01 CS134 1 .977E+04 SR 89 3.384E+03 1135 5.1 95E+01 SR 90 8.1 83E+04 XE1 35 3.065E+01 Y 90 2.113E+01 XE135M 4.1 40E-01 SR 91 3.683E+01 CS135 7.841 E+04 Y 91 4.939E+03 CS136 9.71 5E+01 SR 92 1.123E+01 CS1 37 1 .832E+05 Y 92 1.471 E+01 BA1 37M 2.807E-02 Y 93 4.768E+01 XE1 38 1 .745E+00 ZR 95 7.341 E+03 CS1 38 4.348E+00 NB 95 4.054E+03 BA1 39 1 .092E+01 ZR 97 8.560E+01 BA1 40 2.359E+03 MO 99 3.720E+02 LA1 40 3.1 68E+02 TC 99M 2.971 E+01 LA1 41 2.883E+01 RU103 4.575E+03 CE1 41 5.541 E+03 RU1 05 1.51 9E+01 LA1 42 1.1 15E+01 RH105 1.1 46E+02 CE143 2.342E+02 RU106 1.81 7E+04 PR1 43 2.241 E+03 SB1 27 3.811 E+01 CE1 44 3.962E+04 TE1 27 3.826E+00 ND1 47 8.039E+02 TE1 27M 1.436E+02 NP239 8.1 73E+03 1127 8.040E+03 PU238 3.685E+04 SB129 5.397E+00 PU239 6.782E+05 TE1 29 1 .426E+00 PU240 1.986E+05 TE1 29M 1 .478E+02 PU241 2.1 08E+05 1129 2.727E+04 AM241 9.755E+03 TE131M 1.704E+01 CM242 2.537E+03 1131 7.61 5E+02 CM244 1.1 30E+04 Attachment 8, Page 5 of 37

Attachment 8 Technical Parameters for AST Calculations Maintaining Operator Doses ALARA in the Vicinity of the ECCS Pipe at the Control Room North Wall The purpose of this attachment is to reassess gamma shine from sources external to the control room, using alternative source terms to assess contained radioactivity, and in some cases, a more detailed geometry treatment.

The dominant radiation source outside of the control room is a Unit I vertical run of 14 inch NPS core spray piping that is located 18 inches from the Reactor Enclosure - Control Room wall. This wall is 36 inch thick concrete. The historically determined dose contribution in the control room from this pipe and other lesser piping contributors is 4.2 rem whole body. The following give some consideration to impacted areas in the control room and the potential for use of local administrative controls:

• The core spray fluid is ECCS (Suppression Pool) water. AST source terms are more favorable because of the reduction in assumed fractions of core iodine released to the suppression pool from 50% to 30%. This is offset only partially by increases in certain other non-halogen, non-noble gas isotopes.

• Doses are evaluated with credit for Control Room occupancy per RG. 1.183 of 1.0 for the first day, 0.6 for the next 3 days, and 0.4 for the following 26 days.

• Dose assessment is performed using the point-kernel method as implemented in MicroShield 5.05.

• Doses are calculated as a function of distance into the Control Room, since doses drop significantly with distance.

• All Control Room equipment in the near vicinity has been identified and evaluated to determine occupancy required and the potential for administrative controls to reduce occupancy in the near vicinity.

o Cabinet IAC696 and IBC696 are for the LOCA H2 Recombiner Control Panels, and are expected to require minimal operator presence. If used, system startup and periodic monitoring is all that is considered to be required.

o Cabinet I AC464 is an ERFDS Multiplexer Enclosure requiring no operator presence.

Cabinets further along this wall will have far lower dose rates than shown because of the severe angle through the shield.

o Cabinets ODZ585 through 10C690 are for loose parts monitoring, meteorological instrumentation, and RDMS and MMDRS displays, printers and terminals. The meteorological stations are used chiefly until the Tech Support Center and (offsite)

Emergency Operations Facility (EOF) is functional. Other equipment will require minimal use during a design basis accident.

o Operator locations for Cabinets 10C626 through 10C610 are outside the 0.22 rem isodose line and therefore require no special administrative controls.

o Some administrative control may be required in the adjacent office, although the transport angle through the wall would minimize this consideration.

• Administrative controls to maintain operator doses as low as is reasonably achievable (ALARA) will include periodic habitability surveys by Radiation Protection personnel as currently required in Emergency Operating Procedures (EOPs). In addition to these surveys, operators will need to sign in on a Radiation Work Permit (RWP) to access the plant. This RWP will contain instructions to minimize time spent in this area. If the dose rate exceeds 2.5 mR/hr, RP personnel will be required to post the area as a Radiation Area.

• A zone is identified where these controls are practical. All other areas of the control room are assumed to have a dose contribution equivalent to that calculated for the boundary of the established zone.

Attachment 8, Page 6 of 37

Attachment 8 Technical Parameters for AST Calculations The potential contributions of other external sources are revisited qualitatively to assure that dose contributions are not significant. Sources considered include (a) The Unit 2 Core Spray line, (b) Reactor Enclosure airborne activity; SGTS Filter Shine; and Control Room Filter Shine.

Core Spray Line Analysis:

The steps performed this analysis were:

1. Use RADTRAD with RG 1.183, Table 1 non-noble gas release fractions and timing, with activity released to the suppression pool.

2. Have RADTRAD calculate the compartment activity as a function of time. Select a number of time steps and spacing such that source integration can be performed linearly by multiplying the step duration times the average activity over the step based on the activity at the start and end. A total of 53 time steps were used, and linear integration is conservative.

3. Extract results from the RADTRAD output into a table of the RADTRAD calculated activities and then perform the integration.

4. Use a spreadsheet to take the integrated sources through I day, 4 days, and 30 days, and apply occupancy factors, and unit conversions to create a MicroShield suitable time integrated source file. (Ifa source in uI'i/cc is input into MicroShield then doss rates are calculated(e.g. nRlir).

- If the integratedsource input is in uCi - hr/c, the result is a dose (e.g. in mR).]

5. Run MicroShield using the above source file to determine doses, with design basis Control Room occupancy, at distances from the Control Room interior wall of 1,3, 6, 9, 12, 15, 18, and 21 feet.

A 20-foot piping segment is used, with doses calculated at 6 feet above the floor. Doses are calculated at these distances perpendicular to the wall. (Note that doses at the same distances from the source are also calculated at 15 and 30 degrees from perpendicular, but are not credited for conservatism, and to avoid oblique angle buildup complications.)

Results of Dose Assessment for Core Spray Pipe to Control Room The Isodose Curve figure on the following page shows the calculated dose rates (perpendicular) overlayed on a small portion of a control room layout drawing to identify locations affected. The underlying layout is taken from LGS Drawing M-602, Rev. 29.

MicroShield Calculated Integrated Operator Doses are:

Table 3 1l Integrated Dose (rem) vs. Distance and Angle 1 Distance From 1 3 6 9 12 15 18 21 Wall (feet)

Perpendicular 1.081 0.808 0.571 0.428 0.334 0.268 0.220 0.183 15 degrees 0.764 0.561 0.402 0.302 0.236 0.189 0.155 0.129 30 degrees 0.242 0.179 0.126 0.094 0.074 0.059 0.049 0.041 potentially non-conservative Other External Sources

1. 18 inch NPS RHR Piping: This source was conservatively modeled as two 34 foot pipe segments located at 54.5 feet from the 1 ft inside the control room dose point perpendicular to the center of the two pipe segments. This is a conservative approximation for the piping visible to the control room wall. A MicroShield result for a 34 foot segment, doubled, is approximately 0.12 rem.

2. Other sources such as reactor enclosure airborne and external cloud and RERS, SGTS, and CREFAS filters are negligible because of shielding, distance or both.

Attachment 8, Page 7 of 37

Attachment 8 Technical Parameters for AST Calculations 30 ayoperator dos, Design Basis Full Ti 11'1 1 4

_4 ZC F. T n R

___ _ Cule. No.

I LM64M, Ro. 0, AfchmeFt Cr Page hC ToA Figure I Attachment 8, Page 8 of 37

Attachment 8 Technical Parameters for AST Calculations Treatment of Alternate Drain Pathway MSIV Leakage MSIV leakage is the only Secondary Containment bypass pathway analyzed for radiological dose consequences.

The radioactivity associated with all MSIV leakage is assumed to be released directly from the Primary Containment and into the Main Steam Lines. MSIV leakage has separate limits and a separately analyzed dose assessment, therefore it is not included in the La fraction limit, and is instead separately controlled.

MSIV leakage assumed in this accident analysis is 200 scfh total for all steam lines and 100 scfh for any one line.

Therefore, at upstream conditions this is a flow rate of:

100 scfhlline

• 14.7 psia / (14.7 psia+22 psig) /60 minhr = 0.668 cfm.

MSIV leakage testing is performed at 22 psig. Containment pressures above the MSIV test pressure persist for only about the first 6.5 minutes of the DBA-LOCA. During this limited time period very little containment air is transported into the inboard piping and even less to outboard components.

Informal test runs suggest that leakage during this period, and for the first 20 minutes, results in negligible dose contributions, even if an adjustment were made to extrapolate leakage to what might be expected if MSIVs were tested at the LGS P. of 44 psig.

However, to provide design margin, the above leak rate is increased by 25% for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to a value of 0.834 cfm, as shown in the attached table ("Determination of MSL Decontamination Factors Due to Iodine Deposition."

Outboard flow rates are based on expansion of this fluid from the MSIV test pressure to atmospheric pressure, and by further expansion based on worst case heating the fluid to steam line temperatures from standard temperatures. Steam line temperatures are derived based on a generic BWR evaluation crediting only conduction through pipe walls and insulation and is discussed in the attached table, "Assessment of Steam Line Temperatures for Piping Deposition Credit." Credit is taken for temperature reductions only at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and at 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br />. This results in the outboard flow rates shown.

Flow rates out of the condenser are similarly calculated with the assumption of a condenser air space temperature of 120 OF for the accident duration.

Determination of inboard steam line, outboard steam line and condenser effective filter efficiencies is shown in the attached table ("Determination of MSL Decontamination Factors Due to Iodine Deposition."

Modeling of Deposition Credit in Pipes and Condenser LGS has previously been analyzed and licensed to no longer credit a MSIV Leakage Control System, and to credit seismically analyzed portions of Turbine Condenser System. This historical evaluation is based on methodology described in NEDC-31858P, Rev. 2, "BWROG Report for Increasing MSIV Leakage Rate Limits and Elimination of Leakage Control Systems". That analysis was based on a design basis recirculation line break and TID-14844 based source terms. In this calculation the analysis of MSIV leakage is updated to reflect Alternate Source Term parameters related to release timing and chemical makeup and more recent approaches regarding fission product settling and deposition.

Attachment 8, Page 9 of 37

Attachment 8 Technical Parameters for AST Calculations Modeling of aerosol settling and elemental iodine deposition is based on methodology used by NRC in AEB-98-03. For the two bounding steam lines modeled, two nodes are used. The first node is from the reactor pressure vessel to the inboard MSIV. The second node is from the inboard MSIV to the Turbine Stop Valve that provides the seismically designed boundary of the MSIV Leakage Control System. For aerosol settling, only horizontal piping runs are credited, and only the bottom surface area is credited. Per AEB-98-03, a median settling velocity is used, given the conservatism in using a well-mixed treatment. For elemental iodine deposition both horizontal and vertical piping is credited, as well as all surfaces since this deposition is not gravity dependent.

The attached table, "Main Steam Piping Summary Unit 1," shows the derivation of piping volumes, surface areas for settling and deposition, and piping effective filter efficiencies for each piping node.

For conservatism, no credit is taken for deposition in the drain lines that provide the previously licensed Alternative Drain Path to the condenser. Credit is taken for deposition in the condenser, where the deposition area is the horizontal surface of the wetwell, and the HP condenser walls. By the time that activity has reached the condenser the aerosols are essentially depleted. Therefore, vertical wall surfaces are credited for elemental iodine removal. No credit is taken for any organic iodine removal in piping or the condenser.

All MS drain lines are routed to a single penetration in the HP condenser at a point below the condenser tubing. Unlike NEDC-31 858P, iodine resuspension from settled or deposited iodines is not calculated. Historically, this phenomenon increased organic iodine release by about a factor of two based on resuspension of TID-14844 based elemental iodine fractions. The presence of this phenomenon is questionable with aerosols with significant cesium loadings. Furthermore, while deposition on condenser tubing is not formally credited, test cases have shown that substantial removal of elemental and even organic iodine would be predicted that would more than offset any resuspension. These results are shown in the attached table, "LGS Condenser Characterization."

The condenser tubing provides a surface area that is 40 times that of the credited wall and bottom surface areas. It should also be noted that the HP, IP, and LP condensers are interconnected by substantial openings, but flow the IP and LP condensers for further holdup is not credited.

Flow rates out of the condenser are assumed to be at 120OF and atmospheric pressure. A factor of 1.25 is applied, as is done with leakage and flow through outboard steam lines. This leak rate is also reduced by 50% after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, consistent with the change in Containment conditions.

Attachment 8, Page 10 of 37

Attachment 8 Technical Parameters for AST Calculations Figure 2 Assessment of Steam Line Temperatures Value for Piping Deposition Credit Analysis Tim e Temperature Used (hrs O°K OF OF Main steam line wall temperatures following a 0 565.3 557.9 558 LOCA have not been calculated specifically 1 561.1 550.3 for LGS. A generic cooldown analysis 2 557.0 542.9 developed by GE and reported in August 1990 552.9 535.5 Cline Report (Reference 28) is used. Cooling T(0 K) = 299.7 + 265 .6 *e .428*11 548.9 528.3 Is considered Independent of leakage flow. 545.0 521.2 Only conduction Is considered. The function where to the right Is from Reference 28 and fits the 541.1 514.3 figure below (from Reference 29) and is used t time, sec.

for Limerick: 537.3 507.4 8 533.5 500.6 TEMPERATURES OF THE MSIV LEAKAGE LINES 9 529.8 494.0 10 526.2 487.4 11 522.6 481.0 12 519.1 474.6 13 515.6 468.4 14 512.2 462.2 15 508.8 456.2 20 16 505.5 450.2 17 502.3 444.4 Ba 18 499.0 438.6 Do e 199 495.9 432.9 20 492.8 427.4 21 489.7 421.9 22 486.7 416.5 233 483.8 411.1 244 480.9 405.9 410 48 423.3 302.2 I 72 384.0 231.5 210 96 357.2 183.3 0 0.2 0.4 VA TIUE AMTRM N(cci FiGURE 7. Temperaturi of the MSIV Leakage Pathway Piping as a Function of TMme afler Shutdown. The Rate of Cooling IsNearly Independent of the Flow Rate at the Low Rates Considered Inthe Present Analysis.

Attachment 8, Page II of 37

Attachment 8 Technical Parameters for AST Calculations Table 4 Determination of MSL Decontamination Factors Due to Iodine Deposhlon Condesr Coniemer Candemor hbMt A laboard D "Gor C inboard 0 Outboard A 0bo0d 11 Ostboerd C Outebard 0 No Tube Cred 10% Tbbo 190% Tubes Total Pipesorfses AMrna 512 W5s3 et 232-5 293 2. 237 b=59 443701is 349838

,TOtratPievotirn (19) 335 mg 30S 218 1le2 1t51 1236 10 547405 7445 3 547455 qHG140ntaI TOWo Pip. 8ur1ac4 Area (15') 219 315 311 21a 7232 1931 23ED 22s tVA WA tl'A Holer"otf Seffling MipeSurface lisa[1' 3oea 117C0 13762 C91go 112598 954A 1`17982 119342 134.0C 1t96281 1705406 110 129 159 110 1133 1M21 11C7 1t'6 345705 34702 a 4745 5 JAarosc Sells6" (nt)

Cdey 1.1TCE43t 1.03 1772CE 1751.03 I t7CE43 1.17750t2 1 17M.03 170E4.3 1 T2E-0 11705r0 1.17DE40 0 Awa0406Set"~n Velocity Mr9 s) 3 33E4.03 3 835-0 38231.3 3e83E4-0 3 8394.03 3QE330 282150E3 11136E3 3853E-0 38355.03 S839E43 t

' 51.mnlat 0.posl~an Velocity "Owis (f31106) o,395E-0 51E9-05 0531E.L1 53G5 -S 533275- 09385-Cl 93350 53303.06 229SE-04 22985E04 2.29eE44 5MonentiW Dopoome Vetocity l4.in (Mnac) I 2495E0 5 240E0M 1 248-C 1 24&E. 12035 5254951C501240Eas 16.243C50 22199E44 2298Et4 2259.E44 "ElemefiW Deves Nwa vei06cyl6 9:wn foMeci 7.8475.0 784150 7 945.C 7.e E 53 79-sE-s 7.55-C 7 943.o 7.943E05 22BE-44 2298E-04 1295E.04 EloenwdADepositio Velocity .24hv8(19t-46I t.7ttE4D 1 7ItEam 1 ee.C t 7I Es5 1778E.0 1.725-Cl 7Sn.l m 1759E. T735E44 70715.04 7C71E-04 floev"Wa Deposition V~O"li2"Itirs; (111141 4 tE405 4 w e5 AM 05- 4 E M 40a5E4O5 4UE-C 4D09E43 40915435 7IE-04 70TE-1-4 7D.07E-04 Etienwei Ceol" o Vs"oil 99-739111 tftise 2+/-17E.04 2 s57-04 2075E-4 2 075.04 20 37.04 297-94 2395E704 260704 71070E44 707E.4 7907E.04 59H9M-0 53e99.0 5 wEM rs9 59MEto 39me7e3 515-C 3913509 a 9911A39 2049E-s7 2b4eE-07 2U-495-07 tsolgaflkc p.9.18cm viotoM124hMrs (0f65*)

1.32ME5 1t3935.0 1 26NI 135Et. 1 3249 1.2rE-EB 1 3.29MO01.78mEa 294rE507 24b9E47 2.948E.7

'1orgamse Deposm Vs" fltl-r bnvw a 949"-0 B345.0 a 8435.068 845-CS 68-4203 1405-C 88435E08 e835a.0 25493E7 2s5EE4.7 2-549E507 Organic Deposiio Velocity 0.24hes tt(lsle 1.9615E.09 t955.5E 1 95MCS It 5E. 19 .5-7 1 MCR0-Cl 9SC£49 1195.E01 &3E1E-7 39515E.7 D15E47 organic D"poeto Velocity 24-06hws (Riseel 495615.084tIE-Of 4015CGI AN11-C 42Et1 a 42815-Cl 4551E1.8 4M15E-0 93615E.7 £315t.07 361E4117 Organi Depoofiom Vetocety 96L720h1s (Mooec 2 0E4 2 23E.7 2B35.07 2803E57 2 0E47 20335-7 29035-7 2933E8.0 8315.E7 831E-07 83615E.7 Uncorrected Flow Rate (sclh) 1C0 103 103 933 1G0 103 10m tm0 RFc~osn low Rose0.2 ri; (061 .: 00s e 4 - Sin 98D = A' : NAr' 7 'A ' .;tA . i 'PA 'WA KM919 From CoanenIM l Roomt~ 24.720 Wns(cini4 041 .8. 41T . 0417 . 417  :.'A -VAWA, tVA: 2 . .

WA FC 0125 0939: GM05 '.0352 4.D17,:: J017- . -01A 4017Jr : 4*577 4377: . 40g7?-

PopeFlow Rae Zs."0 aws(06. 041? 047 0 01 i1 1716, . 1714 .171 I-.1711 -? 2299 . 2293 2-93 047:417: 041 041 3 a41f:. c41r e"t .. D~~,

111299 ' 1.23- r .. 229 _.._' 2.23 "'2219 93 50 co C 50085 100085 24134 24104G 241C04 241-- 4 27421 274821 .74M21 Pip Flow Rate 2446Ins (Fee 250mg 25 MA 25834 25034 19181 11D29 102983 12m93 137311 137311 137311 PlpeFM" Rose24.73Ws, (061) 25034 25(94 25C34 25.024 782US 78.123 78 1M 75123 t173`11 927.11 137311

'leissol Set11g RateCocsante (hVr) 1217E-0 1 3-01 1t37Et t ME-01 1.275,01 13TE.51 1375.01 1 3TEd1I *.15101 3045ECt 47sE00 4 485.01

'EierenWtisDposli~om RafteCntn &W1918.28 P(IW 1 2 MI E412s I 2t1E4t I 2T-t I 2561.01 1.2EE-Ct

• 29E41 126E M 4 45E-01 218E.00 1 7T5*.0

'Elmeoaw DepositionA Rare,C0.19.19 244Jelwphil 2t3E01 293E-01 250E47t 25D301 293801 293E4t 2 93EM 2 93.01 4 425.1 21t4CO 177501

'ElemtentatDePozilcm Rate Coeslant 66721111 WI I 871E.3 I 375 1 975.01 1 81t.00 1875.03 1870E-M 1 8?E7M 4 4s5E1 2.t3 1ICD77E01 32835.01 sor93dc Dtpep~w" RxotConstant ma.w (Srl 1 4CE-O 143E.04 1 .c.4 1E44 I A4350 1 4DE04 I12XCE-04 E4J I MM5'3 1 4104 488.4 242-E43 ts18E.2 01911r61 Opposliton Rate Constant 24461w (Wt) 32NE-4 3 2E5-04 32 M504 225E4.0 3291-041 325WE-4 32.w4 4 sE504 2 TEW 1-98 2

'Organic 0.9005061 RafteConstant rftV' 2 MM.3 2 NE.43 260E.0 2 9.E3 20449-0 26E5-3 20C8l0m 4 s6E4 2 2E5-3 13E-02

'Asosel PilterEl~ciney (0.14111hi) 9838% 9885% 9591 8628%X 8934% 3 29 mm 9 3S% 9936% M27% 9511, no009%

'Awast ollr Efficlency (2446219s) 9924% 9241% MM7% 0952%

. 93s% e565% 90SM g M38% 527% 9992%. mm9%

A.Eonesot Plaer EffkklenyIU273 Iis) 335% 434A%. .4242% 2230% .%  % 35.3% 3588'5 997Tr% . 397%

1. ElaonentalFlw Efficisey (026-9 hi's) 7`11% 7818% 7811%' 7311% 7709% 7494% 77Tn% 77.432 M54A 998"% mm9%

'Elarnunt FOUe Efficiency (24.9 his) 95 MM . 027% 931M% 9102% 9s5o% 9S17% MM7 'De" Sisment Felo Efild~efiy (0624 *51s 9044% 9245% 000%

0o 00%

D I% 0%

a 0 07% 0 07% o00% . D07$% n%

'Orgaofic Flow 879dMeay (.24-9 tirsi 1959% 43.17% 707`9%

0 332% a 0024% OM 037 032% 2oM% 0.8% -

0.pnoic Roes Efficiency (I.720 hrs) 203% 248% 24% 209% 305% 272% 319% ,3611% 1441% Av* M%

r'Ppe Wall Tenmperatwe0-2411r (F) Mg803 Pipe IVA Tooilpereture. 0-24hr(14)CASS37

'Pipe Ws" Tempealmer.34-06Win)4100on PlpewasTemperature. 2448hl(4) 483.113 UlillO oy0uorw 158.00.03, 1231096M P"y A-Z FbrnvJa 2

'Plpe Wall Tempweraue.88-720h11 () 20 000 *US'PFCDse. AEG88.03. 1ZOV094.P33 A.Z PMNwj 4 Pipe WASTenmperaftule 96.7=r1(4 MA64 " ClrE. .1,'AsrVLNM"lt3 odrarra-upot AnayeW6'. 309991 t FP.ME~CR.5994. RA0TRADlWsMM.4J129,8507181.t11. 61181 Condenser Tempyeratue. constavdt(K) 12002 1 0C0ne.JE. SM ,oshumn n VosIM AnalysisI.M.Cl6l.Rev 0, Alaxivorl F Mm Laic Coondamld Tevnpoature. Constant (14)CO2

& TMCseU7MWatsWfr031A lb' H P Caeei'3iVOUr

'tNWa0 Te#t tanure. 104,t0.4(V)fg 98 L08 Ute(rica Spooncaton C.3 31129a 9

"Standard TesdPressure. constant (psip) 44 LOSe7grrcel Spe13:ouS8R 48 6Ia2 9 Abnosphoetc PmsweM e~ost9.t M"l 14 7 Flow Rate Conservatism Fector, comsdanl Attachment 8, Page 12 of 37

Attachment 8 Technical Parameters for AST Calculations Table 5 LGS Condenser Characterization High Pressure Conderlser Intermediate Pressure Condenser Low Pressure Condenser Eras Section t1~ectanquilar Volume) Rams Sectim~jRectangularVolumsel Rase, Seodiou (Recannuilar Volume) 33 96 wge*It.L) 29 00 Loq,1lh (Cast. YKA irection,~) ft) 29 03 Lergth (0Eat W1 Divo00rctio) (M) 2900 Length (E11 Wst MUDiectionl (it) 10412Wulthi(Mortj,. Sbuth vDi~necti) 31 42 MAI15 ilbr~u SoulDirCectionr) (8.)

45710 75 Voekve M1~ 3902M00 Vdw*w M 3835 Vobarse (6'~)

Flue Sectioni (Traptrofid Volume) MmuSection (Trapazoidai Volume) Ft," Section (Trapaxeold Volume.)

is 67 Nsi.1tfoleelci=0 it) 18 67 Heitet owfcliait) 29 00 Loiith (Eass Vhi+/- Direction) (A) 29 00 Lenglth (0st West Direction) (MI) 2900 Length (E:01 Wes Of Drection) (ft) 23 67 Top Vfiaih (14,r&hCouth DOnctiam) (6) 29 67 Top V&Ih 01th(ib

• oct ire09ctlbi) (8.) 23607Top Width (tiwtm. South Directitor (6t) 46 42 fte"miYA:M okwn1'* South rem Y)(8.) 10 42 Scotom Wiolh Okteoli* Souls Diahdas) (It) 3442 bollbm Wulish (Nortli

• Souti, resction) (IL')

21h6 Sfarf Le.gth M 20.45SbrtLesslih (fq I"4 StartLeegth (1) 690922 Vckrm~ift~) I 73422 Volurne t t572122 Vobume (le)

Top S-coI tMatugutVotwme) Top Section (becanegul Volume) Top Section If~teWWuAr Volume) awmewplt(6j 6900Length EdWe ietin 29 30 Leqgts(E211VWe010icli) 4t1) N200 Lwngti (East. Wed tOhetior* (it) 23 BY Wells(?lrh. 6euth Dhc 0uift) 23 67 Wodh -1ot SouitufDirection) (8t) 23697Wdh,Ololh. Soulh Cirect1*611 (IL)

.Isoo Vofuyme t(" 4118law Vlume M6' 41 1800o Vobnme (f6')

422 Wstervqal(it) 4AMVfter Laof Olt) 1346e8 C~ordcenBase Area 17206 1 Cor keIea His. AM (6') Well6 Condenser Bass Arm (6')

MM M0Co r~aired VMer VOIur!, ift'i 507903 Cortisie Water Vob',. M6 432503 Cortirve d113twVolurne (M MUMe9Cordncme Tube Surbace Ame(ft'6 ma1o).0Cordemrec Tube Sulam Am Ee) N2,0636 Condenser Tube Suofaoe Area (16' IJKM C31cuta dLergof Tubes Ift) 1.041.225 C31cufted Lermh of Tubes M6) t91.726 CakJstedLer~n of rubes (6)

&2944 Vckr*', Lizoiaced by 1o4,es (R)' 7.1S7Ag VdL-me Cqsntced by Tubls (j05 6S5531 Vobume Dtspbccd by Tubos (hI')

11.5.*8.CSeerg Area (floor or"y 1t.172 CS Serbg Arco (how only) 99608 Solftrrg Am (6,0 orly)

-9=229Dpston Are (fltot olijOWis) 8299809 De~osoitn Area Olbee pkis Wals) 75ea AO Depositon Ameihloor r Ls Wats)

S!tVoluMe(') 857?97.97 She# Vohine (6') 6 1.2t22 Shiel volume (6') 53,72247 Volurne Above Whve'el MM)62& §4 Volune Above li*etwe (6') %I ea.19 VWU~ret Above W4tweIlIt)6' 49,407.44 rretAkrVoltrma M3) 5d.7L5 M Fret Air Volume. (63)49M01 rrve AlfVolune (C3 43,25-w UTueLengths 48 Tube Lenfttr 42 Tube Leftg"os M Carondwmf Tube 00D. 11) 0.0MM7 Tetal Corederver Tube Surfsce Area M "-OM Tube Surface Arma0ft. by shefi "3078 Tube Surface Ame (f?). by zho WM?66 Tube SursoeAroa (6'.by esle 20.857 Calculated T'ctfsLenh of UTues M.L)3,113(I8 CalduledTz ota Vlme Displaced by Tubes 4If1 211,5MM3 Refirences:

Total Shell Volume (ft3) 183,796 I LOS On, th. 90.1097.5.100. Rlow. Is 2 LOS 0.egL fj. 9N.107.5101, Re*V_ X Total Volume Above Hotvell (It3) 168,559 Total Free Air Volume (ft3) 146,996 I LO lb . 901097.5.102. Rev. E 4 LOS M%~ No. M0.10971112. Revt. K o LOS DM0 W. 1.3.21, Rev. tO. C-donsouea A Condegwer Systen?

Attachment 8, Page 13 of 37

Attachment 8 Technical Parameters for AST Calculations Table 6 Atin Steam Piping Sumrary Unit1 24.14 Main Sleam 2t Inch pipe ID LGS Spec, P.300. Rev. 44. .P4ipin Materilae 8 lns~rurn. PFlrtg Standarlds TOTAL MS PIPING A B C D 5435S75 5122.09 582259 5524 75 26 Inch piping(ehell) 2663 218 30,16 2t10 26 lnch pipngbside euteCe area (sqIL) 14u 1357 1542 1453 2f Inch piping hiside votume (c i It.)

2663 2133 3066 22t0 Total linsde surface aree (sq. .)

1440 1357 1542 1463 Total lnsde tokme (CU.IL) 2183 2693 304e 2110 Totallensiteurtacewea (*4 t.)

144 1357 1542 1463 Total nside votume (cu. IL)

HORIZONTAL MS PIPING ONLY A B C D 46t9250 43785.4 5079 34 4781.50 2f Inch piping (fehees) 2471 2306 2875 2513 26 hcIh piping Inside surface area (pq tL) 1243 116t0 1345 1256 2 finch piping kIside vokime (cL 2.)

2471 2306 2675 251E Total inside surface area (sq. 3.)

1243 1t60 1345 1266 Total Inskie voaie (cu. ft)

Horizontal Totals 4270M CO 378034 44eOO 4 4365.0 2 finch Oultoard Pipng. leengh Cinches) 2252 19S1 1 2360 l 2299 Total Outboard Pipe Surface Area Credit (sq. IL) 1133 100t 1 1187? 11s5 1Totaloutboard Pipe Vokmne Credit (cu. IL) 416f50 595s0 598.50 41 650 2ffinchInboard Piping, trlglnches) 1219 315 3I15s 1 211 IToal Inboard Pipe Surface Area Credit (sq. It) 110 is 1t9I 11e Total inboard Ppe Vome Credit fcul. It.)

Totals 4462.75 37.09 4657.59 4551.75 26 Inch Outbolrd Piping. Mnh (Ches) 12350 2039 2458 1 23*7 ITotal Outboard Pipe Surface Area Credit (sq. 04 1182 1 101 1236 1 120 ITotai Outboard Pipe Vokew Credit Icut. Il) 97303 1155.00 1155C0 9730C 20inchInboard Piping engt Qinches) 512 603 608 512 Total inboard Pipe Surface Area Credit lsq. 114 25S8 306 30B 251i Total tnboard Pipe Volm Crediet u.ft.)

NodeakZaon IHoriontsh l - l I -

416.50 59850 598 50 4t6 50 Node Ltength l )esl 219 315 315 219 Node I Surlace Area (O. t .1 110 159 159 110 Node 1VoDnle cu nt 4276 CO 3780 34 44B 84 4365 0D Node 2 Length ichieeS 2252 1991 2380 2299 Node 2 Surface Area lan.- -

1133 _ COl 1187 1156 Mode 2 Vohate u fl NudJik iton otal s -

97300 1 115500 115500 97300 eNodeltsefi a

512 60 ece 512 - Node I Surtace Are& Sq. IL 258 1 3 3C05 258 Node I Vorne (aCL. 1 446275 3i70D 46e7.59 455t 75 Node 2 Length (cheSl 2350 2C89 2458 2397 Node 2 Surface Area I It le2 1051 1236 1206 Node 2 Vokwwe lea .l I Attachment 8, Page 14 of 37

Attachment 8 Technical Parameters for AST Calculations Atmospheric Dispersion Source Configuration The North and South Stacks are executed by ARCON96 as a vent release. As depicted in Attachment B, both stacks have a height of 416 ft MSL (200 ft above grade). The stacks are located between Reactor Enclosures 1 and 2 with the North Stack situated on the north face of the buildings and the South Stack on the south face of the buildings (designated as Normal Release Point 2, "HVAC VENTS FOR REACTOR ENCLOSURES"). These stacks are less than 2.5 times the 194.75 ft high Reactor Buildings (i.e., the highest adjacent building), and therefore, per Regulatory Guide 1.145, they are modeled as a 'vent' release.

Both the North and South Stacks are conservatively assumed to have a zero (0) flow, for which ARCON96 requires that the exit velocity and stack diameter each be assigned an input value of zero (0). Per Regulatory Guide 1.194, Table A-2, the actual building vertical cross-sectional area perpendicular to the wind direction must be utilized; therefore, the Reactor Enclosures' combined vertical cross-sectional area of 5851 m2 (calculated as height = 59.4 m, and w = 98.5 m), was input into ARCON96 to account for wake effects.

Receptors The model ARCON96 was executed for X/Q at the Control Room Intake, which is centered on the north face of the Control Structure at a height of 124 ft above grade.

The direction, relative to true north (assumed 0°) of a straight line extending from the Control Room Intake towards the stack source location, is also an input parameter required by ARCON96. Attachment D depicts the two (2) separate intake-to-stack direction scenarios analyzed in this calculation. They are as follows:

Direction (degrees) Distance (m)

Intake to Stack Intake to Stack

• North Stack 180 16.5

• South Stack 180 64.8 Meteorological Data The Station's meteorological database for the five-year period from 1996 through 2000 was applied in the ARCON96 modeling analysis. Data measured at two (one primary and one backup) meteorological towers were used.

Meteorological Tower 1 is the primary tower and is located approximately 0.6 miles north-northwest of the North and South Stacks, whereas Tower 2 is the backup tower Attachment 8, Page 15 of 37

Attachment 8 Technical Parameters for AST Calculations and is located west at approximately 0.4 miles from the North and South Stacks. Tower 2 data were used only for substitution of any missing Tower 1 data as follows:

Limerick Meteorological Tower Instrument Levels (Elevations in reference to tower grade)

Tower 1 (primary) Tower 2 (backup)

Wind Speed:

Elevation 1 30 ft 159 ft Elevation 2 175 ft 304 ft Wind Direction:

Elevation 1: 30 ft 159 ft Elevation 2: 175 ft 304 ft The meteorological vendor illustrated that the Tower 2 delta temperature data are sufficiently representative to be substituted for the Tower 1 delta temperature data; however, since the Tower 1 and Tower 2 delta temperature height intervals differ from each other somewhat, and also since for all years shown, the primary Tower 1 has data recovery rates well above the NRC's 90 percent requirements, it was deemed unnecessary to make such substitutions.

Hereinafter, the Tower 1 ARCON96 meteorological input database with applicable Tower 2 values substituted for missing Tower 1 values as indicated above will be identified as the 'Tower 1 Modeling Database".

The designation of 'calm' is made to all wind speed observations 0.5 mph or less. The higher of the starting speeds of the wind vane and anemometer equipment on each of the towers (i.e. 0.5 mph) was used as the threshold for calm winds, per Regulatory Guide 1.145, Section 1.1. to this License Amendment Request contains the lower and upper level joint wind direction, wind speed, and stability class distribution tables, based on the five-year lower and upper level Tower 1 Modeling Database, as used for the ARCON96 modeling analysis. (These data are provided both in the format of number of observations and percent occurrence frequency.)

ARCON96 Run Scenarios Control Room Intake X/Q values were calculated by ARCON96 for various source/receptor scenarios. These two scenarios were analyzed using the five-year hourly meteorological joint wind and stability database, as identified below:

Attachment 8, Page 16 of 37

Attachment 8 Technical Parameters for AST Calculations ARCON96 METEOROLOGICAL DATABASE SCENARIOS RELEASE Wind Speed and Direction I Stability Class SCENARIO Primary Secondary* I (Delta Temperature) 1: North Stack Tower 1: 175 ft Tower 1: 30 ft Tower 1: 171 - 26' 2: South Stack Tower 1: 175 ft Tower 1: 30 ft Tower 1: 171 - 26'

• Secondary data used only for those hours when primary data are missing.

The upper level of the Tower 1 Modeling database is the obvious most representative monitoring location of choice for wind data representing the North and South Stack release points.

The North and South Stacks are not tall enough to avoid building-induced downwash; therefore, with zero (0) exit velocity having been assumed, ARCON96 treats their releases as a 'ground-level' type release.

Calculations The X/Q values resulting from the ARCON96 modeling analysis of each release and meteorological database scenario for the required time intervals are presented as follows:

Table 7 ARCON96 X/O (sec/iM3) RESULTS RELEASE / INTAKE &

METEOROLOGICAL 0-2 hour 2-8 hour 8-24 hour 1-4 day 4-30 day SCENARIO

1. North Stack to Control Room Intake:

• Wind: Tower 1 175' 6.88E-03 5.17E-03 2.04E-03 1.29E-03 9.63E-04 Stability: Tower 1 171 -26'

2. South Stack To Control Room Intake: 1.26E-03 9.64E-04 3.80E-04 2.39E-04 1.80E-04

• Wind: Tower 1 175' Stability: Tower_1 171 - 26' PAVAN MODELING ANALYSES OF CONTROL ROOM, EAB AND LPZ X/Q The model PAVAN is a commercial software package designated by Washington Group, International as MC-131, an "active" program applicable to nuclear safety related analyses as well as non-safety related studies and evaluations. The PAVAN code Revision 1 verification was performed for the 0-2 hour, 0-8 hour, 8-24, 1-4 day, and 4-30 day 0.5-percentile, and annual average direction-specific X/Q values, and the overall site 95-percentile maximum X(Q for each of the 0-2 hour, 0-8 hour, 8-24 hour, 1-4 day, and 4-30 day time-averaging periods. This verification was performed with WGI (formerly Raytheon Engineers & Constructors, Inc.) corporate standards, and is consistent with Computer Software Control, NEP-09. Revision 1 of MC-131 was verified for ground-level (i.e., non-elevated) releases, as well as elevated releases, with zero (0) vertical exit velocity assumed.

Attachment 8, Page 17 of 37

Attachment 8 Technical Parameters for AST Calculations Methodology and Acceptance Criteria The computer code PAVAN is a straight line Gaussian dispersion model utilized to estimate relative ground-level air concentrations (X/Q) for potential accidental releases of radioactive material from nuclear facilities. Such assessment is required by 10 CFR 100 and 10 CFR 50. The program implements the NRC guidance provided in Regulatory Guide 1.145. The technical basis for the program is presented by Snell and Jubach.

Utilizing joint frequency of occurrence distributions of wind direction, wind speed and Pasquill atmospheric stability class, PAVAN calculates X/Q values as a function of direction for various time-averaging periods at the EAB and the outer boundary of the LPZ.

Calculations are made from assumed ground-level (i.e., non-elevated) releases (such as vents and building penetrations), which are less than 2.5 times the height of adjacent solid structures, and from elevated releases (i.e., stacks). Three (3) procedures are utilized for calculating X/Q: a direction-dependent approach, a direction-independent approach, and an overall site X/Q approach.

The PAVAN model contains certain model options for executing the program. The table below summarizes the options invoked for the EAB and LPZ X/Q calculations.

Table 8 Option Description Option No. Invoked?

1 Calculate oa and ox based on desert diffusion. No 2 X/Q values include evaluation for no building wake. No 3 ENVLOP calculations printed which describe upper envelope No curve.

4 Print points used in upper envelope curve and calculation. Yes 5 Null 6 Joint frequency distribution in % frequency format. No 7 Print X/Q calculation details Yes 8 Distribute calm winds observations into first wind speed Yes category.

9 Use site-specific terrain adjustment factors for the annual Yes*

average calculations. es 10 Assume a default terrain adjustment factor for the average annual calculations. Option 10 is applied, which together with Yes application of Option 9 means that site specific terrain factors will be used.

• A uniform value of 1.0 is used.

Source Configuration Releases for Control Room Intake X/O Evaluation The North and South Stacks are the assumed release points. Because these stacks do not qualify as 'elevated' releases as defined by Regulatory Guide 1.145, in accordance with Regulatory Guide 1.194 methodology no PAVAN modeling (i.e., only ARCON96 modeling) is appropriate for the Control Room assessment.

Attachment 8, Page 18 of 37

Attachment 8 Technical Parameters for AST Calculations Releasesfor EAB and LPZ X/Q Evaluation As previously stated, the North and South Stacks have a physical height of 200 ft and are located to the south of the Control Room Intake. These stacks do not quality as elevated releases per Regulatory Guide 1.145. Therefore, the stacks were executed by PAVAN as

'ground' type releases requiring that each of these stack heights be assigned an input value of 10 m. The Reactor Building height of 59.4 m and smallest calculated Reactor Enclosure vertical cross-sectional area of 2426 m2 was used for each of the scenarios.

Receptors For the North and South Stack to the EAB and LPZ scenarios, PAVAN was executed in ground release mode with stack-to-intake horizontal distances of 731 m for the EAB and 2043 m for the LPZ.

Meteorological Data As described in Section 1.0, Limerick meteorological data from the five-year period, 1996-2000 for two meteorological towers (one primary and one backup), was used in the PAVAN analysis.

The format of PAVAN meteorological input consists of a joint wind direction (based on sixteen 22.5 degree sectors), wind speed (7 intervals), and stability class (7 classes) occurrence frequency distribution.

Each such meteorological joint frequency distribution for input to PAVAN was prepared by using the WGI pre-qualified program ARCONtoPAVANMETrevl (Program Number NU-840) to transform the data to a joint wind-stability occurrence frequency distribution. The seven wind speed categories were defined according to Regulatory Guide 1.23 with the first category identified as "calm". The higher of the starting speeds of the wind vane and anemometer equipment on each of the towers (i.e. 0.50 mph) was used as the threshold for calm winds, per Regulatory Guide 1.145, Section 1.1. A midpoint was also assumed between each of the Regulatory Guide 1.23 wind speed categories, Nos. 2-6, as to be inclusive of all monitored wind speeds. The Regulatory Guide 1.23 wind speed categories have, therefore, been refined as follows:

DEFINED WIND SPEED CATEGORY RANGES FOR PAVAN MODELING Catego No Regulatory Guide 1.23 Speed PAVAN-Assumed Category No. Interval (mph) Speed Interval (mph) 1 (Calm) 0 to < 1 0 to<0.50 2 1 to 3 >=0.50 to <3.5 3 4 to 7 >=3.5 to <7.5 4 8 to 12 >=7.5 to <12.5 5 13 to 18 >=12.5 to <18.5 6 19 to 24 >=18.5 to <24 7 >24 >=24 The same delta temperature stability class database utilized for the ARCON96 analysis described above was also adopted for the PAVAN analysis.

Attachment 8, Page 19 of 37

Attachment 8 Technical Parameters for AST Calculations The joint lower level wind-stability class occurrence frequency distribution, based on the five-year Tower 1 Modeling Database, is included in the PAVAN input files.

PAVAN Run Scenarios Stack release scenarios were identified for the purpose of applying the PAVAN model using the selected representative meteorological wind and stability class databases to predict the X/Q values that result at the EAB and LPZ. They are shown as follows:

I PAVAN X/Q SCENARIOS RELEASE/RECEPTOR METEOROLOGICAL DATABASE SCENARIOS SCENARIO (Tower ID: Measurement Height above Tower Grade)

Wind Speed and Direction Stability Class (Delta Temperature)

EAB (731 m):.  ! ' -

• North and South Stack Tower 1: 30' Tower 1: 171 - 26' LPZ(2043 mi):' .  ; . ______:__--_________

• North and South Stack Tower 2: 30' l Tower 2: 171 - 26' The Tower 1 Modeling Database is representative for deriving all required meteorological input for the PAVAN modeling of the North and South Stack release X/Q for each subject receptor.

The EAB and LPZ are located at distances of 731 m and 2403 m, respectively. It should be noted that the lower (30 ft) level wind speeds contained in the Tower 1 Modeling Database were used instead of the upper (175 ft) winds, even though it might be otherwise expected that the 175 ft level winds would better represent the 200 ft North and South stack tops. This is because PAVAN requires that any non-elevated release be assumed as a 'ground level' release, which accordingly requires that whatever the release elevation may actually be, it is reassigned a value of 10 meters above station grade. Thus, using actual 10-meter monitored data (i.e., data from the 30 ft level on Tower 1) is considered to be superior to using data from another level (i.e., 175 feet) that PAVAN would subsequently adjust (but imprecisely so, by power law extrapolation) down to 10 meters.

Calculations The X/Q values for the EAB and LPZ calculated by the PAVAN modeling analysis of each release scenario are presented below for each time interval required by NRC Regulatory Guide 1.145.

Attachment 8, Page 20 of 37

Attachment 8 Technical Parameters for AST Calculations Table 9 PAVAN X/Q (sec/M 3 ) Results North And South Stacks to EAB and LPZ LOCATION X/ PARAMETER 0-2 hour 10-8 hour 8-24 hourl 1-4 day 14-30 day

__EAB }731 I j__

3.18E-04 1.76E-04 1.31 E-04 6.89E-05 2.74E-05 North and South Direction-Specific Max (ESE) (ESE) (ESE) (ESE_ (ESE Stacks' Site Limit 2.79E-04 1.58E-04 1.19E-04 6.39E-05 2.63E-05

'_____ ,..,-LPZ 2043m -:

North and South JDirection-Speciic oJ Max 1.15E-04 (ESE) 5.79E-05 4.10E-05 1.95E-05 6.68E-06 (ESE) (ESE) (ESE) (ESE)

Stacks* . l Site Limit 1.01 E-04 5.18E-05 3.71 E-05 1.81 E-05 6.41 E-06

^ The same PAVAN results apply to the North and South Stacks individually.

"The higher of the direction specific and the site limit values are indicated in bold.

SUMMARY

AND CONCLUSIONS The ARCON96 and PAVAN X/Q modeling calculation results are summarized below for the Control Room, EAB and LPZ for the regulated time-averaging periods. Control Room intake results are calculated using the ARCON96 model and the EAB and LPZ results have been calculated using the PAVAN model. All input files for ARCON96 and PAVAN are provided in Attachment 7.

Table 10 X/Q RESULTS

SUMMARY

(sec/M 3)

RECEPTOR  ; RELEASE POINT Yfb0-2 hour 2-8 hour* 8-24 hour 1-4 day 4-30 day Control North Stack 6.88E-03 5.17E-03 2.04E-03 1.29E-03 9.63E-04 Room Intake South Stack 1.26E-03 9.64E-04 3.80E-04 2.39E-04 1.80E-04 EAB North and South Stacks" 3.18E-04 1.76E-04 1.31E-04 6.89E-05 2.74E-05 (731 m) North andSouth Stacks** (ESE) (ESE) (ESE) (ESE) (ESE)

LPZ North and South Stacks* 1.15E-04 5.79E-05 4.1OE-05 1.95E-05 6.68E-06 (2,043 m) andSouthStacks_

N o rt (ESE) (ESE) (ESE) (ESE)

PAVAN result representing 0-8 hour time period.

^^ The same PAVAN results apply to the North and South Stacks individually.

Attachment 8, Page 21 of 37

Attachment 8 Technical Parameters for AST Calculations RADTRAD Input Information Table 11a Primary. ContainmeintLe- ak ge Pat (LO C)'

!RADTRAD -Compartmebnts.:

Compartment 1 2 3 4 Number:

Name: Containment Reactor Enclosure Environment Control Room Type: Other Other Environment Control Room Volume: 403,120 ft3 1,800,000 ft3 N/A 126,000 ft3 Includes Primary 900,000 cfm used Containment and in analysis to Wetwell Airspace account for 50%

mixing credit Source Term 1.0 0.0 0.0 0.0 Fraction:

Compartment Natural Recirculation N/A Recirculation Features: Deposition: Filter: Filter: CREFAS Powers BWR RERS model Radiation mode:

10% (lower bound) 0 to 15.5-min: 2475 cfm.

Flow: 30,000 cfm Manual start in 30 No filtration mi or less credited. HEPA: 80%

Charcoal: 99%

15.5 min to 720 hrs: Chlorine mode:

Flow: 30,000 cfm 3000 cfm.

HEPA: 70% Auto Starts Charcoal: 70% HEPA: 80%

Charcoal: 99%

Comments: Volume No filter credit determined via the during the 15.5-containment minute drawdown leakage program. period.

No elemental 30,000 cfm RERS iodine removal flow provides coefficient. mixing in 50% of the RE volume.

Instantaneous and homogeneous mixing is conservatively assumed.

Attachment 8, Page 22 of 37

Attachment 8 Technical Parameters for AST Calculations Table 1 b

'PrimaryContainment Leakage Pathway OCA)

.RADTRAD Atve Tranfer Pahways i Pathway Number 1 2 34 and Name: Containment to RE Exhaust to CR UnfilteredF andName:_ Reactor Encl. SGTS Node Inleakage CR Filtered Intake From Containment (I) Reactor Environment (3) Environment (3)

Compartment: Enclosure (2)

To Compartment: Reactor Environment (3) Control Room (4) Control Room (4)

Transfer Air Leakage Filtered Exhaust Unfiltered Filtered intake Mechanism: inleakage Transfer Leak rate: Flows: Flow rate: Radiation Mode:

Mechanism Details: 0-24 hr, 0.50% 0-1 min: 9.00E6 0-720 hours, 525 0-30 min, 2100 cfm; per day; cfm cfm - models the 30 min-720 hours, I min to 15.5 min: assumed bounding 525 cfm 24-720 hr, 0.25% 3000 cfm unfiltered per day period); inleakage. Models the normal 15.5 min to 720 CR filtered intake hrs, 2500 cfm. 0% efficiency for flow rate.

(Post-drawdown). all Filter efficiencies:

RERS Filter 99% for aerosol Efficiencies: (HEPA) and; 0%during 0 to 80% for elemental 15.5 min and organic iodines drawdown period. (charcoal).

70% for aerosol (HEPA),

elemental, and Chlorine Mode has organic iodines no filtered intake.

__ (Charcoal)

Comments: The 0-1 min 9.00E6 cfm flow corresponds to 10 air changes per minute to simulate the last minute of RERS startup (first two minutes are before gap release phase).

The I min to 15.5 min 3000 cfm is during the drawdown period (no filtration).

Attachment 8, Page 23 of 37

Attachment 8 Technical Parameters for AST Calculations Table 11c Primary Containment LeakagegPathway"(LOCA)" -I RADTRAD Actiye.Transfer Pathways (Continued)________

Pathway Number CR 5 6 a ndNambe CR Exhaust SGTS Node to and Name: (Equilibrium) Environment From Control Room (4) SGTS Node (5)

Compartment:_______

To Compartment: Environment (3) Environment (3)

Transfer Filtered Exhaust Filter Mechanism: (Maintains equilibrium)

Transfer Rad Mode Flows: Flow rate:

Mechanism Details: 0-30 min, 2625; 0-1 minute, 9.00E6 cfm 30 min-720 hrs, 1050 cfm Flow rate I minute to 15.5 minutes, 3000 cfm Chlorine Mode Flows: for drawdown period.

0-720 hours, 525 cfm Flow rate for 15.5 minutes until 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br />, Filter Efficiency Panel 2500 cfm.

- Filter efficiency is entered as 100.0% for SGTS Filter Efficiency:

all chemical forms of 0% during 0 to 15.5 min iodine, for all time drawdown period; periods. 97.5% for aerosol (HEPA), elemental, and organic iodines (Charcoal).

Comments: This equals the total flow 0-1 minute, 9.00E6 cfm that was taken into the corresponds to 10 air CR volume, which changes per minute to includes inleakage (525 + simulate the last minute 2100, for manual of RERS startup (first initiation time; 525 + 525 two minutes are before cfm, after initiation). gap release phase).

This is the exit from the control room to the environment; the filtration prevents a double counting of the iodine release, although RADTRAD 3.03 documentation indicates that this effect has been eliminated Attachment 8, Page 24 of 37

Attachment 8 Technical Parameters for AST Calculations Table 1Id Prifmtary Containim ent Leakage Pathway (LOCA)

RATRAD Dose Receptor ocati n Dose Location 1 2 3 Number:

EAB (Distance: 731 LPZ (Distance: 2043 In Compartment: Control RoomMersMtr) Meters) Meters)

Breathing Rate 0 to 720 hrs: 3.5E-04 0 to 8 hrs: 3.5E-04 0 to 8 hrs: 3.5E-04 (m

3

/sec): 8 to 24 hrs: 1.8E-04 8 to 24 hrs: 1.8E-04 24 to 720 hrs: 2.3E-04 24 to 720 hrs: 2.3E-04 Occupancy Fractions: 0 to 24 hrs: 1.0 Highest 2-hr window: 1.0 0 to 720 hrs: 1.0 24 to 96 hrs: 0.6 24 to 720 hrs: 0.4 XIQ (sec/M 3 ): 0 to 2 hr: 6.88E-03 3.2-5.2 hrs: 3.18E-04 0 to 8 hr: 5.79E-05 2 to 8 hr: 5.17E-03 (Highest two hours) 8 to 24 hr: 4.10E-05 8 to 24 hr: 2.04E-03 I to 4 day: 1.95E-05 I to 4 day: 1.29E-03 4 to 30 day: 6.68E-06 4 to 30 day: 9.63E-04 Table 1le Primary Containment Leakage Pathwa'y CA)',

RADTRAD Source Terms & Dose, onversion'Factors 3527 MWt CO-58 & Co-60 values are Nuclide Inventory: LGS Specific NIF for 60 MACCS RADTRAD default values isotopes (See Table I of this document)

Release Fractions & RADTRAD Standard BWR-DBA values, Timing: No delay Dose Conversion RADTRAD Library of FGR I1&12 Factors: values for 60 MACCS isotopes Decay & Daughter Enabled - Decay / Daughter Products Products: considered Iodine Chemical NUREG-1465 based Iodine Chemical Fractions: Form Fractions Aerosol: 0.95 Elemental: 0.0485 Organic: 0.0015 Attachment 8, Page 25 of 37

Attachment 8 Technical Parameters for AST Calculations Table 12a

-MSIV Leage -Pathw ( CA)

• ,RADTRAD ComartmnsI____

Compartment 2 3 4 Number: _ _2_3_4 (Node 1) Inboard (Node I) Inboard (Node 2)

Name: Containment MSL "A" MSL "D" Outboard MSL Volume Volume "A" Volume Type: Other Other Other Other 403,120 ft3 Volume: IncludesPrimary 258 ft3 258 ft3 1182 ft3 Containment and Wetwell Airspace Source Term 1.0 0.0 0.0 0.0 Fraction:

Compartment Natural Deposition: None None None Features: Powers BWR model 10% (lower bound)

No elemental iodine removal coefficient.

Comments: Volume determined Minimum steam Minimum steam Minimum steam via containment line piping line piping line piping leakage program. volume from volume from volume from RPV to inboard RPV to inboard inboard MSIV to Instantaneous and MSIV. MSIV. turbine stop homogeneous valve.

mixing is assumed.

Attachment 8, Page 26 of 37

Attachment 8 Technical Parameters for AST Calculations Table 12b

! 'MSIV-Leakage Pathwvay ..(LOCA)..: -

RADTRAD Com Artmenit )(Continued):

Compartment 5 6 9 No.:6789 (Node 2)

Compartment Outboard HP Name: MSL "D" Condenser Environment Control Room Hold" Volume Type: Other Other Environment Control Room Other Volume: 1206 ft3 54,750 ft3 N/A 126,000 ft3 N/A Source Term 0.0 0.0 0.0 0.0 0.0 Fraction:

Compartment None None N/A Recirculation None Features: Filter:

CREFAS Radiation mode: 2475 cfm.

Manual start at 30 min or less HEPA: 99%

Charcoal: 80%

Chlorine mode:

3000 cfm.

Auto Starts HEPA: 99%

Charcoal: 80%

Comments: Minimum HP This steam line Condenser compartment piping shell free-air holds Primary volume from volume. Containment outboard No credit leakage to MSIV to taken for prevent turbine stop deposition in "double-valve. condenser or counting" of on substantial releases.

surface of condenser tubing.

Attachment 8, Page 27 of 37

Attachment 8 Technical Parameters for AST Calculations Table 12c

-: Lea age P a (LO CA)

RDT-RAD Activ Trans'fer Pathas Pathway Number: 1 2 3 4 From Containment Containment Containment Inboard MSL "A" Compartment: piping (node 1)

Inboard piping MSL Inboard piping MSL "D" Outboard MSL "A" To Compartment: "A" volume (node 1) volume (node 1) "Hold" piping (node 2)

Transfer Filter Filter Air leakage Filter Mechanism:

Transfer Flow rates/Efficiencies: Flow rates/Efficiencies: Leakage rate: Flow rate:

Mechanism 0-24hours, 0.834cfm 0-24hours, 0.834cfm 0-24hrs, 0.5% per 0-24hrs, 0.834cfm Details: Aerosol: 0% Aerosol: 0% day Aerosol: 96.8%

Elemental: 0% Elemental: 0% Elemcntal: 39.3 1%

Organic: 0% Organic: 0% 24-720hrs, 0.25% per Organic: 0%

day 24 to 720hours, 24 to 720hours, 24-96hrs, 0.417cfm; 0.417cfm 0.417cfm Aerosol: 98.38%

Aerosol: 0% Aerosol: 0% Elemental: 75.11%

Elemental: 0% Elemental: 0% Organic: 0%

Organic: 0% Organic: 0%

96-720hrs, 0.417cfm Aerosol: 98.38%

Elemental: 95.05%

Organic: 0%

Comments: 50% of first day value 50% of first day value Flow from PC Flow rates were derived based on a based on a leakage other than from the table conservatively assumed conservatively assumed through MSIVs is "Determination of MSL containment pressure at containment pressure at sent to HOLD Decontamination 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. There is no 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. There is no compartment to Factors Due to Iodine filtration of leakage out filtration of leakage out prevent dose Deposition."

of Containment into of Containment into this contribution in this this first piping node. first piping node. run, as it is not a contributor to this MSIV release model.

Attachment 8, Page 28 of 37

Attachment 8 Technical Parameters for AST Calculations Table 12d

MS L-;

--eakage Ptway(LOC)-

- lRADTRAD Acti'v Transfer:Path'ways(Co ntined)>

Pathway Number: 5 6 7 8 From Inboard MSL "D" Outboard MSL A Outboard MSL D CmrtmenIng Vol (De piping volume (node piping volume (node Condenser Compartment: Piping Volume (node 1) 2) 2)

To Compartment: Outboard MSL "D" Condenser Condenser Environment

_________________Piping (node 2)CodneCnesrEvinmn Mechanismf Filter Filter Filter Filter Transfer Flow rates/Efficiencies: Flow Flow Flow rates/Efficiencies:

Mechanism O-24hr, 0.834 cfm; rates/Efficiencies: rates/Efficiencies: 0-24hr, 4.577cfm Details: Aerosol: 96.8% 0-24hr, 0.834 cfm; 0-24hrs, 4.017 cfm; Aerosol: 98.55%

Elemental: 39.3% Aerosol: 98.47% Aerosol: 98.51% Elemental: 98.89%

Organic: 0% Elemental: 38.17% Elemental: 38.63% Organic: 0%

Organic: 0% Organic: 0%

24-96 hr, 0.417 cfm; 24 to 96hr, 2.289cfm Aerosol: 98.38% 24-96 hr, 0.417 cfm; 24-96 hrs, 1.716 cfm; Aerosol: 99.27%

Elemental: 75.11% Aerosol: 77.09% Aerosol: 99.36% Elemental: 99.44%

Organic: 0% Elemental: 75.11% Elemental: 77.43% Organic: 0%

Organic: 0% Organic: 0%

96-720 hr, 0.417 cfm 96-720hr, 2.289cfm Aerosol: 98.38% 96-720 hr, 0.417 cfm 96-720 hrs, 1.302 Aerosol: 99.27%

Elemental: 95.05% Aerosol: 96.58% cfm Elemental: 99.44%

Organic: 0% Elemental: 95.05% Aerosol: 99.51% Organic: 0%

Organic: 0% Elemental: 96.64%

Organic: 0%

Comments: Flow rates were derived Flow rates per Flow rates per Three different cases run.

from the table attached spreadsheet attached spreadsheet No condenser credit, 10%

"Determination of MSL of condenser tube credit, Decontamination 100% tube credit. See Factors Due to Iodine attached spreadsheet Deposition."

Attachment 8, Page 29 of 37

Attachment 8 Technical Parameters for AST Calculations Table 12e MSlV ..Leake Pathwa '(LOCA)- -

'..'RADTR AD Active Transffer Pathwa ys(Continued Pathway Number: 9 10 1I1 From Environment Environment Control Room Compartment:

To Compartment: Control Room Control Room Environment Transfer Filter Filter Filter (Exhaust)

Mechanism:

Transfer Rad Mode Flow rate: Flow Rate: Rad. Mode Flow Mechanism 0-30min, 2100cfm (0% 0-720hrs, 525cfm for Rate:

Details: for all); Rad & Chlorine 0.30min, 2625cfm, 30min-720hrs, 525cfm Modes 30min-720hrs, 99% HEPA; 1050cfm 80% Charcoal Efficiencies: 0% for Chlorine Mode all Flows:

Chlorine Mode Flow 0-720hrs, 525cfm Rate: 0 cfm Efficiencies:

(No filtered intake) 100% for all iodines for all periods Comments: Flow rates were derived Unfiltered Inleakage Flows balance inputs from the table for each time period.

"Determination of MSL Decontamination 100% Filtration Factors Due to Iodine prevents double Deposition." counting of iodine release.

Table 12f

-MSIV Leakage Pathway (LOCA),

,RADTRA.D`Dos'se, Receptor Locations Dose Location 1 2 3 Number:

In Compartment: Control Room EAB (Distance: 731 Meters) Meters)

Breathing Rate 0 to 720 hrs: 3.5E-04 0 to 8 hrs: 3.5E-04 0 to 8 hrs: 3.5E-04 (m3/sec): 8 to 24 hrs: 1.8E-04 8 to 24 hrs: 1.8E-04 24 to 720 hrs: 2.3E-04 24 to 720 hrs: 2.3E-04 Occupancy Fractions: 0 to 24 hrs: 1.0 Highest 2-hr window: 1.0 0 to 720 hrs: 1.0 24 to 96 hrs: 0.6 24 to 720 hrs: 0.4 X/Q (sec/rn3 ): 0 to 2 hr: 6.88E-03 10.4 to 12.4 hrs: 3.18E-04 0 to 8 hr: 5.79E-05 2 to 8 hr: 5.17E-03 (Highest two hours) 8 to 24 hr: 4.10E-05 8 to 24 hr: 2.04E-03 I to 4 day: 1.95E-05 I to 4 day: 1.29E-03 4 to 30 day: 6.68E-06 4 to 30 day: 9.63E-04 Attachment 8, Page 30 of 37

Attachment 8 Technical Parameters for AST Calculations Table 12g

'.. 'MSV Leakage Pah ay .(LOCA ) -"' '2"'

..R-.aD AD Source Terms &DoeConversionFa C s Nuclide Inventory: 3527 MWt CO-58 & Co-60 values are LGS Specific NIF for 60 MACCS RADTRAD default values.

isotopes (See Table 1 of this document)

Release Fractions & RADTRAD Standard BWR-DBA values, Timing: No delay Dose Conversion RADTRAD Library of FGR 1I&12 Factors: values for 60 MACCS isotopes Decay & Daughter Enabled - Decay / Daughter Products Products: considered _

Iodine Chemical NUREG-1465 based Iodine Chemical Fractions: Form fractions Aerosol: 0.95 Elemental: 0.0485 Organic: 0.0015 Attachment 8, Page 31 of 37

Attachment 8 Technical Parameters for AST Calculations Table 13a 9 ECCS LeaKage Pa (OA)  ;; -  ;

RA-DTRAD Compartmen Containment Number: 2_3_4__

Name: ECCS Fluid Reactor Environmcnt Control Room SGTS Node Enclosure _______

Type: Other Other Environment Control Room CRotrol Volume: 959,900 gallons 900,000 ft N/A 126,000 ft3 1 Source Term 1.0 0.0 N/A 0.0 0.0 Fraction: _ N_ _______

Compartment None Recirc. Filter: N/A Recirc. Filter: None Features: RERS Flow Rad Mode Rate: CREFAS Flow 0-15.5min, Rate:

30,000 cfm, no 0-30min, no filter credit filter credit; 15.5min-720hr: 30min-720hrs, 30,000cfm at 2475cfm at 70% efficiency 99% for for aerosols and aerosols, 80%

all iodines. for all iodines.

Chlorine Mode:

3000cfm for all periods at 99%

for aerosols, 80% for all iodines.

Comments: 118,655 ft-_ Compartment Supp. Pool used to model minimum + condition of 9,663 ft3 reactor SoTS filter coolant = train in series 959,900 gallons with RERS filter train.

Attachment 8, Page 32 of 37

Attachment 8 Technical Parameters for AST Calculations Table 13b

-; . ECCSL ePathway t(ih'Ca',

.. ""!-RADTRAD'Active Transfer Patw y Pathway Number 1 2 34 and Name ECCS Fluid to RE CR Filtered Exhaust CR Unfiltered Intake Inleakage From ECCS Fluid (1) Environment (3) Control Room (4) Environment (3)

Compartment:_________

To Reactor Enclosure (2) Control Room (4) Environment (3) Control Room (4)

Compartment:

Transfer Filter (CREFAS) Filter Filter Mechanism: Flow Rate: 5gpm Transfer 98.57% for iodines. CREFAS Flow Rad Mode Flow Flow rate:

Mechanism The "filter is used to rates/Efficiencies: rates/Efficiencies: 0-720 hours, 525 cfm Details: simulate a 1.43% 0-30 min, 2100 0-30 min, 2625cfm; for Radiation Mode and flashing fraction. cfm; Aerosol: 100% Chlorine Mode CR.

Aerosol: 0% Elemental: 100% Aerosol: 0%

Elemental: 0% Organic: 100% Elemental: 0%

Organic: 0% Organic: 0%

30 min-720 hours, 30 min-720 hours, 1050 cfm 525 cfm - Aerosol: I00%

Aerosol: 99% Elemental: 100%

Elemental: 80% Organic: 100%

Organic: 0%

Comments: Since the gallon Models the normal This equals the Unfiltered inleakage volume value was CR filtered intake total flow that was flow rate.

entered as the ECCS flow rate. Filter taken into the CR Fluid volume, efficiencies are volume, which entering a 99% for aerosol includes inleakage gallon/minute value (HEPA) and 80% (525 + 2100, for here is correct. for elemental and manual initiation organic iodines time; 525 + 525 (charcoal). cfm, after initiation).

Chlorine Mode Flow Rate 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br />, 525 cfm -

which balances the inleakage.

Attachment 8, Page 33 of 37

Attachment 8 Technical Parameters for AST Calculations Table 13c

- ECCS LeakagePathw'ay(CA)(C i .'

RADTRADActive- Trnsfr Pathways (fConitinue)-

Path Paha ubr5 day Number RE Exhaust to 6 to SGTS and Name: SGTS Environment From Reactor Enclosure (2) SGTS Node (5)

Compartment:

To SGTS Node (5) Environment (3)

Compartment:

Mechansfe Filter (RERS) Filter (SGTS)

Transfer Flow rate: Flow rate:

Mechanism 0-1 minute, 9.00E6 0-1 minute, 9.00E6 Details: cfm cfm Flow rate I minute to Flow rate I minute to 15.5 minutes, 3000 15.5 minutes, 3000 cfm for drawdown cfm for drawdown period. period.

Flow rate for 15.5 Flow rate for 15.5 minutes until 720 minutes until 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br />, 2500 cfm. hours, 2500 cfm.

RERS Filter SGTS Filter Efficiency is 70% for Efficiency 0% during aerosol (HEPA), 0 to 15.5 minute elemental, and drawdown period, organic iodines 97.5% for aerosol (Charcoal), for (HEPA), elemental, accident duration. and organic iodines (Charcoal).

Comments: 0-1 minute, 9.00E6 0-1 minute, 9.00E6 cfm - 10 air changes cfm - 10 air changes per minute to per minute to simulate last minute simulate last minute of RERS startup (first of RERS startup (first two minutes are two minutes are before gap release before gap release phase). phase).

This is an intermediary pathway to effectively model SGTS filtration in series.

Attachment 8, Page 34 of 37

Attachment 8 Technical Parameters for AST Calculations Table 13d ECCS'Leakge'Pathway 'L;:-' '

RA-DTADDose Receptor Locations:,

Dose Location Number: 1 2 3 In Compartment: Control Room EAB LPZ InComparment:_ConrolRoom(Distance: 731 Meters) (Distance: 2043 Meters)

Breathing Rate 0 to 720 hrs: 3.5E-04 0 to 8 hrs: 3.5E-04 0 to 8 hrs: 3.5E-04 (m3 /sec): 8 to 24 hrs: 1.8E-04 8 to 24 hrs: 1.8E-04 24 to 720 hrs: 2.3E-04 24 to 720 hrs: 2.3E-04 Occupancy Fractions: 0 to 24 hrs: 1.0 Highest 2-hr window: 1.0 0 to 720 hrs: 1.0 24 to 96 hrs: 0.6 24 to 720 hrs: 0.4 X/Q (sec/m 3 ): 0 to 2 hr: 6.88E-03 3.2-5.2 hrs: 3.18E-04 0 to 8 hr: 5.79E-05 2 to 8 hr: 5.17E-03 (Highest two hours) 8 to 24 hr: 4.1OE-05 8 to 24 hr: 2.04E-03 I to 4 day: 1.95E-05 I to 4 day: 1.29E-03 4 to 30 day: 6.68E-06 4 to 30 day: 9.63E-04 Table 13e ECCS Leakage a

,- (LO CA).: ::'

RADTRAD Sourc eTerms &Dose Conversion Faco'r  :

Nuclide Inventory: 3527 MWt lodines only for ECCS leakage.

LGS Specific NIF for 60 MACCS isotopes 1-131: 0.2687E+05 1-132: 0.3881E+05 1-133: 0.5556E+05 I-134: 0.6165E+05 1-135: 0.5192E+05 Release Fractions & RADTRAD Standard BWR-DBA values, Timing: No delay Dose Conversion RADTRAD Library of FGR 11&12 values for Factors: 60 MACCS isotopes Decay & Daughter Enabled - Decay / Daughter Products Products: considered Iodine Chemical NUREG-1465 based Iodine Chemical fractions Since iodines produced by flashing, Fractions: Aerosol: 0.00 treated as 97% elemental and 3%

Elemental: 0.97 organic.

Organic: 0.03 Attachment 8, Page 35 of 37

Attachment 8 Technical Parameters for AST Calculations RADTRAD LOCA Results Summary The following tables are tabulations of doses from the various activity leakage pathways associated with the analyzed DBA-LOCA. The first table shows results for the Radiation Isolation Mode, while the second shows the Chlorine Isolation Mode dose consequences.

Table 14a Radiation Isolation Mode Runs Dose Location .

ActivityControl LAkagevPthwy RooM- EAB LPZ RADTRAD t . (re nTEDE) (rem TEDE) (rem TEDE) Run Output Filename LGS LOCA PC Leak - 70% RERS Filt Credit - 97,5% SGTS Filter- 99%

Primary Containment Leakage 2.9363E+00 8.7994E-01 1.1356E+00 Aerosol, 80% Eand O CR Filter -

30min CREFAS Delay - Rad Mode 525ctm Unfilt Inleak.oO LGS LOCA MSIV Leak - HP Condenser Rem Credit (No Tubes) -

MSIV Leakage 6.9688E-01 1.6823E-02 1.0943E-01 80% CR Charcoal EH with 30min (No Condenser Tube Credit) CREFAS Delay - Rad Mode 525cfm Unfilt Inleak.oO

- J.- LOS LOCA MSIV Leak - HP

- JCondenser Rem Credit (10th of Tubes MSIV Leakage.-:  :  : 2.3471E-01 ' 1.52E-02 9.49-02 .OrganicRemoval)-

- 80%CR (10% Condenser Tube Credit)'- 4- Charcoal Eff with 30min 0 --.

0x CREFAS  ;.....;iey-RMd 55mUfl

• Delay - Rad Mode 525dm Unfilt

____ ___lnleak.oO LGS LOCA MSIV Leak- HP7 Condenser Rem Credit (All of Tubes -

MSIV Leakage 2.6595E-01 1.52E-02 2.809E-02 Organic Removal)es 808% CR Charcda-(100% Condenser Tube Credit) - E3f with 30mnCREFAS Delay - Rd Mode 525cfm Unfilt lnleak.oO LOS LOCA ECCS Leak - 70% RERS Filt Credit - 97.5% SOTS Filter - 991%

ECCS Leakage 4.6595E-02 1.0167E-03 2.8091 E-03 Aerosol, 80% E and O CR Filter -

30mmn CREFAS Delay - Rad Mode Chlorine Isolation Mode Runs InleakoO Dose Location _____

ActiityControlI LAkagevPthwy Room EAB LPZ RADTRAD

,LaaePtwy(rem TEDE) (rem TEDE) (rem TEDE) Run Output Filename LOS LOCA PC Leak - 70H% ERS Fill Credit - 97.5% SOTS Filter - 99%

Primary Containment Leakage 2.49502E+00 8.7994E-01 1.1356E+00 Ae1sol, 80% Eand Filter -

hoCR Mode 525fm Unfilt Inleak.oO 5Chlorine LGS LOCA MSIV Leak - HP MSIV Leakage 6.1828E-01 1.6823E-02 1.0943E-01 Condenser Rem Credit (No Tubes) -

(No Condenser Tube Credit) 80:%CR Charcoal Eff - Chlorine Mode

-525cfm'-. Unfill lnleak.oO

. LGS LOCA MSIV.Leak - HP -

-Condenser Rem Credit (10th of Tubes-MSIV Leakage (10% Condenser Tube Credit)

-3.961 ' 1.6089E-02 9.4223E-02 -Organic Removal 80% CR

- -Chlorine Charcoal Elf Mode 525dm

.. . -:: . .  ; Unfilt 'nleak.oO LOS LOCA MSIV Leak- HP Condenser Rem Credit (All of Tubes MSVeaag 2932-1< 1.242-2 8.1909E-02 ]Organic Removal) - 800%CR Charcoal Eff - Chlorine Mode 525dm Unfilt

• Inleak.oO' LOS LOCA ECCS Leak - 701%RERS Fill Credit - 97.5% SOTS Filter - 99%

ECCS Leakage 3.95722-02 1.01 672-03 2.80912E-03 Aerosol, 80% 2 and OCR Filter -

______________________________________________ ________________________Chlorine Mode 525cfm Unfilt Inleak.oO Attachment 8, Page 36 of 37

Attachment 8 Technical Parameters for AST Calculations Table 15 RADTRAD Bounding LOCA Results By Pathway LOCATION  ;

Control EAB LPZ DOSE CONTRIBUTOR Room (RemTEDE) (Rem TEDE) (Rem TEDE).

Filtered Primary Containment (PC) Leakage (unfiltered for 2.936 0.880 1.136 15.5 minutes, SGTS filtered thereafter) [100% of LA],

Control Room in Rad Mode 0.017 0.109 MSIV Leakage, without LCS but with piping deposition 0.697 0credit. [200 scfh total all MS lines, 100 scfh max/line) 0.047 0.001 0.003 ECCS Leakage in Secondary Containment (SC) [5 gpm]

0.34 Gamma Shine to Control Room General Area 4.02 0.90 1.25 Total Calculated Value 5 25 25 Regulatory Limits Attachment 8, Page 37 of 37