RS-02-174, Co., Llc. (EGC) Application for Amendment to Licenses DPR-19, DPR-25, DPR-29 & DPR-30, Related to Application of Alternative Source Term, Attachments a - E-1

Co., Llc. (EGC) Application for Amendment to Licenses DPR-19, DPR-25, DPR-29 & DPR-30, Related to Application of Alternative Source Term, Attachments a - E-1

ML022940292
Person / Time
Site:	Dresden, Quad Cities
Issue date:	10/10/2002
From:	Jury K Exelon Generation Co, Exelon Nuclear
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
RS-02-174
Download: ML022940292 (194)

Text

Exelkn,.

Exelon Generation www exeloncorp corn Nuclear 4300 Winfield Road Warrenville, IL 60555 10 CFR 50.90 RS-02-174 October 10, 2002 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Dresden Nuclear Power Station, Units 2 and 3 Facility Operating License Nos. DPR-19 and DPR-25 NRC Docket Nos. 50-237 and 50-249 Quad Cities Nuclear Power Station, Units 1 and 2 Facility Operating License Nos. DPR-29 and DPR-30 NRC Docket Nos. 50-254 and 50-265

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 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 (EGC) hereby 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.

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 AST.

Regulatory 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).

October 10, 2002 U. S. Nuclear Regulatory Commission Page 2 EGC 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 Dresden Nuclear Power Station (DNPS), Units 2 and 3, and Quad Cities Nuclear Power Station (QCNPS), Units 1 and 2, were performed following the guidance in References 1 and 2.

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

"* Technical Specifications (TS) and associated Bases revisions to reflect implementation of AST assumptions;

"* TS and associated Bases revisions to increase primary containment allowable leakage;

"* TS and associated Bases revisions to increase main steam isolation valve allowable leakage;

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

"o standby gas treatment, "o secondary containment, "o control room emergency ventilation, and "o control room emergency ventilation air conditioning;

"* TS and associated Bases revisions to change the applicability requirements for the following systems during movement of irradiated fuel assemblies in secondary containment:

o AC sources - shutdown, o DC sources - shutdown, and o Distribution systems - shutdown;

"* TS and associated Bases revisions to reflect use of the standby liquid control system to buffer suppression pool pH to prevent iodine re-evolution during a postulated radiological release;

"* Revisions to increase allowable engineered safety feature leakage;

"* Revisions to include increased unfiltered control room inleakage into the control room envelope; and

"* Development of new offsite and control room atmospheric dispersion factors (X/Qs) calculated using site-specific meteorology data collected between 1995 and 1999.

The proposed changes related to the applicability requirements during movement of irradiated fuel assemblies are consistent with Technical Specification Task Force Traveler

October 10, 2002 U. S. Nuclear Regulatory Commission Page 3 (TSTF)-51, "Revise Containment Requirements During Handling Irradiated Fuel and Core Alterations," Revision 2. TSTF-51, Revision 2, was approved by the NRC on October 15, 1999. TSTF-51 changes the TS operability requirements for engineered safety features such that they are not applicable 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, EGC 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 dropped fuel assembly.

The amendment request is consistent with submittals associated with application of AST that have been previously provided to the NRC by the Nuclear Management Company for the Duane Arnold Energy Center (Letter NG-00-1 589, dated October 19, 2000), the Carolina Power & Light (CP&L) Company for the Brunswick Steam Electric Plant Units 1 and 2 (Letter BSEP 01-0063, dated August 1, 2001), and Energy Northwest for the Columbia Generating Station (Letter G02-01-156, dated December 3, 2001).

This request is subdivided as follows.

1. Attachment A provides a description and safety analysis of the proposed changes.

2. Attachment B consists of tables describing conformance with Regulatory Guide 1.183.

3. Attachment C describes the evaluation performed using the criteria in 10 CFR 50.91(a), "Notice for public comment," paragraph (1), which provides information supporting a finding of no significant hazards consideration using the standards in 10 CFR 50.92, "Issuance of amendment," paragraph (c).

4. Attachment D provides information supporting an environmental assessment.

5. Attachments E-1 and E-2 contain the marked-up TS pages with the proposed changes indicated for DNPS and QCNPS, respectively. A marked-up copy of the affected TS Bases is also included for informational purposes.

6. Attachments F-1 and F-2 contain revised TS pages with the proposed changes incorporated for DNPS and QCNPS, respectively.

The proposed changes have been reviewed by the Plant Operations Review Committees at each of the stations and approved by the respective Nuclear Safety Review Boards in accordance with the requirements of the EGC Quality Assurance Program.

EGC requests approval of the proposed amendments by October 31, 2003. Once approved, the amendments shall be implemented within 60 days. This implementation

October 10, 2002 U. S. Nuclear Regulatory Commission Page 4 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), EGC is notifying the State of Illinois of this application for changes to the TS by transmitting a copy of this letter and its attachments to the designated State Official.

If you have any questions or require additional information, please contact Mr. Kenneth M. Nicely at (630) 657-2803.

Respectfully, Keith R. Jury Director - Licensing Mid-West Regional Operating Group Attachments:

Affidavit Attachment A: Description and Safety Analysis for the Proposed Changes Attachment B: Regulatory Guide 1.183 Comparison Attachment C: Information Supporting a Finding of No Significant Hazards Consideration Attachment D: Information Supporting an Environmental Assessment Attachment E: Marked-Up Technical Specifications and Bases Pages for Proposed Changes Attachment F: Typed Pages for Technical Specifications Changes cc: Regional Administrator - NRC Region III NRC Senior Resident Inspector - Dresden Nuclear Power Station NRC Senior Resident Inspector - Quad Cities Nuclear Power Station Office of Nuclear Facility Safety - Illinois Department of Nuclear Safety

STATE OF ILLINOIS )

COUNTY OF DUPAGE )

IN THE MATTER OF )

EXELON GENERATION COMPANY, LLC

) Docket Numbers DRESDEN NUCLEAR POWER STATION, UNITS 2 AND 3

) 50-237 and 50-249 QUAD CITIES NUCLEAR POWER STATION, UNITS 1 AND 2

) 50-254 and 50-265

SUBJECT:

Request for License Amendments Related to Application of Alternative Source Term AFFIDAVIT I affirm that the content of this transmittal is true and correct to the best of my knowledge, information and belief.

Keith R. Jury Director - Licensing Mid-West Regional Operating Group Subscribed and sworn to before me, a Notary Public in and for the State above named, this (0 -EA _ day of 0 CJ4ý,{,2002.

Attachment A DESCRIPTION AND SAFETY ANALYSIS FOR THE PROPOSED CHANGES A.

SUMMARY

OF THE PROPOSED CHANGES 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 (EGC) requests a change to Appendix A, Technical Specifications (TS), of Facility Operating License Nos. DPR-29 and DPR-30 for the Quad Cities Nuclear Power Station (QCNPS), Units 1 and 2, and DPR-19 and DPR-25 for the Dresden Nuclear Power Station (DNPS), Units 2 and 3. 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," (Reference 1.1) will continue to be used as the radiation dose basis for equipment qualification.

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 AST. Regulatory guidance for the implementation of AST is provided in Reference 1.2. 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).

EGC has performed radiological consequence analyses of the four 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 as described in Reference 1.2. The AST analyses for DNPS and QCNPS were performed following the guidance in References 1.2 and 1.3.

The proposed changes that are justified by the AST analyses include:

"* TS and associated Bases revisions to reflect implementation of AST assumptions;

"* TS and associated Bases revisions to increase primary containment allowable leakage;

"* TS and associated Bases revisions to increase main steam isolation valve allowable leakage;

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

o standby gas treatment, o secondary containment, o control room emergency ventilation, and o control room emergency ventilation air conditioning;

"* TS and associated Bases revisions to change the applicability requirements for the following systems during movement of irradiated fuel assemblies in secondary containment:

o AC sources - shutdown, o DC sources - shutdown, and Page 1 of 48

Attachment A o Distribution systems - shutdown;

"* TS and associated Bases revisions to reflect use of the standby liquid control system to buffer suppression pool pH to prevent iodine re-evolution during a postulated radiological release;

"* Revisions to increase allowable engineered safety feature leakage;

"* Revisions to include increased unfiltered control room inleakage into the control room envelope; and

"* Development of new offsite and control room atmospheric dispersion factors (X/Qs) calculated using site-specific meteorology data collected between 1995 and 1999.

The proposed changes related to the applicability requirements during movement of irradiated fuel assemblies are consistent with Technical Specification Task Force Traveler (TSTF)-51, "Revise Containment Requirements During Handling Irradiated Fuel and Core Alterations," Revision 2.

TSTF-51, Revision 2, was approved by the NRC on October 15,1999. TSTF-51 changes the TS operability requirements for engineered safety features such that they are not applicable 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, EGC 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 dropped fuel assembly.

The proposed TS changes are described in Section E of this Attachment. The marked-up TS pages are provided in Attachment E. A marked-up copy of the affected TS Bases is also included for informational purposes. Revised TS pages reflecting these changes are provided in Attachment F.

B. DESCRIPTION OF THE CURRENT REQUIREMENTS B.1 TS Section 1.1, "Definitions" TS Section 1.1 defines DOSE EQUIVALENT 1-131 as follows.

DOSE EQUIVALENT 1-131 shall be that concentration of 1-131 (microcuries/gram) that alone would produce the same thyroid dose as the quantity and isotopic mixture of 1-131, 1-132, 1-133, 1-134, and 1-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in Table Ill of TID-14844, AEC, 1962, "Calculation of Distance Factors for Power and Test Reactor Sites;" Table E-7 of Regulatory Guide 1.109, Rev. 1, NRC, 1977; or ICRP 30, Supplement to Part 1, pages 192-212, Table titled, "Committed Dose Equivalent in Target Organs or Tissues per Intake of Unit Activity."

B.2 TS Section 3.1.7, "Standby Liquid Control (SLC) System" TS Section 3.1.7 requires that two SLC subsystems shall be operable in Modes 1 and 2.

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Attachment A B.3 TS Section 3.3.6.1, "Primary Containment Isolation Instrumentation" TS Section 3.3.6.1, Table 3.3.6.1-1, "Primary Containment Isolation Instrumentation,"

requires two channels of the SLC System Initiation Function of the Reactor Water Cleanup System isolation instrumentation to be operable in Modes 1 and 2.

B.4 TS Section 3.3.6.2, "Secondary Containment Isolation Instrumentation" TS Section 3.3.6.2, Table 3.3.6.2-1, "Secondary Containment Isolation Instrumentation,"

footnote (b) requires that the instrumentation for the following functions be operable during core alterations and during movement of irradiated fuel assemblies in secondary containment.

Function 3, "Reactor Building Exhaust Radiation - High" Function 4, "Refueling Floor Radiation - High" B.5-a TS Section 3.3.7.1, "Control Room Emergency Ventilation (CREV) System Isolation Instrumentation" (QCNPS only)

TS Section 3.3.7.1, Table 3.3.7.1-1, "Control Room Emergency Ventilation (CREV)

System Isolation Instrumentation," footnote (b) requires that the instrumentation for the following functions be operable during core alterations and during movement of irradiated fuel assemblies in the secondary containment.

Function 4, "Refueling Floor Radiation - High" Function 5, "Reactor Building Ventilation Exhaust Radiation - High" B.5-b TS Section 3.3.7.1, "Control Room Emergency Ventilation (CREV) System Instrumentation" (DNPS only)

TS Section 3.3.7.1 requires that two channels of the Reactor Building Ventilation System - High High Radiation Alarm Function shall be operable during movement of irradiated fuel assemblies in the secondary containment and during core alterations.

B.6 TS Section 3.6.1.3, "Primary Containment Isolation Valves (PCIVs)"

Surveillance Requirement (SR) 3.6.1.3.10 requires verification that the combined leakage rate for all Main Steam Isolation Valve (MSIV) leakage paths is < 46 scfh when tested at > 25 psig at a frequency in accordance with the Primary Containment Leakage Rate Testing Program.

B.7 TS Section 3.6.4.1, "Secondary Containment" TS Section 3.6.4.1 requires that the secondary containment shall be operable during movement of irradiated fuel assemblies in the secondary containment and during core alterations.

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Attachment A B.8 TS Section 3.6.4.2, "Secondary Containment Isolation Valves (SCIVs)"

TS Section 3.6.4.2 requires that each SCIV shall be operable during movement of irradiated fuel assemblies in the secondary containment and during core alterations.

B.9 TS Section 3.6.4.3, "Standby Gas Treatment (SGT) System" TS Section 3.6.4.3 requires that two SGT subsystems shall be operable during movement of irradiated fuel assemblies in the secondary containment and during core alterations. SR 3.6.4.3.1 requires, every 31 days, operation of each SGT subsystem for

> 10 continuous hours with heaters operating.

B.10 TS Section 3.7.4, "Control Room Emergency Ventilation (CREV) System" TS Section 3.7.4 requires that the CREV System shall be operable during movement of irradiated fuel assemblies in the secondary containment and during core alterations.

B.1 1 TS Section 3.7.5, "Control Room Emergency Ventilation Air Conditioning (AC) System" TS Section 3.7.5 requires that the Control Room Emergency Ventilation AC System shall be operable during movement of irradiated fuel assemblies in the secondary containment and during core alterations.

B.12 TS Section 3.8.1, "AC Sources -Operating" SR 3.8.1.21 contains a note that states:

When the opposite unit is in MODE 4 or 5, or moving irradiated fuel assemblies in secondary containment, the following opposite unit SRs are not required to be performed: SR 3.8.1.3, SR 3.8.1.10 through SR 3.8.1.12, and SR 3.8.1.14 through SR 3.8.1.17.

B.13 TS Section 3.8.2, "AC Sources - Shutdown" TS Section 3.8.2 requires that the following AC electrical power sources shall be operable in Modes 4 and 5, and during movement of irradiated fuel assemblies in the secondary containment:

a. One qualified circuit between the offsite transmission network and the onsite Class 1 E AC electrical power distribution subsystem(s) required by LCO 3.8.8, "Distribution Systems - Shutdown"; and

b. One diesel generator (DG) capable of supplying one division of the onsite Class 1 E AC electrical power distribution subsystem(s) required by LCO 3.8.8.

B.14 TS Section 3.8.5, "DC Sources -Shutdown" TS Section 3.8.5 requires that one 250 VDC electrical power subsystem shall be operable to support the 250 VDC and one 125 VDC Class I E electrical power Page 4 of 48

Attachment A distribution subsystems required by LCO 3.8.8, "Distribution Systems - Shutdown" in Modes 4 and 5, and during movement of irradiated fuel assemblies in the secondary containment.

B.15 TS Section 3.8.8, "Distribution Systems- Shutdown" TS Section 3.8.8 requires that the necessary portions of the AC, DC, and the opposite unit's electrical power distribution subsystems shall be operable to support equipment required to be operable in Modes 4 and 5, and during movement of irradiated fuel assemblies in the secondary containment.

B.16 TS Section 5.5.7, "Ventilation Filter Testing Program (VFTP)"

TS Section 5.5.7 requires testing of engineered safety feature filter ventilation systems.

TS Section 5.5.7.c provides methyl iodide penetration acceptance criteria and test conditions for a laboratory test of a sample of the charcoal adsorber. TS Section 5.5.7.e specifies dissipation values for the filter ventilation systems heaters.

B.17-a TS Section 5.5.12, "Primary Containment Leakage Rate Testing Program" (QCNPS only)

TS Section 5.5.12 requires that the maximum allowable primary containment leakage rate, La, at Pa, is 1% of primary containment air weight per day.

B.17-b TS Section 5.5.12, "Primary Containment Leakage Rate Testing Program" (DNPS only)

TS Section 5.5.12 requires that the maximum allowable primary containment leakage rate, La, at Pa, is 1.6% of primary containment air weight per day.

C. BASES FOR THE CURRENT REQUIREMENTS C.1 TS Section 1.1, "Definitions" In 10 CFR 100, "Reactor Site Criteria," Section 100.11, "Determination of exclusion area, low population zone, and population center distance," criteria are provided for evaluating the radiological aspects of the proposed site. A footnote to 10 CFR 100.11 states that the fission product release assumed for these calculations should be based upon a major accident assumed to result in substantial meltdown of the core with subsequent release of appreciable quantities of fission products. Reference 1.1 is cited in 10 CFR 100 as a source of further guidance on these analyses. The current licensing basis for DNPS and QCNPS utilizes the TID 14844 methodology, including the thyroid dose conversion factors used for the calculation of dose equivalent 1-131.

C.2 TS Section 3.1.7, "Standby Liquid Control (SLC) System" The SLC System is designed to provide the capability of bringing the reactor, at any time in a fuel cycle, from full power and minimum control rod inventory to a subcritical condition with the reactor in the most reactive, xenon free state without taking credit for Page 5 of 48

Attachment A control rod movement. The SLC System provides backup capability for reactivity control independent of normal reactivity control provisions provided by the control rods. The system is currently required to be operable while in Modes 1 and 2, since shutdown capability is required in these modes. In Modes 3 and 4, control rods are not able to be withdrawn since the reactor mode switch is in shutdown and a control rod block is applied. This provides adequate controls to ensure that the reactor remains subcritical.

In Mode 5, only a single control rod can be withdrawn from a core cell containing fuel assemblies. Demonstration of adequate shutdown margin ensures that the reactor will not become critical with a single control rod withdrawn.

C.3 TS Section 3.3.6.1, "Primary Containment Isolation Instrumentation" Isolation of the Reactor Water Cleanup System is required when the SLC System has been initiated to prevent dilution and removal of the boron solution by the Reactor Water Cleanup System. Two channels (i.e., one channel per trip system) of the SLC System Initiation Function of the Reactor Water Cleanup System isolation instrumentation are required to be operable in Modes 1 and 2, since these are the only modes where the reactor can be critical. These modes are consistent with the existing applicability statement for the SLC System (i.e., LCO 3.1.7).

C.4 TS Section 3.3.6.2, "Secondary Containment Isolation Instrumentation" High reactor building exhaust radiation or refuel floor radiation is an indication of possible gross failure of the fuel cladding. The release may have originated from a break in the reactor coolant pressure boundary (RCPB) or from a FHA on the refueling floor. The radiation monitoring instrumentation required to be operable by TS 3.3.6.2, Table 3.3.6.2-1, Functions 3 and 4, initiates an isolation of secondary containment and actuation of the SGT System to support actions to limit the release of fission products as assumed in the Updated Final Safety Analysis Report (UFSAR) safety analyses.

These functions are required to be operable during core alterations and during movement of irradiated fuel assemblies in secondary containment because the capability of detecting radiation releases due to fuel failures from dropped fuel assemblies must be provided to ensure that offsite dose limits are not exceeded.

C.5-a TS Section 3.3.7.1, "Control Room Emergency Ventilation (CREV) System Isolation Instrumentation" (QCNPS only)

High radiation in the refueling floor area or in the reactor building ventilation exhaust could be an indication of possible gross failure of the fuel cladding. The release may have originated from a break in the RCPB or from a FHA on the refuel floor. The radiation monitoring instrumentation required to be operable by QCNPS TS 3.3.7.1, Table 3.3.7.1-1, Functions 4 and 5, will automatically initiate isolation of the control room emergency zone upon detection of a high radiation condition, since this radiation release could result in radiation exposure to control room personnel.

These functions are required to be operable during core alterations and during movement of irradiated fuel assemblies in the secondary containment to ensure that control room personnel are protected during a FHA.

Page 6 of 48

Attachment A C.5-b TS Section 3.3.7.1, "Control Room Emergency Ventilation (CREV) System Instrumentation" (DNPS only)

The CREV System is designed to provide a radiologically controlled environment to ensure the habitability of the control room for the safety of control room operators under all plant conditions. The CREV System instrumentation provides control room alarms so that manual action can be taken to start the CREV System and pressurize the control room emergency zone to minimize the consequences of radioactive material in the control room environment. When a Reactor Building Ventilation System - High High Radiation alarm is received, operator action is required to switch the CREV System to the isolation/pressurization mode of operation and close required dampers to maintain the control room emergency zone slightly pressurized with respect to the adjacent zones. CREV System operation ensures that the radiation exposure of control room personnel during postulated accidents does not exceed the limits set by GDC 19 of 10 CFR 50, Appendix A.

TS 3.3.7.1 requires two channels of the Reactor Building Ventilation System - High High Radiation Alarm to be operable during movement of irradiated fuel assemblies in the secondary containment and during core alterations to ensure that control room personnel can be protected during a FHA.

C.6 TS Section 3.6.1.3, "Primary Containment Isolation Valves (PCIVs)"

The function of the PCIVs, in combination with other accident mitigation systems, is to limit fission product release during and following postulated DBAs to within limits. The PCIVs help ensure that an adequate primary containment boundary is maintained during and after an accident, thus providing assurance that primary containment will function as assumed in the safety analyses. The DBAs that result in a release of radioactive material for which the consequences are mitigated by PCIVs are a LOCA and MSLB. In the analysis for each of these accidents, it is assumed that PCIVs are either closed or close within the required isolation times following event initiation. This ensures that potential paths to the environment through PCIVs are minimized.

The analyses for the MSLB and LOCA accidents are based on leakage that is less than the leakage rate specified in SR 3.6.1.3.10 (i.e., a combined leakage rate for all MSIV leakage paths of less than or equal to 46 scfh when tested at greater than or equal to 25 psig). The leakage rate of each MSIV path is assumed to be the maximum pathway leakage (i.e., leakage through the worse of the two isolation valves). The specified leakage rate limit ensures that MSIV leakage is properly accounted for in determining the overall primary containment leakage rate.

C.7 TS Section 3.6.4.1, "Secondary Containment" The function of the secondary containment is to contain, dilute, and hold up fission products that may leak from primary containment following a DBA. In conjunction with operation of the SGT System and closure of certain valves whose lines penetrate the secondary containment, the secondary containment is designed to reduce the activity level of the fission products prior to release to the environment and to isolate and contain fission products that are released during certain operations that take place inside Page 7 of 48

Attachment A primary containment, when primary containment is not required to be operable, or that take place outside primary containment.

Secondary containment is required to be operable during movement of irradiated fuel assemblies in the secondary containment and during core alterations to ensure a control volume is available into which fission products that bypass or leak from primary containment, or are released from the RCPB components located in secondary containment, can be diluted and processed prior to release to the environment. The accident analysis for the FHA takes credit for operability of secondary containment during movement of irradiated fuel assemblies in the secondary containment and during core alterations.

C.8 TS Section 3.6.4.2, "Secondary Containment Isolation Valves (SCIVs)"

The function of the SCIVs, in combination with other accident mitigation systems, is to limit fission product release during and following postulated DBAs. This ensures that fission products that leak from primary containment following a DBA, or that are released during certain operations when primary containment is not required to be operable or take place outside primary containment, are maintained within the secondary containment boundary.

SCIVs are required to be operable during movement of irradiated fuel assemblies in the secondary containment and during core alterations to ensure that an adequate secondary containment boundary is maintained during and after an accident by minimizing potential paths to the environment. The accident analysis for FHA takes credit for operability of the secondary containment boundary, including SCIVs, during movement of irradiated fuel in secondary containment and during core alterations.

Maintaining SCIVs operable ensures that fission products will remain trapped inside secondary containment so that they can be treated by the SGT System prior to discharge to the environment.

C.9 TS Section 3.6.4.3, "Standby Gas Treatment (SGT) System" The function of the SGT System is to ensure that radioactive materials that leak from the primary containment into the secondary containment following a DBA are filtered and adsorbed prior to exhausting to the environment. The system is initiated after a LOCA or FHA to reduce, via filtration and adsorption, the radioactive material released to the environment.

The systems at both DNPS and QCNPS also automatically start and operate in response to actuation signals indicative of conditions or an accident that could require operation of the system.

The SGT System is required to be operable during movement of irradiated fuel assemblies in the secondary containment and during core alterations to limit the release of fission products that leak from the primary containment into the secondary containment following a FHA.

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Attachment A Operating each SGT subsystem for > 10 continuous hours ensures that both subsystems are operable and that all associated controls are functioning properly. It also ensures that blockage, fan or motor failure, or excessive vibration can be detected for corrective action. Operation with the heaters on for > 10 continuous hours every 31 days eliminates moisture on the adsorbers and HEPA filters.

C.10 TS Section 3.7.4, "Control Room Emergency Ventilation (CREV) System" The CREV System provides a radiologically controlled environment from which the unit can be safely operated following a DBA. The control room emergency zone served by the CREV System at DNPS consists of the main control room and the Train B Heating Ventilation and Air Conditioning (HVAC) equipment room. The control room emergency zone served by the CREV System at QCNPS consists of the main control room, cable spreading room, auxiliary electric equipment room, computer room, and the Train B HVAC equipment enclosure. The CREV System is a standby system, parts of which also operate during normal unit operations to maintain the control room emergency zone environment.

The system is designed to minimize infiltration of contaminated air into the control room emergency zone. At DNPS, upon receipt of a reactor building ventilation system high high radiation alarm, indicative of conditions that could result in radiation exposure to control room emergency zone personnel, operator action is required within 40 minutes to switch to the isolation/pressurization mode of operation and close the kitchen and locker room exhaust fan dampers. At QCNPS, the control room emergency zone is automatically isolated upon receipt of an isolation signal.

The CREV System is designed to maintain the control room emergency zone environment for a 30 day continuous occupancy after a DBA without exceeding 5 rem whole body dose or its equivalent to any part of the body. The CREV System applicability requirements ensure that the habitability of the control room emergency zone is maintained following a LOCA, FHA, MSLB, or CRDA, since these DBAs could lead to a fission product release.

C. 1 TS Section 3.7.5, "Control Room Emergency Ventilation Air Conditioning (AC) System" The Control Room Emergency Ventilation AC System provides temperature control for the control room emergency zone following isolation of the control room emergency zone. The Control Room Emergency Ventilation AC System is a single zone system that services only those rooms that are a part of the control room emergency zone as described above for DNPS and QCNPS. The system is designed to maintain the control room emergency zone temperature for a 30 day continuous occupancy following isolation of the control room emergency zone. During emergency operation, the system maintains a habitable environment and ensures the operability of components in the control room emergency zone.

The system applicability requirements ensure that the control room emergency zone temperature will not exceed equipment operability limits following an accident that could result in significant radioactive releases.

Page 9 of 48

Attachment A C.12 TS Section 3.8.1, "AC Sources - Operating" SR 3.8.1.21 of TS Section 3.8.1 is provided to direct that appropriate surveillances for the required opposite unit AC sources are governed by the applicable opposite unit TS.

SR 3.8.1.21 is modified by a note that states:

When the opposite unit is in MODE 4 or 5, or moving irradiated fuel assemblies in secondary containment, the following opposite unit SRs are not required to be performed: SR 3.8.1.3, SR 3.8.1.10 through SR 3.8.1.12, and SR 3.8.1.14 through SR 3.8.1.17.

The note ensures that a given unit SR will not require an opposite unit SR to be performed, when the opposite unit TS exempts performance of an opposite unit SR.

This exemption applies when the opposite unit is in the applicability for TS Section 3.8.2 (i.e., Mode 4 and 5, and during movement of irradiated fuel assemblies in the secondary containment).

C.13 TS Section 3.8.2, "AC Sources - Shutdown" The AC sources are required to be operable during movement of irradiated fuel assemblies in the secondary containment to provide assurance that systems needed to mitigate a fuel handling accident are available. The actions for TS Section 3.8.2 are modified by a note that states LCO 3.0.3 is not applicable. LCO 3.0.3 is not applicable while in Mode 4 or 5. However, irradiated fuel assembly movement can occur in Mode 1, 2, or 3. Ifmoving irradiated fuel assemblies while in Mode 1, 2, or 3, the fuel movement is independent of reactor operations. Entering LCO 3.0.3 while in Mode 1, 2, or 3 would require the unit to be shutdown, but would not require immediate suspension of movement of irradiated fuel assemblies. The note ensures that the actions for immediate suspension of irradiated fuel assembly movement are not postponed due to entry into LCO 3.0.3.

C.14 TS Section 3.8.5, "DC Sources - Shutdown" The DC electrical power sources are required to be operable during movement of irradiated fuel assemblies in the secondary containment to provide assurance that required features needed to mitigate a fuel handling accident are available. The actions for TS Section 3.8.5 are modified by a note that states LCO 3.0.3 is not applicable. LCO 3.0.3 is not applicable while in Mode 4 or 5. However, irradiated fuel assembly movement can occur in Mode 1, 2, or 3. If moving irradiated fuel assemblies while in Mode 1, 2, or 3, the fuel movement is independent of reactor operations. Entering LCO 3.0.3 while in Mode 1, 2, or 3 would require the unit to be shutdown, but would not require immediate suspension of movement of irradiated fuel assemblies. The note ensures that the actions for immediate suspension of irradiated fuel assembly movement are not postponed due to entry into LCO 3.0.3.

C.15 TS Section 3.8.8, "Distribution Systems - Shutdown" The AC and DC electrical power distribution subsystems are required to be operable during movement of irradiated fuel assemblies in the secondary containment to provide Page 10 of 48

Attachment A assurance that systems needed to mitigate a fuel handling accident are available. The actions for TS Section 3.8.8 are modified by a note that states LCO 3.0.3 is not applicable. LCO 3.0.3 is not applicable while in Mode 4 or 5. However, irradiated fuel assembly movement can occur in Mode 1, 2, or 3. If moving irradiated fuel assemblies while in Mode 1, 2, or 3, the fuel movement is independent of reactor operations.

Entering LCO 3.0.3 while in Mode 1, 2, or 3 would require the unit to be shutdown, but would not require immediate suspension of movement of irradiated fuel assemblies. The note ensures that the actions for immediate suspension of irradiated fuel assembly movement are not postponed due to entry into LCO 3.0.3.

C.16 TS Section 5.5.7, "Ventilation Filter Testing Program (VFTP)"

The VFTP establishes the testing requirements for engineered safety feature filter ventilation systems. The testing requirements ensure that both the SGT and CREV Systems filtering systems are capable of controlling the offsite and control room radiological consequences during and following a DBA.

C.17-a TS Section 5.5.12, "Primary Containment Leakage Rate Testing Program" (QCNPS only)

The primary containment is designed to withstand the pressures and temperatures during a DBA LOCA without exceeding the design leakage rate. TS 5.5.12 provides a limit for the maximum allowable primary containment leakage rate. The safety analyses assume a nonmechanistic fission product release following a DBA, which forms the basis for determination of offsite doses. The fission product release is, in turn, based on an assumed leakage rate from the primary containment. Maintaining the maximum leakage rate within the limits of TS 5.5.12 ensures that the leakage rate assumed in the safety analyses is not exceeded.

C.17-b TS Section 5.5.12, "Primary Containment Leakage Rate Testing Program" (DNPS only)

The primary containment is designed to withstand the pressures and temperatures during a DBA LOCA without exceeding the design leakage rate. TS 5.5.12 provides a limit for the maximum allowable primary containment leakage rate. The safety analyses assume a nonmechanistic fission product release following a DBA, which forms the basis for determination of offsite doses. The fission product release is, in turn, based on an assumed leakage rate from the primary containment. Maintaining the maximum leakage rate within the limits of TS 5.5.12 ensures that the leakage rate assumed in the safety analyses is not exceeded.

D. NEED FOR REVISION OF THE REQUIREMENTS The proposed changes to the TS will allow QCNPS and DNPS to apply the results of the plant specific AST analyses using the guidance in Reference 1.2 and meeting the requirements of 10 CFR 50.67. Approval of this change will provide a realistic source term for QCNPS and DNPS that will result in a more accurate assessment of DBA radiological doses. This allows relaxation of some current licensing basis requirements as described in Section E. Adopting the AST methodology may also support future evaluations and license amendments.

Page 11 of 48

Attachment A In addition, this change will allow increasing MSIV leakage. Unplanned MSIV repairs are a significant contributor to increased outage duration and unplanned exposure during refueling outages. This leakage limit is associated with the TS for the operability of the primary containment isolation valves (i.e., TS 3.6.1.3). Revision to this TS will relax operational constraints during outage activities. This relaxation will result in a reduction in personnel exposure due to valve maintenance being performed on the MSIVs. Under the AST assumptions proposed, the MSIV work can be strategically planned to maintain work ALARA and maximize the incremental benefit of work being performed in high dose areas.

Another potential benefit involves the performance of maintenance or repair work on redundant divisionalized safety systems. This work is usually scheduled to ensure that one division is operable while work is performed on the opposite division. Unanticipated problems with the operable division could require the unnecessary suspension of the movement of irradiated fuel or other core alterations, such as control rod drive testing, until the problem is corrected and the system returned to operable status. The proposed change could also facilitate maintenance or repairs on non-redundant portions of systems (e.g., CREV System) without suspending refueling activities.

The implementation of these changes would provide schedule flexibility during outages while maintaining safety margin. For example, moving large equipment into secondary containment in preparation for an outage must be coordinated with TS requirements for secondary containment operability. This limits the method and timing associated with equipment movement, which in turn, can result in delays to certain critical path activities and extend outage duration. This change increases flexibility to move personnel and equipment and perform work which would affect secondary containment operability during the handling of irradiated fuel.

E. DESCRIPTION OF THE PROPOSED CHANGES E.1 TS Section 1.1, "Definitions" The proposed change revises the definition of dose equivalent 1-131 in TS Section 1.1 to remove the word "thyroid" and to add a reference to Federal Guidance Report 11, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion," 1989. This change reflects the application of AST assumptions. The revised MSLB accident analysis uses inhalation committed effective dose equivalent dose conversion factors from Federal Guidance Report 11 for calculation of normalized 1-131 dose equivalent activity.

E.2 TS Section 3.1.7, "Standby Liquid Control (SLC) System" The proposed change revises the applicability of TS Section 3.1.7 to add the requirement for the LCO to be met in Mode 3. This change implements AST assumptions regarding the use of the SLC System to buffer the suppression pool following a LOCA involving significant fission product release. The required actions for Condition C are being revised to add an additional requirement to be in Mode 4.

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Attachment A E.3 TS Section 3.3.6.1, "Primary Containment Isolation Instrumentation" TS Section 3.3.6.1, Table 3.3.6.1-1 lists the applicability requirements for Primary Containment Isolation Instrumentation. The proposed change revises the applicability of the SLC System Initiation Function of the Reactor Water Cleanup System isolation instrumentation to add the requirement for this function to be operable in Mode 3. The revised applicability for this function is consistent with the revised applicability for the SLC System.

E.4 TS Section 3.3.6.2, "Secondary Containment Isolation Instrumentation" The proposed change revises footnote (b) of TS Table 3.3.6.2-1 to eliminate the requirement for Function 3 (i.e., Reactor Building Exhaust Radiation - High) and Function 4 (i.e., Refueling Floor Radiation - High) of the Secondary Containment Isolation Instrumentation to be operable during core alterations. The proposed change also relaxes TS requirements to require these functions to be operable only when handling recently irradiated fuel. Changes to the TS Bases define the time period that must elapse to consider fuel to be beyond recently irradiated. With the application of AST, secondary containment is not credited for the FHA after a 24-hour decay period.

E.5-a TS Section 3.3.7.1, "Control Room Emergency Ventilation (CREV) System Isolation Instrumentation" (QCNPS only)

The proposed change revises footnote (b) of TS Table 3.3.7.7-1 to eliminate the requirement for Function 4 (i.e., Refueling Floor Radiation - High) and Function 5 (i.e.,

Reactor Building Ventilation Exhaust Radiation - High) of the CREV System Isolation Instrumentation to be operable during core alterations. The proposed change also relaxes TS requirements to require these functions to be operable only when handling recently irradiated fuel. Changes to the TS Bases define the time period that must elapse to consider the fuel to be beyond recently irradiated. With the application of AST, the CREV System is not credited for the FHA after a 24-hour decay period.

E.5-b TS Section 3.3.7.1, "Control Room Emergency Ventilation (CREV) System Instrumentation" (DNPS only)

The proposed change deletes "During CORE ALTERATIONS" from the applicability statement for TS LCO 3.3.7.1 and relaxes TS requirements to require LCO 3.3.7.1 to be applicable only when handling recently irradiated fuel. Changes to the TS Bases define the time period that must elapse to consider the fuel to be beyond recently irradiated.

With the application of AST, the CREV System is not credited for the FHA after a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> decay period.

E.6 TS Section 3.6.1.3, "Primary Containment Isolation Valves (PCIVs)"

The proposed change revises SR 3.6.1.3.10 to increase the allowable limit for the combined leakage rate for all MSIV leakage paths from less than or equal to 46 scfh to less than or equal to 144 scfh when tested at greater than or equal to 25 psig. In addition, a new leakage rate limit through each MSIV leakage path is being added. The revised SR 3.6.1.3.10 reads:

Page 13 of 48

Attachment A Verify the leakage rate through each MSIV leakage path is < 57 scfh when tested at

> 25 psig, and the combined leakage rate for all MSIV leakage paths is < 144 scfh when tested at > 25 psig.

The frequency for SR 3.6.1.3.10 is "In accordance with the Primary Containment Leakage Rate Testing Program," and this frequency is not being changed. Application of AST supports the increase in MSIV leakage, as discussed in Section F.

E.7 TS Section 3.6.4.1, "Secondary Containment" The proposed change deletes "During CORE ALTERATIONS" from the applicability statement for TS LCO 3.6.4.1 and relaxes TS requirements to require LCO 3.6.4.1 to be applicable only when handling recently irradiated fuel. Changes to the TS Bases define the time period that must elapse to consider the fuel to be beyond recently irradiated. In addition, the proposed change revises Condition C, and associated required actions and completion times, to reflect the revision of the applicability requirements for LCO 3.6.4.1.

With the application of AST, secondary containment is not credited for the FHA after a 24-hour decay period.

E.8 TS Section 3.6.4.2, "Secondary Containment Isolation Valves (SCIVs)"

The proposed change deletes "During CORE ALTERATIONS" from the applicability statement for TS LCO 3.6.4.2 and relaxes TS requirements to require LCO 3.6.4.2 to be applicable only when handling recently irradiated fuel. Changes to the TS Bases define the time period that must elapse to consider the fuel to be beyond recently irradiated. In addition, the proposed change revises Condition D, and associated required actions and completion times, to reflect the revision of the applicability requirements for LCO 3.6.4.2.

With the application of AST, closure of secondary containment isolation valves is not credited for the FHA after a 24-hour decay period.

E.9 TS Section 3.6.4.3, "Standby Gas Treatment (SGT) System" The proposed change deletes "During CORE ALTERATIONS" from the applicability statement for TS LCO 3.6.4.3 and relaxes TS requirements to require LCO 3.6.4.3 to be applicable only when handling recently irradiated fuel. Changes to the TS Bases define the time period that must elapse to consider the fuel to be beyond recently irradiated. In addition, the proposed change revises Condition C and Condition F, and associated required actions and completion times, to reflect the revision of the applicability requirements for LCO 3.6.4.3. These changes are being made to reflect that with application of AST, the SGT System is no longer required to be operable during movement of irradiated fuel assemblies, that have decayed at least 24-hours, in the secondary containment, or during core alterations, since this system is not credited for the FHA after a 24-hour decay period.

For QCNPS, the proposed change also revises SR 3.6.4.3.1 to remove the requirement for operating the SGT System heaters during performance of the surveillance to operate each SGT subsystem for > 10 continuous hours. As described in Section E.16 of this attachment, TS Section 5.5.7.c is being revised for QCNPS to increase the relative humidity from 70% to 95% for methyl iodide penetration testing. The proposed change Page 14 of 48

Attachment A also revises Section 5.5.7.e to delete the requirement for periodic heater testing. With the lower methyl iodide penetration percentage supported by AST, the requirement to verify operation of the SGT System heaters is being eliminated.

E.10 TS Section 3.7.4, "Control Room Emergency Ventilation (CREV) System" The proposed change deletes "During CORE ALTERATIONS" from the applicability statement for TS LCO 3.7.4 and relaxes TS requirements to require LCO 3.7.4 to be applicable only when handling recently irradiated fuel. Changes to the TS Bases define the time period that must elapse to consider the fuel to be beyond recently irradiated. In addition, the proposed change revises Condition C, and associated required actions and completion times, to reflect the revision of the applicability requirements for LCO 3.7.4.

The AST analyses do not take credit for CREV System operation during movement of irradiated fuel, that has decayed at least 24-hours, in secondary containment, or during core alterations.

E.1 1 TS Section 3.7.5, "Control Room Emergency Ventilation Air Conditioning (AC) System" The proposed change deletes "During CORE ALTERATIONS" from the applicability statement for TS LCO 3.7.5 and relaxes TS requirements to require LCO 3.7.5 to be applicable only when handling recently irradiated fuel. Changes to the TS Bases define the time period that must elapse to consider the fuel to be beyond recently irradiated. In addition, the proposed change revises Condition C, and associated required actions and completion times, to reflect the revision of the applicability requirements for LCO 3.7.5.

AST analyses do not take credit for operation of the Control Room Emergency Ventilation AC System during movement of irradiated fuel, that has decayed at least 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, in secondary containment, or during core alterations.

E.12 TS Section 3.8.1, "AC Sources - Operating" The proposed change revises the note for SR 3.8.1.21 to add "recently" to the phrase referencing movement of irradiated fuel assemblies in secondary containment. This change is needed to be consistent with the proposed change to TS Section 3.8.2, as described below.

E.13 TS Section 3.8.2, "AC Sources - Shutdown" The proposed change relaxes TS requirements to require LCO 3.8.2 to be applicable only when handling recently irradiated fuel. Changes to the TS Bases define the time period that must elapse to consider the fuel to be beyond recently irradiated. In addition, the proposed change revises Required Actions A.2.2 and B.2 to reflect the revision of the applicability requirements for LCO 3.8.2. AST analyses do not take credit for operation of the CREV and SGT Systems, or secondary containment, following a FHA involving movement of irradiated fuel assemblies that have decayed at least 24-hours.

E.14 TS Section 3.8.5, "DC Sources - Shutdown" The proposed change relaxes TS requirements to require LCO 3.8.5 to be applicable only when handling recently irradiated fuel. Changes to the TS Bases define the time Page 15 of 48

Attachment A period that must elapse to consider the fuel to be beyond recently irradiated. In addition, the proposed change revises Required Action A.2.2 to reflect the revision of the applicability requirements for LCO 3.8.5. AST analyses do not take credit for operation of the CREV and SGT Systems, or secondary containment, following a FHA involving movement of irradiated fuel assemblies that have decayed at least 24-hours.

E.15 TS Section 3.8.8, "Distribution Systems-Shutdown" The proposed change relaxes TS requirements to require LCO 3.8.8 to be applicable only when handling recently irradiated fuel. Changes to the TS Bases define the time period that must elapse to consider the fuel to be beyond recently irradiated. In addition, the proposed change revises Required Action A.2.2 to reflect the revision of the applicability requirements for LCO 3.8.8. AST analyses do not take credit for operation of the CREV and SGT Systems, or secondary containment, following a FHA involving movement of irradiated fuel assemblies that have decayed at least 24-hours.

E.16 TS Section 5.5.7, "Ventilation Filter Testing Program (VFTP)"

The proposed change revises Section 5.5.7.c to increase methyl iodide penetration acceptance criteria for the SGT System from 2.5% to 50% and for the CREV System from 0.5% to 5%. Application of AST supports increasing the methyl iodide penetration percentages.

For QCNPS, the proposed change revises test conditions in Section 5.5.7.c to increase the relative humidity from 70% to 95% for methyl iodide penetration testing. The proposed change also revises Section 5.5.7.e to delete the requirement for periodic heater testing. With the lower methyl iodide penetration percentage supported by AST, the need to periodically verify a SGT System heater dissipates the required wattage is eliminated. Although the requirement to verify the heater dissipation is being eliminated, QCNPS has no current plan to remove this heater from the station.

The proposed change to eliminate the requirement to verify a SGT System heater dissipates the required wattage is not being made for DNPS, due to a design difference between DNPS and QCNPS. DNPS vents the High Pressure Coolant Injection System Gland Seal Condenser through SGT, which results in additional moisture in the SGT System. Thus, at DNPS, the verification of heater dissipation is still required to ensure that the heater is capable of adequately removing moisture from the system.

E.17-a TS Section 5.5.12, "Primary Containment Leakage Rate Testing Program" (QCNPS only)

The proposed change increases the maximum allowable primary containment leakage rate, La, at Pa, to 3% of primary containment air weight per day. Application of AST supports the increase in maximum allowable primary containment leakage rate, as discussed in Section F.

Page 16 of 48

Attachment A E.17-b TS Section 5.5.12, "Primary Containment Leakage Rate Testing Program" (DNPS only)

The proposed change increases the maximum allowable primary containment leakage rate, La, at Pa, to 3% of primary containment air weight per day. Application of AST supports the increase in maximum allowable primary containment leakage rate, as discussed in Section F.

F. SAFETY ANALYSIS OF THE PROPOSED CHANGES 1.0 Introduction 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 Reference 1.1, significant advances have been made in understanding the timing, magnitude, and chemical form of fission product releases from severe nuclear power plant accidents. Many of these insights developed out of the major research efforts started by the NRC and the nuclear industry after the accident at Three Mile Island. NUREG-1465 (Reference 1.4) 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 Reference 1.2.

EGC has performed radiological consequence analyses of the four DBAs that result in offsite exposure (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 Page 17 of 48

Attachment A accordance with the guidance in References 1.2 and 1.3. 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 could potentially result in 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

" Evaluation of suppression pool pH to ensure that the particulate 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 Reference 1.2 for the four analyzed DBAs.

2.0 Evaluation 2.1 Scope 2.1.1 Accident Radiological Consequence Analyses The DBA accident analyses documented in the DNPS and QCNPS UFSARs that could potentially result in control room and offsite doses were addressed using methods and input assumptions consistent with the AST. The following DBAs were addressed:

"* CRDA, UFSAR Section 15.4.10;

"* MSLB, UFSAR Section 15.6.4;

"* LOCA, UFSAR Section 15.6.5; and

"* FHA, UFSAR Section 15.7.2 (QCNPS) and 15.7.3 (DNPS).

The analyses were performed in accordance with Reference 1.2 to confirm compliance with the acceptance criteria presented in 10 CFR 50.67.

Page 18 of 48

Attachment A 2.1.2 NUREG-0737, Item lI.B.2 EGC 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.

2.2 Method of Evaluation 2.2.1 Fission Product Inventory The ORIGEN code (Reference 1.5) was used in the calculation of the plant-specific fission product source term inventories. The inventories were determined based on the licensed core power level following extended power uprate (i.e., 2957 megawatts thermal (MWt)) and further adjusted by 102% (3016 MWt) in support of the AST evaluations. The fission product inventory is based on an average core burnup of 1600 effective full power days (EFPD).

2.2.2 Radiological Consequence New calculations were prepared for the simulation of the radionuclide release, transport, removal, and dose estimates associated with the postulated accidents listed in Section 2.1.1.

The RADTRAD computer code (Reference 1.6) was used for these calculations. 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 atmospheric dispersion factors (X/Qs) were calculated using the guidance of Regulatory Guide 1.145 (Reference 1.7) and the PAVAN computer code (Reference 1.8). The code has been used by the NRC in safety reviews.

Control room atmospheric dispersion factors (X/Qs) were calculated with the ARCON96 computer code (Reference 1.9). The application of the code was consistent with the guidance provided in References 1.10 and 1.11. The ARCON96 code calculates relative concentrations in plumes from nuclear power plants at control room air intakes in the vicinity of the release point. Figures 1a and 1b show the layout of intakes and release points for DNPS and QCNPS, respectively. The code has been used by the NRC in safety reviews.

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Attachment A 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.

The post-accident dose rate in the control room and adjacent areas is due to the shine from the refueling floor airborne source. An additional low level post-accident external-source gamma ray dose rate component in the control room and adjacent areas is due to the shine from the refueling floor noble gas airborne source. These dose rates were calculated in the immediate post-TMI-2 period (i.e., 1980-1982) and are included in the UFSAR, Section 12A-3.2.3.4. The shielding provided by the refueling floor slab and the reactor building walls, as well as other concrete structures significant to dose points in the yard, turbine building, control room and radiation monitor dose points, were modeled. The post accident dose rate in the control room due to airborne activity is also directly calculated using the RADTRAD program (Reference 1.6).

In addition to the calculation tools described above, the radiological consequence analyses made use of hand calculations and spreadsheets, supported by appropriate references, to determine inputs and outputs such as plant specific source terms, filter efficiency determinations, and suppression pool pH analyses.

2.3 Inputs and Assumptions 2.3.1 Accident Radiological Consequence Analyses Release Mode The accident analyses were performed for a core inventory based on 3016 MWt (102% of the extended power uprate licensed power level of 2957 MWt) in accordance with Regulatory Guide 1.49 (Reference 1.12).

The reactor core inventory for the analyses was based on an assumed average fuel exposure of 1600 EFPD that develops equilibrium activities in the fuel.

Onsite Meteorological Measurements Program The meteorological measurements program at DNPS and QCNPS consists of monitoring wind direction (WD), wind speed, temperature, and precipitation. Two methods of determining atmospheric stability are used.

The principal method is delta T, which measures the vertical temperature difference. When the delta T method is not available, the sigma theta method is used, which is based on the standard deviation of the horizontal WD. These data, referenced in ANSI/ANS-2.5-1984 Page 20 of 48

Attachment A (Reference 1.13), are used to determine the meteorological conditions prevailing at the plant sites. The meteorological program includes site specific information and calibration procedures.

The meteorological tower is equipped with instrumentation that conforms with the system accuracy recommendations of References 1.13 and 1.14.

The equipment is placed on booms oriented into the generally prevailing wind at the site. Equipment signals are brought to an instrument shack with controlled environmental conditions. The shack at the base of the tower houses the recording equipment, signal conditioners, etc., used to process and retransmit the data to the end-point users.

Recorded meteorological data are used to generate wind roses and to provide estimates of airborne concentrations of gaseous effluents and projected offsite radiation dose. Instrument calibrations and data consistency evaluations are performed routinely to ensure maximum data integrity. Better than 90% data recovery is attained from each measuring and recording system.

Transport Mode Atmospheric dispersion coefficients were calculated, for the identified release paths, based on site-specific meteorology data collected between January 1995 through December 1999. The dispersion coefficients developed represent a change to those used in the current UFSAR analyses. The values currently in the UFSAR are based on Regulatory Guide 1.3 (Reference 1.15) results from the PAVAN code. The Regulatory Guide 1.145 results were used for the offsite atmospheric dispersion coefficients for the AST analyses.

The infiltration of unfiltered air into the control room emergency zone occurs through three different paths: (1) through the emergency zone boundary, (2) through the system components located outside the emergency zone, and (3) through backflow at the zone boundary doors as a result of ingress or egress to or from the emergency zone.

During emergency pressurized modes of operation, the control room ventilation system supplies 2000 scfm of outdoor air to maintain the control room at 1/8-inch water column positive pressure with respect to the adjacent areas. Intentionally admitting outdoor air into the emergency zone facilitates reduction of infiltration through the emergency zone boundary by assuring that air is exfiltrating from the zone at an adequate velocity (i.e., a velocity through the emergency zone boundary to develop and maintain a pressure of 1/8-inch water column).

During the isolation mode, infiltration through the emergency zone boundary is initially negligible because the control room will be at a positive pressure at the time of system isolation. Infiltration following Page 21 of 48

Attachment A isolation is assumed to be 600 cfm for the unfiltered inleakage which includes impact of ingress and egress.

The infiltration through the system components located outside the emergency zone 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 emergency zone from non-habitable areas. The inleakage has been measured in tracer gas tests and determined to be less than 263 cfm. The present analysis assumes a value of 600 cfm.

The opening and closing of boundary doors can induce infiltration to the emergency zone. The backflow infiltration is conservatively assumed at 10 cfm as recommended by Reference 1.16.

Potential adverse interactions between the control room emergency zone and adjacent zones that may allow the transfer of toxic or radioactive gases into the control room are minimized by maintaining the control room at a positive pressure of 1/8-inch water column, with respect to adjacent areas, during emergency pressurized modes. In addition, both the intake dampers and the dampers which isolate the emergency zone are automatically isolated or actuated by the operator in response to the odor of toxic gas or the reactor building ventilation system high radiation alarm.

The standard breathing rates used for control room personnel dose assessments and for the offsite personnel are shown in Table 1. Control room occupancy factors used are also included in Table 1.

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

2.3.1.1 LOCA Inputs and Assumptions The key inputs used in this analysis are included in Tables 2 through 5. These inputs and assumptions are grouped into three main categories (i.e., release, transport, and removal).

LOCA Release Inputs The LOCA analysis assumes the total containment leakage rate at the limit of 3.0% of primary containment air weight per day. The primary containment leakage is reduced by 50% after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, based on the post-LOCA drywell pressure history. No primary containment leakage, with the exception of MSIV leakage, has been identified which bypasses the secondary containment and is released unfiltered to the atmosphere. The analysis assumes the Page 22 of 48

Attachment A maximum MSIV leakage rates of the proposed SR 3.6.1.3.10 to the environment. As discussed in Section E.6 of this attachment, the proposed change revises SR 3.6.1.3.10 to increase the allowable limit for the combined leakage rate for all MSIVs to less than or equal to 144 scfh when tested at greater than or equal to 25 psig. In addition, a leakage rate limit of less than or equal to 57 scfh is being added for each MSIV leakage path, when tested at greater than or equal to 25 psig. These values correspond to a combined leakage rate of 250 scfh and an individual MSIV leakage rate of 100 scfh, when tested at 48 psig. Conservative estimates of MSIV transport times indicated that the leading edge of MSIV leakage would not exit the main steam line prior to 40 minutes after the accident. MSIV leakage to the environment is therefore initiated at T = 40 minutes.

The analysis assumes an engineered safety feature systems leakage rate outside of containment of 2 gpm. Ten percent of the activity in the leakage is assumed to become airborne. This leak rate is consistent with Regulatory Guide 1.183. Although the engineered safety feature systems leakage rate may realistically be assumed to begin approximately 15 minutes following the accident, with the actuation of the drywell sprays, the present analysis conservatively is assumed leakage to begin at the onset of the accident and to continue throughout the 30-day duration of the postulated accident. Prior to 15 minutes there is no engineered safety feature systems operation, hence no leakage, assumed since an ECCS failure is an implicit assumption of the core damage leading to the AST.

Regulatory Guide 1.183 accident isotopic release specification allows deposition of iodine in the suppression pool. Essentially all of the iodine is assumed to remain in solution as long as the pool pH is maintained at or above a level of 7. DNPS and QCNPS procedures will be revised to direct operators, upon detection of symptoms indicating that core damage is occurring (e.g., primary containment high radiation), to manually initiate the SLC System.

The analysis includes the assumptions that: (1) borated solution injection is initiated within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following the accident, and (2) a minimum mass of 3769.4 pounds of sodium pentaborate or equivalent is delivered into the suppression pool. The calculation results demonstrate the buffering effect of the boron solution maintains the suppression pool pH above 7 for the 30-day duration of the postulated LOCA. Maintaining suppression pool pH at or above a level of 7, as an assumption in support of radiological consequence analysis, represents a change to the DNPS and QCNPS design and licensing bases.

Page 23 of 48

Attachment A LOCA Transport Inputs At the beginning of the event, the reactor building exhaust fans are tripped. The reactor building (i.e., secondary containment) is then exhausted by the SGT System continuing the building's negative pressure thus precluding unfiltered exfiltration.

If the main steam lines and the main condenser were to remain intact, the MSIV leakage would eventually collect in the main condenser. However, the present analysis assumes that only the main steam lines between the reactor and the MSIVs remain intact. The analysis further assumes that one of the four outboard MSIVs on the shortest main steam line fails to close. Although the main steam lines from the outboard MSIVs to the condenser and the condenser itself are expected to remain intact, no credit is taken for these in the analysis.

In the analysis, the accident activity was assumed to enter the control room unfiltered for the first 40 minutes of the LOCA at a nominal CREV System filtered ventilation flow rate of 2000 scfm

+/- 10%. After 40 minutes, the CREV System is manually initiated and filtered flow of 2000 scfm +/- 10%, along with 600 cfm of unfiltered leakage, is assumed.

LOCA Removal Inputs The activity of elemental iodine and aerosols released from the core into the drywell is reduced by deposition (i.e., plate-out) and settling in the drywell utilizing the natural deposition values identified in the RADTRAD code. Containment leakage into the reactor building is collected by the SGT System which exhausts the reactor building, via filters, and reduces releases. The deposition removal mechanisms are characteristics of the AST methodology and represent a change in the plant design and licensing basis.

Main steam line pipe deposition was modeled using the RADTRAD code with removal coefficients based on gravitational settling. Five well-mixed control volumes are utilized (i.e., one vertical and four horizontal), with sizing based on convective flow considerations. Gravitational settling is applied to aerosols in horizontal segments. The Cline equations (i.e., basis for Bixler model) are used for organic and elemental iodine plate-out.

No credit is taken for holdup or plate-out in the main steam lines beyond the outboard MSIV in three lines and beyond the inboard MSIV in the fourth line. Additionally, no credit is taken for holdup and plate-out in the main condenser. Main steam line deposition was based on using the shortest line (i.e., most rapid transport) for Page 24 of 48

Attachment A the worst case line (i.e., the one with the assumed failed inboard isolation valve) and the next shortest line of the remaining three lines.

A filter efficiency of 99% for aerosols and 50% for elemental and organic iodine were used in the analysis for the SGT System. For the CREV System, the filter efficiency of 99% was used for aerosols and 95% was used for elemental and organic iodines.

The filter efficiencies are consistent with the proposed DNPS and QCNPS TS.

2.3.1.2 MSLB Accident and Assumptions The key inputs used in this 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. The break mass released includes that amount in the steam line and connecting lines at the time of the break, plus the amount that passes through the valves prior to closure. Two activity release cases corresponding to the pre-accident spike and maximum equilibrium concentration allowed by TS of 4.0 pCi/gm and 0.2 gCi/gm dose equivalent 1-131 respectively were assumed.

These released activity assumptions are consistent with Regulatory Guide 1.183.

The analysis assumes an instantaneous ground level release.

The released reactor coolant and steam is assumed to expand to a hemispheric volume at atmospheric pressure and temperature.

This hemisphere is assumed to move at a speed of one meter per second downwind past the control room intake and then to offsite locations. No credit is taken for buoyant rise of the steam cloud or for operation of the CREV System.

For offsite locations, dispersion is based on the conservative and simplified methodology of Reference 1.17.

2.3.1.3 FHA Inputs and Assumptions The key inputs used in this analysis are included in Table 7.

The postulated FHA involves the drop of a fuel assembly on top of the reactor core during refueling operations. The analysis assumes that 160 fuel pins 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 2.3.1. A post shutdown 24-hour decay period was used to determine the release activity inventory. This assumption is consistent with plant Page 25 of 48

Attachment A procedures, but 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 analysis assumes the fuel bundle is dropped 34 feet, but assumes a water depth of only 19 feet above the assemblies seated in the reactor pressure vessel. The decontamination provided by the 19 feet is determined from guidance in Regulatory Guide 1.183 and is consistent with the limits in the TS.

In accordance with Regulatory Guide 1.183, the analysis assumes that the activity in the reactor building environment is released within two hours, from the reactor building through the reactor building vent stack, as a zero velocity vent release with no further credit for reactor building holdup or dilution, or SGT System operation.

The analysis assumes that the CREV System and control room isolation are not initiated.

2.3.1.4 CRDA Input and Assumptions The key inputs and scenarios used in this analysis are included in Tables 8 and 9.

The CRDA involves the rapid removal of a highest worth control rod resulting in a reactivity excursion that encompasses the consequences of any other postulated CRDA. The core performance analysis shows that the energy deposition that results from this event is inadequate to damage fuel pellets or cladding. However, for the dose consequence analysis, EGC assumed that approximately 1.86% of the fuel pins in the full core were damaged, with melting occurring in 0.77% of the damaged rods (e.g., 0.014% of the core). A core average radial peaking factor of 1.70 was assumed in the analysis.

Three scenarios are considered as shown in Table 9. The limiting case (i.e., Scenario 2 of Table 9) is that of a release which is below the analytic limit of the main steam line radiation monitor (MSLRM) which, if exceeded, trips the mechanical vacuum pump (MVP).

As noted in Table 9, releases to the environment are possible via four pathways. In all scenarios, 99.85% of the activity released from the damaged fuel reaches the turbine and condenser, and 0.15% is assumed to be released directly through the turbine gland seal system (i.e., pathway 1). Releases from the main turbine and condenser are to the turbine building at a rate of one percent per day for a period of 24-hours (i.e., pathway 2) No Page 26 of 48

Attachment A credit is taken for turbine building holdup or dilution and the release from the turbine building is conservatively assumed to be at ground level.

There are two potential forced flow paths from the condenser that also have been evaluated. The first, and the principal CRDA dose contributor, is via the MVP which exhausts to the station chimney from the condenser at a large flow rate when the reactor is at a low power level (i.e., pathway 3). The MSLRM provides a trip function for this pathway. If the MSLRM radiation level exceeds the analytic limit, the MVP will trip and the release through this pathway is stopped and dose consequences are minimal and associated with the one percent per day condenser leak rate (i.e.,

Scenario 1). If dose rate from the release is below the trip setpoint, then the MVP release will continue until it is manually terminated (i.e., Scenario 2). This analytic limit is determined to assure that doses will not exceed 50% of regulatory limits. This manual trip scenario is the limiting case for the CRDA assessment.

The final assessed scenario (i.e., Scenario 3) is for a CRDA at higher power levels when the condenser vacuum is maintained by the steam jet air ejectors (SJAE). This component of the release is via the large charcoal delay beds of the augmented off-gas (AOG) System driven by the SJAE. This pathway would eliminate all iodine releases and greatly delay noble gases (i.e., pathway 4).

The radiological modeling takes credit for securing the MVP pathway if the radiation level at the MSLRM exceeds the analytic limit for this monitor. This analytic limit is determined such that one-half of the limiting regulatory dose limit for the control room, exclusion area boundary (EAB), and low population zone (LPZ) is reached for releases at or below that limit. For the present analysis, the control room dose is limiting. The actual MSLRM setpoint is well below the analytic limit, thus assuring that regulatory limits will not be challenged.

The analysis takes no credit for the filtration of the control room intake air for the duration of the event.

3.0 Results 3.1 Evaluation Results 3.1.1 Accident Radiological Consequence Analyses The postulated accident radiological consequence analyses were reviewed and updated for AST implementation impact.

Page 27 of 48

Attachment A 3.1.1.1 LOCA The radiological consequences of the DBA LOCA were analyzed with the RADTRAD code, using the inputs and assumptions discussed in Section 2.3.1.1. The post-accident doses are the result of four distinct activity releases as discussed below.

Primary to Secondary Containment Leakage This leakage is captured by the secondary containment (reactor building), is filtered by the SGT System, and is then released to the environment through the stations' chimney as an elevated release. No unfiltered exfiltration from the reactor building at any time in the course of a LOCA is considered in this assessment as the DNPS and QCNPS current licensing basis does not take credit for it.

Primary Leakage, Secondary Containment Bypass This pathway considers piping systems from primary containment to points outside of secondary containment and then to the environment. Except for MSIV leakage, no secondary bypass leakage pathways have been identified for DNPS and QCNPS.

Engineered Safety Feature Systems Leakage into the Secondary Containment This leakage is assumed to start immediately after the onset of a LOCA and continue for 30 days. It is filtered by the SGT System prior to release to the environment.

MSIV Leakage from the Primary Containment into the Environment or Turbine Buildinq The MSIV leakage is released as an unfiltered ground level release.

The postulated exposure to the control room occupants includes inleakage internal cloud immersion and inhalation contribution from the primary containment, leakage from engineered safety feature systems, and MSIV leakage releases.

The dose from the following external sources are expected to be much less than 0.5 rem TEDE.

External cloud contribution from the primary containment, secondary containment bypass, engineered safety feature systems, and MSIV leakage releases. This term takes credit for control room structural shielding.

Page 28 of 48

Attachment A

"* A direct dose contribution from the secondary containment contained accident activity. This term takes credit for both reactor building and control room structural shielding.

"* A direct shine from the CREV System and SGT System filters.

The LOCA control room dose corresponds to an assumed unfiltered inleakage rate of 600 cfm. Table 1Oa and 1Ob for DNPS and QCNPS, respectively, present 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.

3.1.1.2 MSLB The radiological consequences of the MSLB accident were analyzed using the inputs and assumptions discussed in Section 2.3.1.2. The doses resulting from a postulated MSLB are given in Tables 11 a and 11 b for DNPS and QCNPS, respectively.

As indicated, the control room, EAB, and LPZ calculated doses are within regulatory limits for AST implementation.

3.1.1.3 FHA The radiological consequences of the FHA were analyzed using the inputs and assumptions discussed in Section 2.3.1.3. The doses resulting from a postulated FHA are given in Tables 12a and 12b for DNPS and QCNPS, respectively. As indicated, the control room, EAB, and LPZ calculated doses are within regulatory limits for AST implementation.

3.1.1.4 CRDA The radiological consequences of the CRDA were analyzed using the inputs and assumptions discussed in Section 2.3.1.4. The doses resulting from a postulated CRDA are given in Tables 13a and 13b for DNPS and QCNPS, respectively. As indicated, the control room, EAB, and LPZ calculated doses are within regulatory limits for AST implementation.

3.1.2 Atmospheric Dispersion Factors Figures la and lb show the release and intake points for DNPS and QCNPS, respectively. The X/Q values for these release-intake combinations are summarized in Tables 14a, 14b, 15a, 15b, 15c, and 15d.

Page 29 of 48

Attachment A Tables 14a and 14b list 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 elevated release X/Q values (i.e., LOCA Chimney release) were calculated using Regulatory Guide 1.145 methodology and include an initial 30-minute fumigation period. These results are based on site specific hourly meteorological data in a five-year period of record.

Tables 15a, 15b, 15c and 15d list X/Q values for the EAB and LPZ boundaries. These were calculated using the PAVAN code and Regulatory Guide 1.145 guidance using a five-year record of site hourly meteorological data.

3.1.3 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 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of the onset of a DBA LOCA and mixed within the suppression pool. The modeling of the DNPS and QCNPS containment cabling maximized the production of hydrochloric acid. The analysis demonstrated that the suppression pool pH 30 days post-LOCA is greater than 7. The final pH and other related parameters are presented in Table 16.

3.2 Evaluation Conclusions As shown in Tables 10 through 13, the plant accident radiological consequence analyses demonstrate that the post-accident offsite and control room doses can be maintained within regulatory limits following AST implementation.

Furthermore, it has determined that continued compliance with NUREG-0737, Item ll.B.2, will be maintained.

4.0. 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 DNPS and QCNPS. The principal changes affect primary containment and MSIV allowable leakage, and elimination of requirements for several systems during movement of irradiated fuel assemblies that have decayed at least 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and during core alterations.

In the dose consequence analyses for the control room occupants, the assumed unfiltered leakage was increased to a value that continues to bound the measured data.

Further evaluation of the analyses performed in support of the AST implementation Page 30 of 48

Attachment A support the conclusion that exposures to onsite and offsite receptors would not result in doses exceeding the values specified in 10 CFR 50.67.

Page 31 of 48

Attachment A Figure Ia: Layout of Intakes and Release Points for DNPS DBA Release Paths LOCA Station chimney and MSIV room CRDA Station chimney and MSIV room FHA Reactor building stack MSLB Steam cloud, transported over CR intake Page 32 of 48

Attachment A Figure 1b: Layout of Intakes and Release Points for QCNPS VN Station Chimney DBA Release Paths LOCA Station chimney and MSIV room CRDA Station chimney and MSIV room FHA Reactor building stack MSLB Steam cloud, transported over CR intake Page 33 of 48

Attachment A

.. Table 1: Personnel Dose Inputs niputeAssuthion .4Val4e Onsite Breathing Rate 3.47E-04 m3 /sec Offsite Breathing Rate 0-8 hours: 3.47E-04 m3 /sec 8-24 hours: 1.75E-04 m 3/sec 1-30 days: 2 32E-04 m 3/sec Control Room Occupancy 0-1 day: 1.0 Factors 1-4 days: 0.6 4-30 days: 0.4 Table'2a,:, Key"Analysis Inputqs and Assumptions Reease6 Inputs - LOCA Radionuclide Source Term 1Input/Assumption, Value Core Fission Product ORIGEN-2 Inventory Only the 60 nuclides considered by RADTRAD are utilized in the analysis Core Power Level 3016 MWt Core Burnup 1600 EFPD Fission Product Release RG 1.183, Table I Fractions for 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 RG 1.183, Table 4 Timing LOCA Release Phases (Per RG 1.183, the release BWRs phases are modeled Phase Onset Duration sequentially) Gap Release 2 m in 0.5 hr Early In-Vessel 0.5 hr 1.5 hr Page 34 of 48

Attachment A Table 2b:,t 'KAy

"'Inp. and' ssumptions Release lInput's Non-LOCA Radionuclide Source Term

!nputlAssumptikon. Value Core Fission Product ORIGEN-2 Inventory Only the 60 nuclides considered by RADTRAD are utilized in the analysis Core Power Level 3016 MWt Core Burnup 1600 EFPD Fission Product Gap RG 1.183, Table 3 Release Fractions for Non- Non-LOCA Fraction of Fission LOCA Accidents 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 Page 35 of 48

Attachment A Table3:A Analysis Inputs and Assumptions

.Release Inputs Primary and Secondary Containment Parameters InputlAs~sump"tion I Value Drywell Free Volume 1.58E+05 cubic feet Surface Area in Drywell 32,250 square feet (DNPS) 32,430 square feet (QCNPS)

Suppression Pool Volume 110,000 cubic feet Primary Containment Total Leak Rate 3.0% per day for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (includes MSIV leakage) 1.50% per day after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Total MSIV leak rate @ 48 psig 250 scfh for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> containment pressure (100 scfh max per line) 50% of that after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Leading edge of MSIV leakage conservatively assumed to exit main steam lines at T = 40 minutes Secondary Containment Volume 4.5E+06 cubic feet (DNPS) 4.7E+06 cubic feet (QCNPS)

Fraction of Secondary 0.5 Containment Available for Mixing SGT System Flow Rate (with 10% margin) 4400 cfm Secondary Containment Drawdown Time 0 Secondary Containment Bypass 0 ESF Systems Leak Rate Outside 2 gpm of Primary Containment (includes factor of 2 margin)

ESF Leakage Duration 0-30 days Release Location ESF/Containment Leakage Station Chimney (elevated release)

MSIV Leakage MSIV Room (ground release)

Release Duration ESFlContainment Leakage 0-30 days MSIV Leakage 40 minutes to 30 days Page 36 of 48

Attachment A Taible 4:,,Key L-OC4AAn~a'l'ysi's'lInputs and Assumptions Transport InpUts -Control Room Parameters Input/Assumption Value Nuclide Release Locations See Figures la and lb CREV System Initiation (manual) 40 minutes after LOCA initiation Control Room Free Volume 81,000 cubic feet (DNPS) 184,000 cubic feet (QCNPS)

CREV System Air Intake Flow Rate 2000 cfm +/- 10%

(normal and accident)

Control Room Unfiltered Inleakage Rate 600 cfm Table"5: Key LOCA Analysis Inputs and Assumptions

• Rem o-Va'l Inp uts Input/?Assum ton Value Aerosol DW Spray Removal Rates N/A Aerosol Natural Deposition Coefficients Credit is taken for natural deposition of Used in the Drywell aerosols based on equations for the Power's model in NUREG/CR 6189 and input directly into RADTRAD as natural deposition time dependent lambdas.

Elemental Iodine Removal in the Drywell Based on Standard Review Plan 6.5.2 methodology, credit is taken until a DF of 200 is reached.

Main Steam Lines Deposition Five well-mixed control volumes utilized (one vertical and four horizontal), with sizing based on convective flow considerations. Gravitational settling applied to aerosols in horizontal segments.

Cline equations (basis for Bixler model) are used for organic and elemental iodine plate-out.

Main Steam Line and Condenser No credit is taken for holdup or plate-out in Holdup Credit for MSIV Leakage the main steam lines beyond the outboard MSIV in three lines and beyond the inboard MSIV in the fourth line. No credit is taken for holdup and plate-out in the main condenser.

Page 37 of 48

Attachment A Table 5: 'Key LOCA Analysis Inputsand Assumptions (cont.)

..Removal Inputs,

-lnputlAssumptikon Value SGT System Filter Efficiency HEPA: Particulate aerosol 99%

Charcoal: Elemental and organic iodine 50%

CREV System Filter Efficiency HEPA: Particulate aerosol 99%

Charcoal: Elemental and organic iodine 95%

Table 6: Key MSLB Accident Analysis Inputs and Assumptions lnputlAssumipption Value Mass Release 20,000 Ibm of steam and 10,125 Ibm of reactor coolant Pre-Accident Spike Iodine Concentration 4 pCi/gm 1-131 equivalent Maximum Equilibrium Iodine Concentration 0.2 pCi/gm 1-131 equivalent Transport model for Control Room Steam cloud moves past the Control Room intake at 1 m/sec Control Room Filtration No Credit Taken Table 7:-Key FHA Analysis Inputs and Assumptilons.

In6p ut/ Assu""m-"ptio~n Value Core Damage 7x7 in a 49 pin bundle plus 111 other pins 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 /> Fuel Pool Water Iodine DF = 135 (19 feet depth)

Decontamination Factor Release Period 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Release Location Reactor building vent stack Unfiltered, zero velocity vent release Dresden: 302 feet from intake Quad Cities: 310 feet from intake CREV System Initiation No Credit Taken Page 38 of 48

Attachment A Table 8:key CR'.DA Anlyi Inputs and Assumptions lnputlAssumption Value Core Damage 850 fuel rods failed (45,612 fuel rods in core)

Percent of Damaged Fuel with Melt 0.77%

Radial Peaking Factor 1.7 Condenser Free Volume Dresden: 55,000 cubic feet Quad Cities: 92,000 cubic feet Mechanical Vacuum Pump Operation MVP flow rate Before Isolation Dresden: 2332 cfm Quad Cities: 5000 cfm MVP discharges to the station chimney MVP isolation within 10 minutes of CRDA initiation Condenser Leak Rate 1% per day Release Period 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Release Location See Figures Ia and lb MVP releases to station chimney Condenser release is at ground level CREV System Initiation Not utilized Charcoal Delay Bed 14.6 days for Xe Noble Gas Delay for SJAE pathway 19.4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for Kr Page 39 of 48

Attachment A ITable 9: CRiD A S6e'n'a'r ios Timing S6e'n a~rio I 'Scenarior 2 Scenario 3, Initial Plant

• In startup/power

• In startup/power 9 In startup/power Condition ascension ascension ascension MVPin

• MVPin 0 SJAEin operation for operation for operation for condenser condenser condenser vacuum vacuum vacuum

• Gland sealing

• Gland sealing 0 Gland sealing steam flow in steam flow in steam flow in operation operation operation CRDA and Activity Full CRDA Small CRDA Full CRDA Transport 0 MSLRM high Possibly,

• No dependence

" CRDA occurs, radiation alarms MSLRM high on MSLRM releasing all in control room radiation alarms response associated within seconds in control room radioactivity to reactor coolant MSLRM high- MSLRM high with a fraction high radiation high radiation quickly released signal initiates analytic limit not MVP shutdown reached, thus no to steam within seconds auto MVP trip

" Activity quickly Reactor Trip and mixes into the other neutron 5000 to 8000 instrument cubic feet reactor dome responses alert and volumesteam line control personnel room of CRDA

" Reactor trips on In response to neutron MSLRM alarms instrumentation and other

" Steam carries indications activity to operator trips condenser, MVP within except for 10 minutes 0.15% which goes to gland sealing steam system Page 40 of 48

Attachment A Table 9:; CRDA Scenarios (cont.)

Timing ý_Sc~enariio 1I Scena'rio,.2 Scenario 3, Release

• 0 15% of activity

• 0.15% of activity ° 0.15% of activity Mechanisms released through released through released gland seal gland seal through gland condenser to condenser to seal condenser station chimney station chimney to station (no delay) (no delay) chimney (no Activity delay)

Balance of activity released exhausted at ° Activity from from condenser high flow from condenser at 1 volume % condenser exhausted per day for through station through AOG 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period, chimney using System, at ground level MVPs for 10 eliminating minutes before iodine releases manual isolation and delaying noble gas Balance of release for decay activity released from condenser at 1 volume %

per day for 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period, at ground level Dose Implications Not calculated, since Design basis This is bounded by this is bounded by MSLRM high-high Scenario 2, which Scenario 2, which analytic limit has a significantly has both selected to control quicker release and significantly larger dose to less than includes iodines and quicker release 50% of regulatory limit Table 1Oa:, Dresden - LA adiological Consequence Analysis

~Regulatory Limit Location ,DurationI n TEDE (rem) TEDE (rem)

Control Room 30 days 4.53* 5 EAB Maximum, 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 1.3 25 LPZ 30 days 0.4 25 The doses here include the direct shine and inhalation doses Trom radioactivity drawn into the control room. The dose from external sources (e.g., refuel floor, passing cloud, and radioactivity accumulated on CREV/SGT System filters is expected to be much less than 0.5 rem TEDE.

Page 41 of 48

Attachment A Tabl6e lýb Quad Cities- LOCA Radiollogical Coriseen4ceAnalysis Rultry Li mit Location Duration TEDE (rem), TEDE(rem)

Control Room 30 days 3.9* 5 EAB Maximum, 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 6.5 25 LPZ 30 days 1.1 25 The doses here include the direct shine and inhalation doses from radioactivity drawn into the control room. The dose from external sources (e.g., refuel floor, passing cloud, and radioactivity accumulated on CREV/SGT System filters is expected to be much less than 0.5 rem TEDE.

Table h11a: *;Dresden ,- MSLB Accident Radiolog!cal'Consequence Analysis I T Dos e 4.0 jiCilgm 0.2 g ICi/gm Dose E'quivalent 1-131 Eq~uivalent 1-131 Regulatory Limit "Location'! Duration TEDE (rem) TEDE (rem) TEDE,(rem)

Control 30-day 2.45E-1 1.23E-2 5 Room integrated dose EAB Worst 4.84E-2 2.42E-3 25 (4.0 gCi/gm) 2-hour 2.5 (02 gCi/gm) integrated dose LPZ 30-day 6.05E-3 3.02E-4 25 (4.0 gCi/gm) integrated 2.5 '02 p dose 2. ( i/gm)

Table l11b: Quad Cities,-MSLB Accident Radiological Consequence Analysis 4.0 pgigm Do6se ý0.2 4pCilgm Dose Equivalent 1-131 Equivalent 1-131 Regulatory Limit Location. Duration' TEDEt (rem) T EDE ,(reim). TEIDE (rem)

Control 30-day 2.45E-1 1.23E-2 5 Room integrated dose EAB Worst 9.50E-2 4.75E-3 25 (4.0 gCi/gm) 2-hour 2.5 (0.2 gCi/gm) integrated dose LPZ 30-day 9.55E-3 4.77E-4 25 (4.0 OCi/gm) integrated 2.5 (0.2 gCi/gm) dose Page 42 of 48

Attachment A

- Table 12a: Dre'sden'- FHA Radiological Cn.sequence Analysis

[wvith 19 fe~et w~ater, c'ove'ra ge]

Regulatory Limit Location Duration TEDEq(rem) TEDE (remn)

Control Room 30 days 2.00 5 EAB Maximum, 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 0.713 6.3 LPZ 30 days 0.072 6.3 Table 12b: Quad Cities - FHA Radiological Consequence Analysis,

[with 1feet water cverage Regulatory Limit Location Duration TEDE (rem) TEE(re m)

Control Room 30 days 1.69 5 EAB Maximum, 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 3.03 6.3 LPZ 30 days 0.181 6.3 Table 13a: Dresden - CRDA Radilogical Consequence Analysis, Regulatory Limit Location Duration. TEDE (rem) TEDE (rem')

Control Room 30 days Less than 2.5 5 EAB Maximum, 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Less than 3 6.3 LPZ 30 days Less than 3 6.3 Tab~lel 3b: Quad Cities CRDA Radiological Consequence Analys is Regul~atory Limit Location Duration: TIEDE (rem) TEDE (rem)

Control Room 30 days Less than 2.5 5 EAB Maximum, 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Less than 3 6.3 LPZ 30 days Less than 3 6.3 Page 43 of 48

Attachment A Table 14a: Dresden 'Control Room XIQ Values for the Different Release and Intake Combinations.',

XIQ (sec/m3); XIQ (sec/m

• Time Time ~ ~ .II Period - IOCA LOCA Period Chimney MSIV. FHA CRDA 0 - 0.5 hrs 4.17E-04 1.24E-03 6.62E-04 0 - 10 min 4.17E-04 0.5 - 2 hrs 1.41 E-08 1.24E-03 6.62E-04 10 min - 2 hrs 1.24E-03 2 - 8 hrs 5.57E-09 1.08E-03 2 - 8 hrs 1.08E-03 8 - 24 hrs 3.50E-09 5.29E-04 8 - 24 hrs 5.29E-04 1 - 4 days 1.28E-09 3 43E-04 -

4-30 days 3.01 E-10 2.72E-04 Notes:

1. Ground level X/Q values are based on ARCON 96; elevated release X/Q values are based on Regulatory Guide 1.145 methodology.

2. Control room intake XIQ values are applicable for control room inleakage.

Table 14b: Quad Cities Control Room X/Q Values for the 1

Different R-elea s e. and Intake Com bination S 2h Tm

-hs 1.50 XIQ (sec/-)

Time~m Time Period LOCA LOCA Perio d

_____Chimney MSIV* FHA CRDA 0 - 0.5 hrs 4.16E-04 1.13E-03 5.58E-04 0 - 10 min 4.16E-04 0.5 -2 hrs 2.35E-09 1.13E-03 5.58E-04 10 min - 2 hrs 1.13E-03 2 -8 hrs 1.15E-09 9.45E-04 22-8 hrs 9.45E-04 8 - 24 hrs 8.02E-10 4.54E-04 8 - 24 hrs 4.54E-04 1 - 4 days 3.96E-10 2.68E-04 4-30 days 1.21 E-100 1.67E-04 Notes:

1. Ground level X/Q values are based on ARCON 96, elevated release X/Q values are based on Regulatory Guide 1.145 methodology

2. Control room intake XIQ values are applicable for control room inleakage.

Page 44 of 48

Attachment A Table I5a: Dresden Elevated Release X/Q (sec/m) Values Using RG 1.145 Methodology for the EAB and LPZ Time Period' A I (sec/rn) PZXQ (sc/rn3) 0 - 0.5 hrs 6.98E-5 (0- 10 min for CRDA) 0.5 - 2 hrs 3.59E-6 0 - 0.5 hrs 8.72E-6 (0- 10 min for CRDA) 0.5 - 2 hrs - 2.48E-6 2 - 8 hrs - 1.17E-6 8 - 24 hrs - 8.08E-7 1-4d - 3.58E-7 4 - 30 d - 1.12E-7 Table 15b6: D resdenbGround Leve!"Release

.IQ (sec/rm3 Values Using RG 1.145 Methodology for the EAB and LPZ,.

J ime Period EAB XIQ (seCrn 3) LPZ XIO (sec/rn) 0 - 2 hrs 2.02E-4 (LOCA and CRDA) 1.85E-4 (FHA) 4.40E-4(MSLB) 0 - 2 hrs 2.10E-5 (LOCA and CRDA) 1.88E-5 (FHA) 5.5E-05 (MSLB) 2 - 8 hrs 9.08E-6 8 - 24 hrs 5.98E-6 1-4d 2.41 E-6 4 - 30 d 6.56E-7 Page 45 of 48

Attachment A

, ,Table 15c: Quad Cities Elevated Release

3) Values Using RG 1.145 Methodology for the EAB and LPZ X/Q (sec/m Time Period EAB X/

• ..(sec/rn) LPZ X/Q (seclm3 )

0 - 0.5 hrs 1.37E-4 (0- 10 min for CRDA) 0.5 - 2 hrs 3.21 E-6 0 - 0.5 hrs - 1.38E-5 (0- 10 min for CRDA) 0.5 - 2 hrs 3.09E-6 2 - 8 hrs 1.52E-6 8 - 24 hrs 1.07E-6 1-4d 4.95E-7 4 - 30 d 1.64E-7

- - Table 15d: Qu'ýd Cities Ground Level release X/Q (sec/m) Values Using RG 1.145 Methodology for the EAB and LPZ Time Period EAB XIQ (seclm 3 ) LPZ %IQ(s'c'ým 3 )

0 - 2 hrs 1.25E-3 (LOCA and CRDA) 7.87E-4 (FHA) 8.64E-4 (MSLB) 0 - 2 hrs 6.68E-5 (LOCA and CRDA) 4.69E-5 (FHA) 8.68E-5 (MSLB) 2 - 8 hrs - 3.07E-5 8 - 24 hrs - 2.08E-5 1 -4d - 8.95E-6 4 - 30 d 2.67E-6 Page 46 of 48

Attachment A Table 16: Suppression Pool pH Results (SLC Systemn

.. Sodium Pentaborate Inventory= 3769.4 Ibm)

Time' pH Initial Pool pH 5.6 0-24 hrs; All Sodium Pentaborate Is in Suppression Pool 30 days 7.53 G. IMPACT ON PREVIOUS SUBMITTALS EGC has reviewed the proposed changes for impact on previous submittals awaiting NRC approval, and has determined that there is no impact on any of them.

H. SCHEDULE REQUIREMENTS EGC requests approval of the proposed changes by October 31, 2003, with a 60-day implementation period.

I. REFERENCES 1.1 U. S. Atomic Energy Commission, Technical Information Document (TID) 14844, "Calculation of Distance Factors for Power and Test Reactor Sites," March 23, 1962 1.2 U. S. Nuclear Regulatory Commission Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors,"

July 2000 1.3 U. S. Nuclear Regulatory Commission Standard Review Plan 15.0.1, "Radiological Consequence Analyses Using Alternative Source Terms," Revision 0, July 2000 1.4 NUREG-1465, "Accident Source Terms for Light-Water Nuclear Power Plants,"

February 1995 1.5 A. G. Croff, "A User's Manual for the ORIGEN 2 Computer Code," ORNL/ITM-7175, Oak Ridge National Laboratory, July 1980 1.6 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 Page 47 of 48

Attachment A 1.7 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 1.8 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 1.9 J. V. Ramsdell and C. A. Simonen, "Atmospheric Relative Concentrations in Building Wakes," NUREG-6331, Revision 1, U. S. Nuclear Regulatory Commission, May 1997 1.10 U. S. Nuclear Regulatory Commission Memorandum from J. L. Birmingham to C. A.

Carpenter, "Summary of February 1-2, 2000, Meeting with the Nuclear Energy Institute (NEI) Regarding Control Habitability and NEI 99-03," March 23, 2000 1.11 Nuclear Energy Institute, NEI 99-03, "Control Room Habitability Assessment Guidance,"

Revision A, May 2001 1.12 U. S. Nuclear Regulatory Commission Regulatory Guide 1.49, "Power Levels of Nuclear Power Plants," Revision 1, December 1973 1.13 ANSI/ANS-2.5-1984, "Standard for Determining Meteorological Information at Nuclear Power Sites" 1.14 U. S. Nuclear Regulatory Commission Safety Guide 23, "Onsite Meteorological Programs," February 17, 1972 1.15 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 1.16 U. S. Nuclear Regulatory Commission Standard Review Plan 6.4, "Control Room Habitability Systems," Revision 2, July 1981 1.17 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 Page 48 of 48

Attachment B REGULATORY GUIDE 1.183 COMPARISON Table 1: Conformance with Regulatory Guide (R1G)1A.83 MVain Se~ctions RG DresdenlPuad Section RG,:P.osition Cities Analysis* 'Comments 3.1 The inventory of fission products in the reactor core and available for Conforms ORIGEN 2 was used to release to the containment should be based on the maximum full power determine core inventory.

operation of the core with, as a minimum, current licensed values for fuel Power level used was enrichment, fuel burnup, and an assumed core power equal to the 3016 MWt to account for current licensed rated thermal power times the ECCS evaluation two% uncertainty (2957 x uncertainty. The period of irradiation should be of sufficient duration to 1.02 = 3016). Fission product allow the activity of dose-significant radionuclides to reach equilibrium or inventory is based on an to reach maximum values. The core inventory should be determined average core burnup of using an appropriate isotope generation and depletion computer code 1600 effective full power such as ORIGEN 2 or ORIGEN-ARP. Core inventory factors (Ci/MWt) days.

provided in TID 14844 and used in some analysis computer codes were derived for low burnup, low enrichment fuel and should not be used with higher burnup and higher enrichment fuels.

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

Page 1 of 27

Attachment B Table,1: Conformance with Regulator Guide(RG) 1.183 Main Sections.

~RG Dresden/Quad Section -:RG Position Cities Analysis Comments 3.1 No adjustment to the fission product inventory should be made for Conforms events postulated to occur during power operations at less than full 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 gap Conforms The fractions from Table 1 are release and early in-vessel damage phases for DBA LOCAs are listed in used.

Table 1 for BWRs and Table 2 for PWRs. These fractions are applied to the equilibrium core inventory described in Regulatory Position 3.1.

Table I 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 Page 2 of 27

Attachment B Table 1: Conformance with Regulatory Guide- (RG) 1.183 Main Sections,

~RG Dresden/Quad Section, RG Position Cities Analy sis Comments 3.2 For non-LOCA events, the fractions of the core inventory assumed to be Conforms Complies with Note 11 of in the gap for the various radionuclides are given in Table 3. The Table 3.

release fractions from Table 3 are used in conjunction with the fission product inventory calculated with the maximum core radial 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 3.3 Table 4 tabulates the onset and duration of each sequential release Conforms The BWR durations from phase for DBA LOCAs at PWRs and BWRs. The specified onset is the Table 4 are used.

time following the initiation of the accident (i.e., time = 0). The early in- LOCA is modeled in a linear vessel phase immediately follows the gap release phase. The activity fashion.

released from the core during each release phase should be modeled as increasing in a linear fashion over the duration of the phase. For non- Non-LOCA is modeled as an LOCA DBAs, in which fuel damage is projected, the release from the instantaneous release.

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 0 sec 0.5 hr 2 min 0.5 hr Early In-Vessel 0.5 hr 1.3 hr 0.5 hr 1.5 hr Page 3 of 27

Attachment B Table 1: Conformance with Regulatory Guide (RG) 1.183 Mai Sectins.

~RG b,resden/Quad Section ,RG, Position .Cities Analysis, Commenits, 3.3 For facilities licensed with leak-before-break methodology, the onset of Not Applicable Dresden and Quad Cities do the gap release phase may be assumed to be 10 minutes. A licensee not use leak-before-break may propose an alternative time for the onset of the gap release phase, methodology for DBA based on facility-specific calculations using suitable analysis codes or on analyses.

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 used.

3.4 Table 5 lists the elements in each radionuclide group that should be Conforms The nuclides used are the considered in design basis analyses. 60 identified as being Table 5 potentially important dose Radionuclide Groups contributors to total effective dose equivalent (TEDE) in the Group Elements RADTRAD code, which Noble Gases Xe, Kr encompasses those listed in Halogens I, Br RG 1.183, Table 5.

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, Np 3.5 Of the radioiodine released from the reactor coolant system (RCS) to the Conforms 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 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 Page 4 of 27

Attachment B

.Table 1: Conformance with Regulatory Guide (RG),.1.183 Main Sections RG DresdenlQuad Section~ RG Position .~Cities Analyis Comments 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 provide additional details.

3.6 The amount of fuel damage caused by non-LOCA design basis events Conforms The current design basis for should be analyzed to determine, for the case resulting in the highest fuel damage from a CRDA is radioactivity release, the fraction of the fuel that reaches or exceeds the not impacted by application of initiation temperature of fuel melt and the fraction of fuel elements for AST and is, therefore, which the fuel clad is breached. Although the NRC staff has traditionally unchanged. The fuel damage relied upon the departure from nucleate boiling ratio (DNBR) as a fuel assumptions correspond to damage criterion, licensees may propose other methods to the NRC those of the Dresden and staff, such as those based upon enthalpy deposition, for estimating fuel Quad Cities Updated Final damage for the purpose of establishing radioactivity releases. Safety Analysis Reports.

4.1.1 The dose calculations should determine the TEDE. TEDE is the sum of Conforms TEDE calculated. Significant the committed effective dose equivalent (CEDE) from inhalation and the progeny included.

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 should be derived from the data provided in ICRP Publication 30, "Limits dose conversion factors for Intakes of Radionuclides by Workers" (Ref. 19). Table 2.1 of Federal (DCFs) are used.

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

Page 5 of 27

Attachment B Table 1: Conformance, with-RegltryGie(R-18 Main Sections RG Dresden/Quad

.Section. RGPosition: Cities Analysis, Comments 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 analysis uses RADTRAD assumed to be 3.5 x 10"4 cubic meters per second. From 8 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> default values (three following the accident, the breathing rate should be assumed to be significant figures), which 1.8 x 10-4 cubic meters per second. After that and until the end of the corresponds to the values in accident, the rate should be assumed to be 2.3 x 104 cubic meters per Section 4.1.3 of RG 1.183.

second.

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

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 EAB. The maximum EAB TEDE for any two-hour period following the start of the radioactivity release should be determined and used in determining compliance with the dose criteria in 10 CFR 50.67. The maximum two-hour TEDE should be determined by calculating the postulated dose for a series of small time increments and performing a "sliding" sum over the increments for successive two-hour periods. The maximum TEDE obtained is submitted. The time increments should appropriately reflect the progression of the accident to capture the peak dose interval between the start of the event and the end of radioactivity release (see also Table 6).

Page 6 of 27

Attachment B

'Table ,1: Conformance with Regulatory Guide (RG)1.183 Main Sections * ..

RG~ Die'sdenlQuad "Section' RG Position Cities Analysis Commentss 4.1.6 TEDE should be determined for the most limiting receptor at the outer Conforms 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 deposition on the ground.

4.2.1 The TEDE analysis should consider all sources of radiation that will Conforms The principal source of dose cause exposure to control room personnel. The applicable sources will within the control room is due vary from facility to facility, but typically will include: to airborne activity. The dose estimates from post LOCA sources erna tost co

"* Contamination of the control room atmosphere by the intake or sources external to the control infiltration of the radioactive material contained in the radioactive 14844 source terms. The plume released from the facility,

"* Contamination of the control room atmosphere by the intake or principal source from external infiltration of airborne radioactive material from areas and structures sources is shine from noble adjacent to the control room envelope, gases in the reactor building "abovethe refueling floor. The tRadiation shine from the external radioactive plume released from noble gas source term of TID the facility, 14844 is not appreciably

"* Radiation shine from radioactive material in the reactor containment, different from the AST source term. Therefore, the shine dose contribution is expected

"* Radiation shine from radioactive material in systems and to be comparable in both components inside or external to the control room envelope, e.g., cases.

radioactive material buildup in recirculation filters.

4.2.2 The radioactive material releases and radiation levels used in the control Conforms The source term, transport, room dose analysis should be determined using the same source term, and release methodology is transport, and release assumptions used for determining the EAB and the same for both the control the LPZ TEDE values, unless these assumptions would result in non- room and offsite locations.

conservative results for the control room.

Page 7 of 27

Attachment B Table .,:IConformance with Regulatory Guide(RG) 1,183 Man Sections RG Dresden/Quad

~Section~ RG Position C6ities Analy~s'is. Comments~

4.2.3 The models used to transport radioactive material into and through the Conforms 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 radioactive Conforms Pressurization and intake material within the control room may be assumed. Such features may filtration are credited in the include control room isolation or pressurization, or intake or recirculation LOCA accident analysis. No filtration. Refer to Section 6.5.1, "ESF Atmospheric Cleanup System," of credit is taken in the MSLB, the SRP (Ref. 3) and Regulatory Guide 1.52, "Design, Testing, and FHA, and CRDA accident Maintenance Criteria for Post-accident Engineered-Safety-Feature analyses.

Atmosphere Cleanup System Air Filtration and Adsorption Units of Light Water-Cooled Nuclear Power Plants" (Ref. 25), for guidance.

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

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 Based on RADTRAD default exposed individual who is present in the control room for 100% of the values (three significant 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 figures), which corresponds to 1 and 4 days, and 40% of the time from 4 days to 30 days. For the the values in Section 4.2.6 of duration of the event, the breathing rate of this individual should be RG 1.183.

assumed to be 3.5 x 104 cubic meters per second.

4.2.7 Control room doses should be calculated using dose conversion factors Conforms The equation given is utilized identified in Regulatory Position 4.1 above for use in offsite dose for finite cloud correction analyses. The DDE from photons may be corrected for the difference when calculating external between finite cloud geometry in the control room and the semi-infinite doses due to the airborne cloud assumption used in calculating the dose conversion factors. The activity inside the control following expression may be used to correct the semi-infinite cloud dose, room.

DDE., to a finite cloud dose, DDEfn,t,, where the control room is Page 8 of 27

Attachment B Table 1: Conformance with Reg ulatory,GUide (RG),1.183 Maine 6ir*Sectionss RG Dresden/Quad Section RG Position Ci.ties.Anal.sis..Comm.ents .

modeled as a hemisphere that has a volume, V, in cubic feet, equivalent to that of the control room (Ref. 22).

338 finite= DDE_VO DDE 1173 4.3 The guidance provided in Regulatory Positions 4.1 and 4.2 should be Conforms A qualitative assessment of used, as applicable, in re-assessing the radiological analyses identified the regulatory positions on in Regulatory Position 1.3.1, such as those in NUREG-0737 (Ref. 2). source term indicate that with Design envelope source terms provided in NUREG-0737 should be no new operator actions updated for consistency with the AST. In general, radiation exposures required in areas such as to plant personnel identified in Regulatory Position 1.3.1 should be ECCS pump rooms, doses expressed in terms of TEDE. Integrated radiation exposure of plant would be lower than currently equipment should be determined using the guidance of Appendix I of reported.

this guide.

5.1.1 The evaluations required by 10 CFR 50.67 are re-analyses of the design Conforms basis safety analyses and evaluations required by 10 CFR 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 classified Conforms The analyses take credit for as safety-related, are required to be operable by technical specifications, SLC System operation. The are powered by emergency power sources, and are either automatically SLC System is safety-related, actuated or, in limited cases, have actuation requirements explicitly required to be operable by addressed in emergency operating procedures. The. single active technical specifications, and component failure that results in the most limiting radiological powered by emergency consequences should be assumed. Assumptions regarding the power. The SLC System is Page 9 of 27

Attachment B Tablel: 'Conformance with Regulatory Guide,(RG) 1.183 Main Sections RG - Dr sd'n/Quad

ýSection GPsto Cities Analysis Comments occurrence and timing of a loss of offsite power should be selected with manually initiated from the the objective of maximizing the postulated radiological consequences. main control room, as directed by the emergency operating procedures.

5.1.3 The numeric values that are chosen as inputs to the analyses required Conforms 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 nonconservative in another portion of the same analysis.

5.1.4 Licensees should ensure that analysis assumptions and methods are Conforms compatible with the AST and the TEDE criteria.

5.3 Atmospheric dispersion values (X/Q) for the EAB, the LPZ, and the Conforms ARCON96 was used to control room that were approved by the staff during initial facility determine control room licensing or in subsequent licensing proceedings may be used in atmospheric dispersion performing the radiological analyses identified by this guide. values for ground level releases and for reactor Methodologies that have been used for determining X/Q values are building stack releases.

documented in Regulatory Guides 1.3 and 1.4, Regulatory Guide 1.145, RG 1.145 methodology was "Atmospheric Dispersion Models for Potential Accident Consequence used to determine EAB and Assessments at Nuclear Power Plants," and the paper, "Nuclear Power LPZ atmospheric dispersion Plant Control Room Ventilation System Design for Meeting General values and control room Criterion 19" (Refs. 6, 7, 22, and 28). atmospheric dispersion The methodology of the NRC computer code ARCON96 (Ref 26) is values for elevated releases.

generally acceptable to the NRC staff for use in determining control room X/Q values.

Page 10 of 27

Attachment B Table'2: Conformance withRG 1.183 Appendix A(Loss-of oolant Accident)

RG ~Dresde'n/Quiad Section -RGPosition Citi els Analysi's Comrments Acceptable assumptions regarding core inventory and the release of Conforms radionuclides from the fuel are provided in Regulatory Position 3 of this guide.

2 Ifthe sump or suppression pool pH is controlled at values of 7 or Conforms The stated distributions of greater, the chemical form of radioiodine released to the containment iodine chemical forms are should be assumed to be 95% cesium iodide (Csl), 4.85 percent used.

elemental iodine, and 0.15 percent organic iodide. Iodine species, An evaluation has been including those from iodine re-evolution, for sump or suppression pool completed to demonstrate pH values less than 7 will be evaluated on a case-by-case basis. that the long term post-LOCA Evaluations of pH should consider the effect of acids and bases created suppression pool pH is during the LOCA event, e.g., radiolysis products. With the exception of greater than 7.

elemental and organic iodine and noble gases, fission products should be assumed to be in particulate form.

3.1 The radioactivity released from the fuel should be assumed to mix Conforms Radioactivity released from instantaneously and homogeneously throughout the free air volume of the fuel is assumed to mix the primary containment in PWRs or the drywell in BWRs as it is instantaneously and released. This distribution should be adjusted if there are internal homogeneously throughout compartments that have limited ventilation exchange. The suppression the free air volume of the pool free air volume may be included provided there is a mechanism to drywell as it is released.

ensure mixing between the drywell to the wetwell. The release into the containment or drywell should be assumed to terminate at the end of the early in-vessel phase.

3.2 Reduction in airborne radioactivity in the containment by natural Conforms Credit is taken for elemental deposition within the containment may be credited. Acceptable models iodine plate-out within the for removal of iodine and aerosols are described in Chapter 6.5.2, drywell for the first 3.1 hours1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> "Containment Spray as a Fission Product Cleanup System," of the (i.e., until a DF of 200 is Standard Review Plan (SRP), NUREG-0800 (Ref. A-I) and in reached) following the NUREG/CR-6189, "A Simplified Model of Aerosol Removal by Natural accident using SRP 6.5.2 Processes in Reactor Containments" (Ref. A-2). The latter model is methodology.

Page 11 of 27

Attachment B Table 2: Conformance with RG1.183Appendix ss-Coolt Acci dnt)

RG' resdenlQuad Section RG-Position C61ities',Analysis Comments incorporated into the analysis code RADTRAD (Ref. A-3). Credit is taken for natural deposition of aerosols based on equations for the Power's model in NUREG/CR 6189 and input directly into RADTRAD as natural deposition time dependent lambdas.

3.3 Reduction in airborne radioactivity in the containment by containment Not Applicable No credit is taken for drywell spray systems that have been designed and are maintained in sprays accordance with Chapter 6.5.2 of the SRP (Ref. A-I) may be credited.

Acceptable models for the removal of iodine and aerosols are described in Chapter 6.5.2 of the SRP and NUREG/CR-5966, "A Simplified Model of Aerosol Removal by Containment Sprays"l (Ref. A-4). This simplified model is incorporated into the analysis code RADTRAD (Refs. A-1 to A-3).

3.3 The evaluation of the containment sprays should address areas within Not Applicable No credit is taken for drywell the primary containment that are not covered by the spray drops. The sprays.

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.

3.3 The SRP sets forth a maximum decontamination factor (DF) for Not Applicable No credit is taken for drywell elemental iodine based on the m6ximum iodine activity in the primary sprays.

containment atmosphere when the sprays actuate, divided by the I activity of iodine remaining at some time after decontamination. The I Page 12 of 27

Attachment B Table 2": Conformance with RG 1.83 Appendix A (Loss-of-Coolant Accident)

RG:,f D.esden/Q.uad.

Section RG, Position Cities Ayisi Comments 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 oarticulate in SRP methodology).

3.4 Reduction in airborne radioactivity in the containment by in-containment Not Applicable Dresden and Quad Cities recirculation filter systems may be credited if these systems meet the have no in-containment guidance of Regulatory Guide 1.52 and Generic Letter 99-02 (Refs. A-5 recirculation filter systems.

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 Pool scrubbing not credited.

pool scrubbing in BWRs should generally not be credited. However, the staff may consider such reduction on an individual case basis. The evaluation should consider the relative timing of the blowdown and the fission product release from the fuel, the force driving the release through the pool, and the potential for any bypass of the suppression pool (Ref. 7). Analyses should consider iodine re-evolution if the suppression pool liquid pH is not maintained greater than 7.

3.6 Reduction in airborne radioactivity in the containment by retention in ice Not Applicable Dresden and Quad Cities do condensers, or other engineering safety features not addressed above, not have ice condensers.

should be evaluated on an individual case basis. See Section 6.5.4 of the SRP (Ref. A-1).

3.7 The primary containment (i.e., drywell for Mark I and II containment Conforms. Leakage reduced after designs) should be assumed to leak at the peak pressure technical 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> based upon reduced 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 Page 13 of 27

Attachment B Table 2:, Conformance with RG 1.183 Appendix A (Loss-of-Coolant Accident)

RG Dresdenlduad Section RG, Position' Cities Anal ys'is Clomments 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 containment pressure.

specification leak rate. For BWRs, leakage may be reduced after the Primary containment pressure 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 not brought subatmospheric.

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.

3.7 For BWRs with Mark III containments, the leakage from the drywell into Not Applicable Dresden and Quad Cities the primary containment should be based on the steaming rate of the have Mark I containments.

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 Ifthe primary containment is routinely purged during power operations, Not Applicable Dresden and Quad Cities releases via the purge system prior to containment isolation should be containments are inerted and analyzed and the resulting doses summed with the postulated doses not routinely purged during from other release paths. The purge release evaluation should assume power operations.

that 100% of the radionuclide inventory in the reactor coolant system 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 SGT System discharges to collected, processed by engineered safety feature (ESF) filters, if any, the station chimney, which is and released to the environment via the secondary containment exhaust described in the UFSARs for Page 14 of 27"

Attachment B Table 2: Conformance with RG 1,183 Appendix A (Loss-of-Coolant Accident)

~RG DresdirnfQuad Section RG Position Cities Analysis Comments, system during periods in which the secondary containment has a Dresden and Quad Cities as negative pressure as defined in technical specifications. Credit for an an elevated release point.

elevated release should be assumed only if the point of physical release is more than two and one-half times the height of any adjacent structure.

4.2 Leakage from the primary containment is assumed to be released Conforms The present licensing design directly to the environment as a ground-level release during any period basis assumes that in which the secondary containment does not have a negative pressure secondary containment is as defined in technical specifications. always negative and no exfiltration will occur in the LOCA accident sequence.

4.3 The effect of high wind speeds on the ability of the secondary Conforms The existing design basis containment to maintain a negative pressure should be evaluated on an assumptions as described in individual case basis. The wind speed to be assumed is the 1-hour Section 6.2.3.1 of the average value that is exceeded only 5% of the total number of hours in Updated Final Safety Analysis the data set. Ambient temperatures used in these assessments should Reports for Dresden and be the 1-hour average value that is exceeded only 5% or 95% of the Quad Cities were used.

total numbers of hours in the data set, whichever is conservative for the intended use (e.g., if high temperatures are limiting, use those exceeded only 5%).

4.4 Credit for dilution in the secondary containment may be allowed when Conforms Fifty percent of the reactor adequate means to cause mixing can be demonstrated. Otherwise, the building volume is credited for leakage from the primary containment should be assumed to be dilution due to mixing caused transported directly to exhaust systems without mixing. Credit for by SGT System operation.

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.

Page 15 of 27

Attachment B "Table2: Conformance with RG1.183 Appendix A (Loss-of-Coolant Accident)

~RG' Dres'denlQuad Section RG Position Cities Analysis' Comments,.

4.5 Primary containment leakage that bypasses the secondary containment Conforms Except for MSIV leakage, no should be evaluated at the bypass leak rate incorporated in the technical secondary containment specifications. Ifthe bypass leakage is through water, e.g., via a filled bypass has been identified.

piping run that is maintained full, credit for retention of iodine and aerosols may be considered on a case-by-case basis. Similarly, deposition of aerosol radioactivity in gas-filled lines may be considered on a case-by-case basis.

4.6 Reduction in the amount of radioactive material released from the Conforms Credit is taken for the secondary containment because of ESF filter systems may be taken into performance of the SGT account provided that these systems meet the guidance of Regulatory System charcoal and HEPA Guide 1.52 (Ref. A-5) and Generic Letter 99-02 (Ref. A-6). filters. The SGT System meets the guidance of RG 1.52 and Generic Letter 99-02.

5.1 With the exception of noble gases, all the fission products released from Conforms Fission products mixed into the fuel to the containment (as defined in Tables 1 and 2 of this guide) suppression pool during should be assumed to instantaneously and homogeneously mix in the release.

primary containment sump water (in PWRs) or suppression pool (in BWRs) at the time of release from the core. In lieu of this deterministic approach, suitably conservative mechanistic models for the transport of airborne activity in containment to the sump water may be used. Note that many of the parameters that make spray and deposition models conservative with regard to containment airborne leakage are nonconservative with regard to the buildup of sump activity.

5.2 The leakage should be taken as two times the sum of the simultaneous Conforms ESF leakage is conservatively leakage from all components in the ESF recirculation systems above assumed to begin at the start which the technical specifications, or licensee commitments to item of the LOCA event and III.D.1.1 of NUREG-0737 (Ref. A-8), would require declaring such continue for 30 days.

systems inoperable. The leakage should be assumed to start at the Page 16 of 27

Attachment B Table 2:, Conformance with RG 1.183 Appendix A (Loss-of-Coolant Accident)

RG DrsdenldQua'd Section RG Position Citi1esAnailysis Comments earliest time the recirculation flow occurs in these systems and end at The analysis assumed an the latest time the releases from these systems are terminated. ESF leakage of two gpm, Consideration should also be given to design leakage through valves which is two times the typical isolating ESF recirculation systems from tanks vented to atmosphere, industry ESF leakage rate of e.g., emergency core cooling system (ECCS) pump miniflow return to one gpm.

the refueling water storage tank.

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

5.4 Ifthe temperature of the leakage exceeds 212°F, the fraction of total Not Applicable The temperature of the iodine in the liquid that becomes airborne should be assumed equal to leakage does not exceed the fraction of the leakage that flashes to vapor. This flash fraction, FF, 2127F.

should be determined using a constant enthalpy, h, process, based on the maximum time-dependent temperature of the sump water circulating outside the containment:

FF = h -hf2 hfg Where: hf1 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 hfg is the heat of vaporization at 212 0F.

5.5 Ifthe temperature of the leakage is less than 212°F or the calculated Conforms A release fraction of 10% is flash fraction is less than 10%, the amount of iodine that becomes assumed.

airborne should be assumed to be 10% of the total iodine activity in the leaked fluid, unless a smaller amount can be justified based on the actual sump pH history and area ventilation rates.

5.6 The radioiodine that is postulated to be available for release to the Conforms Credit is taken for holdup and environment is assumed to be 97% elemental and 3% organic. dilution of ESF leakage in the Reduction in release activity by dilution or holdup within buildings, or by reactor building and for Page 17 of 27

Attachment B Table2: Conformance with RG 1.183 Appendix A (Loss-of-CoolantAccident)

S Gi RG,,osition -. Citieles'"AnalIy^,sis Commrents ESF ventilation filtration systems, may be credited where applicable, release through SGT System Filter systems used in these applications should be evaluated against filters in the same way as the guidance of Regulatory Guide 1.52 (Ref. A-5) and Generic Letter 99- containment leakage. The 02 (Ref. A-6). SGT System meets the guidance of RG 1.52 and Generic Letter 99-02.

6.1 For the purpose of this analysis, the activity available for release via Conforms MSIV leakage should be assumed to be that activity determined to be in the drywell for evaluating containment leakage (see Regulatory Position 3). No credit should be assumed for activity reduction by the steam separators or by iodine partitioning in the reactor vessel.

6.2 All the MSIVs should be assumed to leak at the maximum leak rate Conforms above which the technical specifications would require declaring the MSIVs inoperable. The leakage should be assumed to continue for the duration of the accident. Postulated leakage 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 />, if supported by site-specific analyses, to a value not less than 50% of the maximum leak rate.

6.3 Reduction of the amount of released radioactivity by deposition and Conforms Five well-mixed control plateout on steam system piping upstream of the outboard MSIVs may volumes are utilized (i.e., one be credited, but the amount of reduction in concentration allowed will be vertical and four horizontal),

evaluated on an individual case basis. Generally, the model should be. with sizing based on based on the assumption of well-mixed volumes, but other models such convective flow as slug flow may be used if justified. considerations. Gravitational settling is applied to aerosols in horizontal segments. Cline equations (basis for Bixler model) are used for organic and elemental iodine plate out. The analysis takes credit Page 18 of 27

Attachment B Table 2: Conýfo-rmance',with RG1IA83 Ap pendix A (Lossý-'of-,Coola"n't Acc'ide'nt)

RG Dre'sd e'nl6ad.

ISection RG Position Cities Analysis Comments for a time delay of 40 minutes before initiation of MSIV leakage to the environment.

6.4 In the absence of collection and treatment of releases by ESFs such as Conforms MSIV leakage is assumed to the MSIV leakage control system, or as described in paragraph 6.5 be an unprocessed, ground below, the MSIV leakage should be assumed to be released to the level release.

environment as an unprocessed, ground- level release. Holdup and dilution in the turbine building should not be assumed.

6.5 A reduction in MSIV releases that is due to holdup and deposition in Not Applicable No credit is taken for the main steam piping downstream of the MSIVs and in the main steam piping beyond the condenser, including the treatment of air ejector effluent by offgas outboard MSIVs in three lines; systems, may be credited if the components and piping systems used in no credit is taken for the the release path are capable of performing their safety function during steam piping beyond the and following a safe shutdown earthquake (SSE). The amount of inboard MSIV of the fourth reduction allowed will be evaluated on an individual case basis. line. This model assumes References A-9 and A-10 provide guidance on acceptable models. that the additional single failure concurrent with the LOCA is that the outboard MSIV of the worst leaking inboard MSIV fails to close.

No credit is taken for holdup or plate-out in the main condenser.

7.0 The radiological consequences from post-LOCA primary containment Conforms No purge is assumed.

purging as a combustible gas or pressure control measure should be analyzed. If the installed containment purging capabilities are maintained for purposes of severe a6cident management and are not credited in any design basis analysis, radiological consequences need not be evaluated. If the primary containment purging is required within Page 19 of 27

Attachment B Table 2: Conformance with RG 1.183 ApendixA (Loss-of-Coolant Accdent)

RG DregdenlQuad Seton.RG Position Ctes Analysis Comments, 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).

Table 3: .Conformance with --RG 1.1 83'AppniB(ue a-n Accdet RG -Dresd'du'l d, section. RG Position Cities. Analysi s Comments 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 on a conservative analysis that considers the most limiting case. This analysis should consider parameters such as the weight of the dropped heavy load or the weight of a dropped fuel assembly (plus any attached handling grapples), the height of the drop, and the compression, torsion, and shear stresses on the irradiated fuel rods. Damage to adjacent fuel assemblies, if applicable (e.g., events over the reactor vessel), should be considered.

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

Page 20 of 27

Attachment B Table 3. Conformance with RG 1.183Appendix B (FuelHandlng Accident)

RG D-resdenlQuad

.Sec.tion RG Position Cities Analysis .. Comments 1.3 The chemical form of radioiodine released from the fuel to the spent fuel Conforms All iodine added to pool is pool should be assumed to be 95% cesium iodide (Csl), 4.85 percent assumed to dissociate.

elemental iodine, and 0.15 percent organic iodide. The Csl released 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 Two cases were considered decontamination factors for the elemental and organic species are 500 in the analyses. The first and 1, respectively, giving an overall effective decontamination factor of case assumed a water depth 200 (i.e., 99.5% of the total iodine released from the damaged rods is of 19 feet and a retained by the water). This difference in decontamination factors for decontamination factor of elemental (99.85%) and organic iodine (0.15%) species results in the 135. The Technical iodine above the water being composed of 57% elemental and 43% Specification 3.7.8 limit for the organic species. If the depth of water is not 23 feet, the decontamination minimum spent fuel storage factor will have to be determined on a case-by-case method (Ref. B-i). pool water level is 19 feet.

The second case assumed a water depth of 23 feet and a decontamination factor of 200. The decontamination factors were determined in accordance with RG 1.183.

3 The retention of noble gases in the water in the fuel pool or reactor Conforms 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 No credit is taken for the SGT building is assumed to be released to the environment over a 2-hour System or its elevated Page 21 of 27

Attachment B Table 3:Co*nformance with RG-1-183Ap HanidlingAccident)

RG~ DresdenlQuad Section 7RGPosition .. Cities Aný.a.lysis Comments time period, release.

4.2 A reduction in the amount of radioactive material released from the fuel Not Applicable No credit is taken for filtration pool by engineered safety feature (ESF) filter systems may be taken into from the reactor building.

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.

4.3 The radioactivity release from the fuel pool should be assumed to be Not Applicable Two-hour release to the drawn into the ESF filtration system without mixing or dilution in the fuel environment is assumed.

building. If mixing can be demonstrated, credit for mixing and dilution 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 Ifthe containment is isolated during fuel handling operations, no Not Applicable Containment is not isolated.

radiological consequences need to be analyzed.

5.2 Ifthe containment is open during fuel handling operations, but designed Not Applicable Containment is not isolated.

to automatically isolate in the event of a fuel handling accident, the release duration should be based on delays in radiation detection and completion of containment isolation. Ifit 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., Conforms personnel air lock or equipment hatch is open), the radioactive material that escapes from the reactor cavity pool to the containment is released Page 22 of 27

Attachment B Table 3: Conformance with RG 1.183,Appendix B.(Fuel Handling Accident)

RG, Dresdenldu'ad

~Section RG Position Cities'An'alysis Comments" to the environment over a 2-hour time period.

5.4 A reduction in the amount of radioactive material released from the Not Applicable No credit is being taken for containment by ESF filter systems may be taken into account provided filtration of release from the that these systems meet the guidance of Regulatory Guide 1.52 and reactor building.

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 Applicable No credit is taken for dilution cavity by natural or forced convection inside the containment may be or mixing of the activity considered on a case-by-case basis. Such credit is generally limited to released from the reactor 50% of the containment free volume. This evaluation should consider cavity.

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.

• , , *; .:Table 4: Conformance with RG1.183 Appendix C (Control Rod Drop Accident)

.RG

• RGDresdenlQuad Section RGPosition~ Cities Analysis 1 Cornments 1 Assumptions acceptable to the NRC staff regarding core inventory are Conforms 100% of the noble gases and provided in Regulatory Position 3 of this guide. For the rod drop 50% of the iodines released accident, the release from the breached fuel is based on the estimate of from melted fuel. Other the number of fuel rods breached and the assumption that 10% of the releases also based on core inventory of the noble gases and iodines is in the fuel gap. The Regulatory Position 3.

release attributed to fuel melting is based on the fraction of the fuel that Page 23 of 27

Attachment B Table 4: Conformance with RG 1.183 Appenrdix C:(Control Rod Drop Accident)

~RG.Dresde'nlQuad

,Section RG Position ~C itie s~A naIyVs"is Comments reaches or exceeds the initiation temperature for fuel melting and on the assumption that 100% of the noble gases and 50% of the iodines contained in that fraction are released to the reactor coolant.

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

released activity should be the maximum coolant activity (typically 4 Therefore, coolant activity is pCi/gm DE 1-131) allowed by the technical specifications. neglected.

3.1 The activity released from the fuel from either the gap or from fuel Conforms pellets is assumed to be instantaneously mixed in the reactor coolant within the pressure vessel.

3.2 Credit should not be assumed for partitioning in the pressure vessel or Conforms for removal by the steam separators.

3.3 Of the activity released from the reactor coolant within the pressure Conforms vessel, 100% of the noble gases, 10% of the iodine, and 1% of the remaining radionuclides are assumed to reach the turbine and condensers.

3.4 Of the activity that reaches the turbine and condenser, 100% of the Conforms Release rate of 1% per day noble gases, 10% of the iodine, and 1% of the particulate radionuclides for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

are available for release to the environment. The turbine and Decay is assumed in the condensers leak to the atmosphere as a ground- level release at a rate condenser.

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 assumed to terminate. No credit should be assumed for dilution or holdup within the turbine building. Radioactive decay during holdup in the turbine and condenser may be assumed.

3.5 In lieu of the transport assumptions provided in paragraphs 3.2 through See Table 8a Plant-specific scenarios 3.4 above, a more mechanistic analysis may be used on a case-by-case considered - See Table 8a of basis. Such analyses account for the quantity of contaminated steam Attachment A.

carried from the pressure vessel to the turbine and condensers based on a review of the minimum transport time from the pressure vessel to Page 24 of 27

Attachment B 4: Conformance with RG1.183 Appendix C (Control RodDropAccident)

. .Table RG Dresden/Quad

-Section. RG Position, Cities A~nalysis Comments 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 vessel should be assumed to be 95% Csl as an aerosol, 4.85%

elemental, and 0.15% organic. The release from the turbine and condenser should be assumed to be 97% elemental and 3% organic.

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

Foot- If there are forced flow paths from the turbine or condenser, such as Conforms Forced flow paths are note 2 unisolated motor vacuum pumps or unprocessed air ejectors, the considered and the most leakage rate should be assumed to be the flow rate associated with the limiting path is determined.

most limiting of these paths. Credit for collection and processing of releases, such as by off gas or standby gas treatment, will be considered on a case-by-case basis.

Table 5: Conformance withRG 1.183Appendix D(Main Ste Ln B RG~. Dresden/Quad

-Section -RG-Position - ~CitI6ies nalsis -Comm'e'nts~

1 Assumptions acceptable to the NRC staff regarding core inventory and Not Applicable No fuel damage, release the release of radionuclides from the fuel are provided in Regulatory estimate based on coolant Position 3 of this guide. The release from the breached fuel is based on activity.

Regulatory Position 3.2 of this guide and the estimate of the number of Page 25 of 27

Attachment B Table 5:-Conformance with RG1.183 Appendix D (Main Steam Line Break)

RG~ I Dresdenl/Qu'a Section RG Position Cities Analysis Comments fuel rods breached.

2 Ifno or minimal fuel damage is postulated for the limiting event, the Conforms 4 giCi/gm is consistent with released activity should be the maximum coolant activity allowed by the spiking technical technical specification. The iodine concentration in the primary coolant is specification.

assumed to correspond to the following two cases in the nuclear steam supply system vendor's standard technical specifications.

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

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 previous.

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

3 The activity released from the fuel should be assumed to mix Conforms 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 An MSIV closure time of in the maximum time allowed by technical specifications. 5.5 seconds was assumed in the analysis. This is longer than the technical specification maximum closure time of 5 seconds.

be assumed to be that Conforms 4.2 The total mass of coolant released should amount in the steam line and connecting lines at the time of the break 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 released to the atmosphere instantaneously as a ground-level release.

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

Page 26 of 27

Attachment B Table 5: Conformance with RG 1.183Appendix D(Mare Steam LineBreak) .

RG I >~DresdeinIduad Section -RG Position 2 Cities-Anialysis Comments 4.4 The iodine species released from the main steam line should be Conforms assumed to be 95% Csl as an aerosol, 4.85% elemental, and 0.15%

organic.

Page 27 of 27

Attachment C INFORMATION SUPPORTING A FINDING OF NO SIGNIFICANT HAZARDS CONSIDERATION Summary of Proposed Changes Exelon Generation Company, LLC (EGC) is requesting a revision to the Facility Operating Licenses for Dresden Nuclear Power Station, Units 2 and 3, and Quad Cities Nuclear Power 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 Consequences Analyses Using Alternative Source Terms."

The AST analyses have also been performed with an increase in total primary containment and main steam isolation valve allowable leakage without crediting holdup and plate-out in the main steam system beyond the MSIVs; therefore, the licensing basis is being revised to reflect this position.

No Significant Hazards Consideration Determination 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.

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

The implementation of alternative source term (AST) assumptions has been evaluated in revisions to the analyses of the following limiting design basis accidents at Dresden Nuclear Power Station (DNPS) and Quad Cities Nuclear Power Station (QCNPS):

- Loss-of-Coolant Accident,

• Main Steam Line Break Accident,

• Fuel Handling Accident, and

• Control Rod Drop Accident.

Page 1 of 3

Attachment C Based upon the results of these analyses, it has been demonstrated that, with the requested changes, the dose consequences of these limiting events is within the regulatory guidance provided by the NRC for use with the AST. This guidance is presented in 10 CFR 50.67 and associated Regulatory Guide 1.183, and Standard Review Plan Section 15.0.1.

Requirements for secondary containment operability, secondary containment isolation valves, the Standby Gas Treatment (SGT) System, the Control Room Emergency Ventilation (CREV) System, and the Control Room Emergency Ventilation Air Conditioning (AC) System during movement of irradiated fuel assemblies that have decayed at least 24-hours and during core alterations are being eliminated. This is acceptable because, with the application of AST, none of these systems are credited in mitigating the consequences of a fuel handling accident after a 24-hour decay period.

The proposed change also increases the maximum allowable primary containment leakage and the maximum allowable main steam isolation valve leakage limits. This is acceptable due to the new assumptions, used in calculating control room and offsite dose following a design basis loss-of-coolant accident, related to application of AST.

The proposed changes do not affect the design or operation of the facility; rather, once the occurrence of an accident has been postulated, the new source term is an input to evaluate the consequence. The radiological consequences of the above design basis accidents have been evaluated with application of AST assumptions. The results conclude that the radiological consequences remain within applicable regulatory limits.

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

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

The application of AST does not affect the design, functional performance or operation of the facility. Similarly, it does not affect the design or operation of any structures, systems or components involved in the mitigation of any accidents, nor does it affect the design or operation of any component in the facility such that new equipment failure modes are created.

As such the proposed amendment will not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

Approval of the basis change from the original source term developed in accordance with Technical Information Document (TID) 14844 to a new AST, as described in Regulatory Guide 1.183, is requested. The results of the accident analyses revised in support of the proposed changes, and the requested Technical Specification changes, are subject to revised acceptance criteria. These analyses have been performed using conservative methodologies.

Page 2 of 3

Attachment C 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 margin of safety is considered to be that provided by meeting the applicable regulatory limits. Relaxation of these Technical Specification requirements results in an increase in dose following certain design basis accidents. However, since the doses following these design basis accidents remain within the regulatory limits, there is not a significant reduction in a margin of safety. The changes continue to ensure that the doses at the exclusion area and low population zone boundaries, as well as the control room, are within the corresponding regulatory limits.

Therefore, operation of DNPS and QCNPS in accordance with the proposed changes will not involve a significant reduction in a margin of safety.

Conclusion Based on the above, Exelon Generation Company, LLC (EGC) concludes that the proposed amendments 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.

Page 3 of 3

Attachment D INFORMATION SUPPORTING AN ENVIRONMENTAL ASSESSMENT Exelon Generation Company, LLC (EGC) has evaluated the proposed changes against the criteria for identification of licensing and regulatory actions requiring environmental assessment in accordance with 10 CFR 51.21, "Criteria for and identification of licensing and regulatory actions requiring environmental assessments." EGC 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 Attachment C, 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 effluent that may be released offsite.

The following table demonstrates that EGC 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.

Dose'Results (rem)

EAB Doses and Limit LPZ Doses and Limit DNPS QCNPS DNPS: QCNPS Accident Dose Dose Limit Dose Dose Limit Loss of Coolant 1.3 6.5 25 0.4 1.1 25 Accident Main Steam 4.84E-2' 9.50E-2'2 25' 6.05E-3'2 9.55E-3'2 251 2.42E-32 4.75E-3 2.52 3.02E-4 4.77E-4 2.52 Line Break Control Rod <3 <3 6.3 <3 <3 6.3 Drop Accident Fuel Handling 0.713 3.03 6.3 0.072 0.181 6.3 Accident Page 1 of 2

Attachment D Notes: 1. Based on a pre-accident spike concentration of 4.0 ACi/gm dose equivalent 1-131.

2. Based on a maximum equilibrium concentration of 0.2 pCi/gm dose equivalent 1-131.

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 modifications to the plant or changes in its operation which could alter the type or amounts of effluents that may be released offsite.

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

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

CR Dose Results (rem),

DNPS QCNPS'

'Accident Dose Dose Limit Loss of Coolant Accident 4.53 3.9 5.0 Main Steam Line Break 2.45E-11 2.45E-112 5.0 2 1.23E-2 1.23E-2 Control Rod Drop Accident < 2.5 < 2.5 5.0 Fuel Handling Accident 2.00 1.69 5.0 Notes: 1. Based on a pre-accident spike concentration of 4.0 [tCi/gm dose equivalent 1-131.

2. Based on a maximum equilibrium concentration of 0.2 [tCi/gm dose equivalent 1-131.

The alternative source term does not affect the design or operation of the facility; rather, once the occurrence of an accident has been postulated, the alternative source term is an input to evaluate the consequence. The implementation of the alternative source term has been evaluated in revisions to the analyses of the limiting design basis accidents at Quad Cities Nuclear Power Station, Units 1 and 2, and Dresden Nuclear Power Station, Units 2 and 3. 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 requested 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.

Page 2 of 2

Attachment E-1 MARKED-UP TECHNICAL SPECIFICATIONS AND BASES PAGES FOR PROPOSED CHANGES DRESDEN NUCLEAR POWER STATION, UNITS 2 AND 3

Definitions 1.1 1.1 Definitions (continued)

CHANNEL FUNCTIONAL TEST A CHANNEL FUNCTIONAL TEST shall be the injection of a simulated or actual signal into the channel as close to the sensor as practicable to verify OPERABILITY of all devices in the channel required for channel OPERABILITY. The CHANNEL FUNCTIONAL TEST may be performed by means of any series of sequential, overlapping, or total channel steps.

CORE ALTERATION 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 movable 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 The COLR is the unit specific document that REPORT (COLR) provides cycle specific parameter limits for the current reload cycle. These cycle specific limits shall be determined for each reload cycle in accordance with Specification 5.6.5. Plant operation within these limits is addressed in individual Specifications.

DOSE EQUIVALENT 1-131 DOSE EQUIVALENT 1-131 shall be that concentration of 1-131 (microcuries/gram) that alone would produce the same th. dose as the quantity and isotopic mixture of 1-131, 1-132, 1-133, 1-134, and 1-135 actually present. The( dose (continued)

Dresden 2 and 3 1.1-2 Amendment No. 185/180

Definitions 1.1 1.1 Definitions DOSE EQUIVALENT 1-131 conversion factors used for this calculation shall (continued) be those listed in Table III of TID-14844.

AEC, 1962, "Calculation of Distance Factors for Power and Test Reactor Sites;."Table E-7 of Regulatory Guide 1.1 , ev. 1, NRC, 1977; or ICRP 30, Supplement to Part 1. pages 192-212, Table titled, "Committed Dose Equivalent in Target Organs or Tissues per Intake of Unit Activity."

I LEAKAGE

• LEAKAGE shall be:

a. Identified LEAKAGE

1. LEAKAGE into the drywell, such as that from pump seals or valve packing, that is captured and conducted to a sump or 11l. . *

'""_collecting tank; or

2. LEAKAGE into the drywell atmosphere from

/I(/te 1 7R'Y sources that are both specifically located 4,,z*- cf2O/ca/dt7 a Oi,/

• and known either not to interfere with the operation of leakage detection systems or 2AS'*" 4,04.ers/O-, '4-/ -'S not to be pressure boundary LEAKAGE; al'0 -- t*4/a/i e'

a,24 Z/ 5 d/ FO'i", /98 b.

Unidentified LEAKAGE All LEAKAGE into the drywell that is not identified LEAKAGE;

c. Total LEAKAGE Sum of the identified and unidentified LEAKAGE; and

d. Pressure Boundary LEAKAGE LEAKAGE through a nonisolable fault in a Reactor Coolant System (RCS) component body, pipe wall, or vessel wall.

(continued)

Dresden 2 and 3 1.1-3 Amendment No. 191/185

SLC System 3.1.7 3.1 REACTIVITY CONTROL SYSTEMS 3.1.7 Standby Liquid Control (SLC) System LCO 3.1.7 Two SLC subsystems shall be OPERABLE.

APPLICABILITY: MODES Ian ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One SLC subsystem A.1 Restore SLC subsystem 7 days inoperable, to OPERABLE status.

B. Two SLC subsystems B.1 Restore one SLC 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> inoperable, subsystem to OPERABLE status.

C. Required Action and C.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time not met.

C.?

~einMODE,'1 Dresden 2 and 3 3.1.7-1 Amendment No. 185/180

Primary Containment Isolation Instrumentation 3.3.6.1 Table 3.3.6.1-1 (page 3 of 3)

Primary Containment Isolation Instrumentation APPLICABLE CONDITIONS MODES OR REOUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER TRIP REOUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEM ACTION C.1 REGUIREMENTS VALUE

5. Reactor Water Cleanup System Isolation

a. SLC System Initiation 1 1, H SR 3.3.6.1.7 NA

b. Reactor Vessel Water 1.2.3 2 F SR 3.3.6.1.1 > 2.65 inches Level - Low SR 3.3.6.1.2 SR 3.3.6.1.3 SR 3.3.6.1.6 SR 3.3.6.1.7

6. Shutdown Cooling System Isolation

a. Recirculation Line 1.2.3 2 F SR 3.3.6.1.2 < 346 0 F Water SR 3.3.6.1.6 Temperature-High SR 3.3.6.1.7

b. Reactor Vessel Water 3.4.5 2 (b) I SR 3.3.6.1.1 > 2.65 inches Level - Low SR 3.3.6.1.2 SR 3.3.6.1.3 SR 3.3.6.1.6 SR 3.3.6.1.7 (b) In MODES 4 and 5. provided Shutdown Cooling System integrity is maintained, only one channel per trip system with an isolation signal available to one shutdown cooling pump suction isolation valve is required.

Dresden 2 and 3 3.3.6.1-7 Amendment No. 190/184

Secondary Containment Isolation Instrumentation 3.3.6.2 Table 3.3.6.2-1 (page I of 1)

Secondary Containment Isolation Instrumentation APPLICABLE MODES OR REOUIRED OTHER CHANNELS SPECIFIED PER SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS TRIP SYSTEM REDUIREMENTS VALUE

1. Reactor Vessel Water 1.2.3. 2 SR 3.3.6.2.1 > 2.65 inches Level - Low (a) SR 3.3.6.2.2 SR 3.3.6.2.3 SR 3.3.6.2.5 SR 3.3.6.2.6

2. Orywel1 Pressure-High 1.2.3 2 SR 3.3.6.2.2 < 1.94 psig SR 3.3.6.2.4 SR 3.3.6.2.6

3. Reactor Building Exhaust 1.2.3. 2 SR 3.3.6.2.1 < 14.9 mR/hr Radiation - High (a).(b) SR 3.3.6.2.2 SR 3.3.6.2.4 SR 3.3.6.2.6

4. Refueling Floor 1.2.3. 2 SR 3.3.6.2.1 < 100 mR/hr Radiation-High (a).(b) SR 3.3.6.2.2 SR 3.3.6.2.4 SR 3.3.6.2.6 (a) During operations with a potential for draining the reactor vessel.

(b) During ERAT Sand rng movement of irradiated fuel assemblies in secondary containment.

reeel'C Dresden 2 and 3 3.3.6.2-4 Amendment No. 190/184

CREV System Instrumentation 3.3.7.1 3.3 INSTRUMENTATION 3.3.7.1 Control Room Emergency Ventilation (CREV) System Instrumentation LCO 3.3.7.1 Two channels of the Reactor Building Ventilation System-High High Radiation Alarm Function shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3, During movement of irradiated fuel assemblies in the secondary containment.

Gr ij LO R RAT I D-uring operations with a potential for draining the reactor vessel (OPDRVs).

ACTIONS

---

-----NOTE OTE.................... ------------

Separate Condition entry is allowed for each channel.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more channels A.1 Declare CREV System 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> from inoperable. inoperable, discovery of loss of CREV System Instrumentation alarm capability in both trip systems AND A.2 Restore channel to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> OPERABLE status.

(continued)

Dresden 2 and 3 3.3.7.1-1 Amendment No. 185/180

PCIVs 3.6.1.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.1.3.10 Veri y the combined eakage rate for/ all In accordance MS leakage paths is < 46 scfh wh with the t sted at > 25 p g. Primary Can+  ; nm~+

Leakage Rate Testing Program I-l a11 pallj* /Is e1vt I*

/~/se-It,*

Dresden 2 and 3 3.6.1.3-9 Amendment No. 185/180

Secondary Containment 3.6.4.1 3.6 CONTAINMENT SYSTEMS 3.6.4.1 Secondary Containment LCO 3.6.4.1 The secondary containment shall be OPERABLE.

and 3.2 APPLICABILITY: MODES 1, 2, During movement of irradiated fuel assemblies in the econdary containment, Dun i CORE AT S During operations with a potential for draining the reactor vessel (OPDRVs).

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Secondary containment A.1 Restore secondary 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> inoperable in MODE 1, containment to

2. or 3. OPERABLE status.

B. Required Action and B.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time of Condition A AND not met.

B.2 Be in MODE 4. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> C. Secondary containment C.1 -------- NOTE-------

inoperable during LCO 3.0.3 is not movement _ i rradiated applicable.

fuel assemblies in th -------------------

secondary contai nmenV,

-d~wing COX* Suspend movement of Immediately 4LTERAT NS, or during irradiated fuel OPDRVs. -assemblies in the C secondary containment.

AND -&eel (continued)

Dresden 2 and 3 3.6.4.1-1 Amendment No. 185/180

Secondary Containment 3.6.4.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. (continued) Su s d CORE

ýý A ERATIONS C Initiate action to Immediately suspend OPDRVs.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.4.1.1 Verify secondary containment vacuum is 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />

> 0.25 inch of vacuum water gauge.

SR 3.6.4.1.2 Verify one secondary containment access 31 days door in each access opening is closed.

SR 3.6.4.1.3 Verify the secondary containment can be 24 months on a maintained > 0.25 inch of vacuum water STAGGERED TEST gauge for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> using one SGT subsystem BASIS for each at a flow rate < 4000 cfm. SGT subsystem SR 3.6.4.1.4 Verify all secondary containment 24 months equipment hatches are closed and sealed.

Dresden 2 and 3 3.6.4.1-2 Amendment No. 185/180

SCIVs 3.6.4.2 3.6 CONTAINMENT SYSTEMS 3.6.4.2 Secondary Containment Isolation Valves (SCIVs)

LCO 3.6.4.2 Each SCIV shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3,.

During movement of irradiated fuel assemblies in the secondary containment, uri nCORE ERATT ,S During operations w-i"a potential for draining the reactor vessel (OPDRVs).

ACTIONS

---

. NOTES -----------------------------------

1. Penetration flow paths may be unisolated intermittently under administrative controls.

2. Separate Condition entry is allowed for each penetration flow path.

3. Enter applicable Conditions and Required Actions for systems made inoperable by SCIVs.

CONDITION REOUIRED ACTION COMPLETION TIME A. One or more A.1 Isolate the affected 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> penetration flow paths penetration flow path with one SCIV by use of at least inoperable, one closed and de-activated automatic valve, closed manual valve, or blind flange.

AND (continued)

Dresden 2 and 3 3.6.4.2-1 Amendment No. 185/180

SCIVs 3.6.4.2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. Required Action and C.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time of Condition A AND or B not met in MODE 1, 2, or 3. C.2 Be in MODE 4. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> D. Required Action and D.1 -------- NOTE --------

associated Completion LCO 3.0.3 is not Time of Condition A applicable.

or B not met during --------------

mo v e me n t o f i r r a d i a t e d S/m v e oI ei l fuel assemblies in the movement of Immediately secondary containmentl ' -Tmirradiated fuel

,dur* g CORE/ assemblies in the A RATIONXor during secondary OPDRVs. containment.

.2 Suspend C ~a ty AND D Initiate action to Immediately suspend OPDRVs.

Dresden 2 and 3 3.6.4.2-3 Amendment No. 185/180

SGT System 3.6.4.3 3.6 CONTAINMENT SYSTEMS 3.6.4.3 Standby Gas Treatment (SGT) System LCO 3.6.4.3 Two SGT subsystems shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3, During movement of irradiated fuel assemblies in the secondary containment, iDuri.< COREZ TERAT4VS.

During operations with a potential for draining the reactor vessel (OPDRVs).

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One SGT subsystem A.1 Restore SGT 7 days inoperable, subsystem to OPERABLE status.

B. Required Action and B.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time of Condition A AND not met in MODE 1, 2.

or 3. B.2 Be in MODE 4. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> C. Required Action and ------------ NOTE -----------

associated Completion LCO 3.0.3 is not applicable.

Time of Condition A ---------------------------

not met during movement oirradiated C. Place OPERABLE SGT Immediately fuel assemblies in the subsystem in

/ secondary containment/,) operation.

durAg CORE AkTERATIO.S , or during OR OP s. (oDRi r (continued)

Dresden 2 and 3 3.6.4.3-1 Amendment No. 185/180

SGT 3.6.4.3 System ACTIONS REQUIRED ACTION moev2`erne Su sCpe n-dt Immediately C. (continued) ,irradiated fuel assemblies in secondary containment.

Initiate action to Immediately suspend OPDRVs.

D.1 Restore one SGT 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> D. Two SGT subsystems subsystem to inoperable in MODE 1, OPERABLE status.

2, or 3.

E.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> E. Required Action and associated Completion Time of Condition D AND not met.

E.2 Be in MODE 4. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> F.1 -----NOT E ------- ---

F. Two SGT subsystems LCO 3.0.3 is not inoperable during applicable.

movement of4irradiated I - - -- - - - - ---

fIuel assemblies in the secondary containmen ,drjr Suspend movement of Immediately irradiated fuel assemblies in secondary containment.

AND (continued) 3.6.4.3-2 Amendment No. 185/180 Dresden 2 and 3

SGT System 3.6.4.3 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME F. (continued) FF. Sus/end CORE a AND A ERATIONS.

F.F Initiate action to Immediately suspend OPDRVs.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.4.3.1 Operate each SGT subsystem for > 10 31 days continuous hours with heaters operating.

SR 3.6.4.3.2 Perform required SGT filter testing in In accordance accordance with the Ventilation Filter with the VFTP Testing Program (VFTP).

SR 3.6.4.3.3 Verify each SGT subsystem actuates on an 24 months actual or simulated initiation signal.

Dresden 2 and 3 3.6.4.3-3 Amendment No. 185/180

CREV System 3.7.4 3.7 PLANT SYSTEMS 3.7.4 Control Room Emergency Ventilation (CREV) System LCO 3.7.4 The CREV System shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3, During movement of irradiated fuel assemblies in the QDu~ig CORFI.4LTERATI*,5)

During operations wi a potential for draining the reactor vessel (OPDRVs).

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. CREV System inoperable A.1 Restore CREV System 7 days in MODE 1, 2, or 3. to OPERABLE status.

B. Required Action and B.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time of Condition A AND not met in MODE 1. 2, or 3. B.2 Be in MODE 4. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> C. CREV System inoperable ------------ NOTE ------------

during movement of LCO 3.0.3 is not applicable.

• irradiated fuel ----------------------------

assemblies in the Ftainment, C.1 Suspend movement of Immediately (drigOR-jj J Airradiated fuel A RATI S, or during assemblies in the OPDRVs. -secondary containment.

ree~e,7'/p/AND (continued)

Dresden 2 and 3 3.7.4-1 Amendment No. 185/180

CREV System 3.7.4 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. (continued) C.2 Susp dCORE at AL RATIONS.

AND C Initiate action to Immediately suspend OPDRVs.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.4.1 Operate the CREV System for > 10 continuous 31 days hours with the heaters operating.

SR 3.7.4.2 Perform required CREV filter testing in In accordance accordance with the Ventilation Filter with the VFTP Testing Program (VFTP).

SR 3.7.4.3 Verify the CREV System actuates on a manual 24 months initiation signal.

SR 3.7.4.4 Verify the CREV System can maintain a 24 months positive pressure of > 0.125 inches water gauge relative to the adjacent areas during the isolation/pressurization mode of operation at a flow rate of < 2000 scfm.

Dresden 2 and 3 3.7.4-2 Amendment No. 185/180

Control Room Emergency Ventilation AC System 3.7.5 3.7 PLANT SYSTEMS 3.7.5 Control Room Emergency Ventilation Air Conditioning (AC) System LCO 3.7.5 The Control Room Emergency Ventilation AC System shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3,/

During movement of irradiated fuel assemblies in the secondary containment,

(:urg CORELTERAT_ S, During operations with a potential for draining the reactor vessel (OPDRVs).

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Control Room Emergency A.1 Restore Control Room 30 days Ventilation AC System Emergency Ventilation inoperable in MODE 1, AC System to OPERABLE 2, or 3. status.

B. Required Action and B.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time of Condition A AND not met in MODE 1, 2, or 3. B.2 Be in MODE 4. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> (continued)

Dresden 2 and 3 3.7.5-1 Amendment No. 185/180

Control Room Emergency Ventilation AC System 3.7.5 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. Control Room Emergency ------------ NOTE----------

Ventilation AC System LCO 3.0.3 is not applicable.

inoperable during - -

movement oLairradiated fu assemblies in the C.1 uspend move of Immediately secondary containment, irradiated fuel (dung CORE assemblies in the KTERATI; 9,o6-r during secondary OPDRVs. containment.

AND AL RATIONS.

AND SInitiate action to Immediately suspend OPDRVs.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.5.1 Verify the Control Room Emergency 24 months Ventilation AC System has the capability to remove the assumed heat load.

Dresden 2 and 3 3.7.5-2 Amendment No. 185/180

AC Sources-Operating 3.8.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.1.20 ------------------- NOTE -------------------

All DG starts may be preceded by an engine prelube period.

Verify, when started simultaneously from 10 years standby condition, each DG achieves, in

< 13 seconds, voltage > 3952 V and frequency > 58.8 Hz.

SR 3.8.1.21 ------------------- NOTE -------------------

When the opposite unit is in MODE 4 or 5, or movin irradiated fuel assemblies in secon ary containment, the following opposite unit SRs are not required to be performed: SR 3.8.1.3, SR 3.8.1.10 through SR 3.8.1.12, and SR 3.8.1.14 through SR 3.8.1.17.

For required opposite unit AC electrical In accordance power sources, the SRs of the opposite with applicable unit's Specification 3.8.1, except SRs SR 3.8.1.9, SR 3.8.1.13, SR 3.8.1.18, SR 3.8.1.19, and SR 3.8.1.20, are applicable.

Dresden 2 and 3 3.8.1-15 Amendment No. 185/180

AC Sources -Shutdown 3.8.2 3.8 ELECTRICAL POWER SYSTEMS 3.8.2 AC Sources-Shutdown LCO 3.8.2 The following AC electrical power sources shall be OPERABLE:

a. One qualified circuit between the offsite transmission network and the onsite Class IE AC electrical power distribution subsystem(s) required by LCO 3.8.8.

"Distribution Systems -Shutdown"; and

b. One diesel generator (DG) capable of supplying one division of the onsite Class 1E AC electrical power distribution subsystem(s) required by LCO 3.8.8.

APPLICABILITY: MODES 4 and 5, During movement ofirradiated fuel assemblies in the secondary containment.

Dresden 2 and 3 3.8.2-1 Amendment No. 185/180

AC Sources- Shutdown 3.8.2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.2.2 Suspend movement of Immediately irradiated in "Yassemblies fuelthe secondary containment.

A.2.3 Initiate action to Immediately suspend operations with a potential for draining the reactor vessel (OPDRVs).

AND A.2.4 Initiate action to Immediately restore required offsite power circuit to OPERABLE status.

(continued)

Dresden 2 and 3 3.8.2-3 Amendment No. 185/180

AC Sources -Shutdown 3.8.2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. One required DG B.1 Suspend CORE Immediately inoperable. ALTERATIONS.

AND B.2 uspend movement of Immediately C-irradiated fuel assemblies in secondary containment.

AND B.3 Initiate action to Immediately suspend OPDRVs.

AND B.4 Initiate action to Immediately restore required DG to OPERABLE status.

Dresden 2 and 3 3.8.2-4 Amendment No. 185/180

DC Sources -Shutdown 3.8.5 3.8 ELECTRICAL POWER SYSTEMS 3.8.5 DC Sources-Shutdown LCO 3.8.5 One 250 VDC and one 125 VDC electrical power subsystem shall be OPERABLE to support the 250 VDC and one 125 VDC Class 1E electrical power distribution subsystems required by LCO 3.8.8. "Distribution Systems-Shutdown."

APPLICABILITY: MODES 4 and 5, During movement o iirradiated fuel assemblies in the secondary containment.

ACTIONS

---

NOT E . . .OTE..

LCO 3.0.3 is not applicable.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required A.1 Declare affected Immediately DC electrical power required feature(s) subsystems inoperable. inoperable.

OR A.2.1 Suspend CORE Immediately ALTERATIONS.

AND A.2.2 Suspend movement of Immediately irradiated fuel assemblies in the secondary containet ANcDoniud (continued)

Dresden 2 and 3 3.8.5-1 Amendment No. 185/180

Distribution Systems -Shutdown 3.8.8 3.8 ELECTRICAL POWER SYSTEMS 3.8.8 Distribution Systems-Shutdown LCO 3.8.8 The necessary portions of the AC. DC, and the opposite unit's Division 2 electrical power distribution subsystems shall be OPERABLE to support equipment required to be OPERABLE.

APPLICABILITY: MODES 4 and 5, During movement of irradiated fuel assemblies in the secondary containment.

ACTIONS NOT E ------------------------------------

LCO 3.0.3 is not applicable.

CONDITION REHUIRED ACTION COMPLETION TIME A. One or more required A.1 Declare associated Immediately AC or DC electrical supported required power distribution feature(s) subsystems inoperable, inoperable.

OR A.2.1 Suspend CORE Immediately ALTERATIONS.

AND (continued) 3.8.8-1 Amendment No. 185/180 Dresden 2 and 3

Distribution Systems -Shutdown 3.8.8 ACTIONS CONDITION REQUIRED ACTION FCOMPLETION TIME A. (continued) A.2.2 Suspend movement of Immediately irradiated assemblies fuel in the secondary A.2.3 Initiate action to Immediately suspend operations with a potential for draining the reactor vessel.

AND A.2.4 Initiate actions to Immediately restore required AC and DC electrical power distribution subsystems to OPERABLE status.

AND A.2.5 Declare associated Immediately required shutdown cooling subsystem(s) inoperable and not in operation.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.8.1 Verify correct breaker alignments and 7 days voltage to required AC and DC electrical power distribution subsystems.

Dresden 2 and 3 3.8.8-2 Amendment No. 185/180

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.7 Ventilation Filter Testinq Proqram (VFTP) (continued)

ESF Ventilation System Penetration Flowrate Standby Gas < 1.0% > 3600 cfm and Treatment (SGT) < 4400 cfm System Control Room < 0.05% > 1800 scfm and Emergency < 2200 scfm Ventilation (CREV)

System

c. Demonstrate for each of the ESF systems that a laboratory test of a sample of the charcoal adsorber. when obtained as described in Regulatory Guide 1.52, Revision 2, shows the methyl'iodide penetration less than the value specified below when tested in accordance with ASTM D3803-1989 at a temperature of 30 0 C and relative humidity (RH) specified bel ow:

ESF Ventilation System Penetration RH Standby Gas Treatment - 70%

(SGT) System Control Room  % 70%

Emergency Ventilation (CREV) System

d. Demonstrate for each of the ESF systems that the pressure drop across the combined HEPA filters and the charcoal adsorbers is less than the value specified below when tested at the system flowrate specified as follows:

ESF Ventilation System Delta P Flowrate Standby Gas < 6 inches > 3600 cfm and Treatment (SGT) water guage < 4400 cfm System Control Room < 6 inches > 1800 scfm and Emergency water guage < 2200 scfm Ventilation (CREV) System (continued)

Dresden 2 and 3 5.5-7 Amendment No. 185/180

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.12 Primary Containment Leakage Rate Testing Program (continued)

C. The maximuma lowable primary containment leakage rate, L,,

at P, m of primary containment air weight per day.

d. Leakage rate acceptance criteria are:

1. Primary containment overall leakage rate acceptance criterion is < 1.0 La. During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are < 0.60 La for the combined Type B and Type C tests, and < 0.75 La for Type A tests.

2. Air lock testing acceptance criteria is the overall air lock leakage rate is < 0.05 L, when tested at > P,.

e. The provisions of SR 3.0.3 are applicable to the Primary Containment Leakage Rate Testing Program.

Dresden 2 and 3 5.5-12 Amendment No. 185/180

Dresden Bases Inserts Insert A The SLC System is also used to maintain suppression pool pH at or above 7 following a loss of coolant accident (LOCA) involving significant fission product releases. Maintaining suppression pool pH levels at or above 7 following an accident ensures that iodine will be retained in the suppression pool water (Ref. 3).

Insert B Following a LOCA, offsite doses from the accident will remain within 10 CFR 50.67, "Accident Source Term," limits (Ref. 4) provided sufficient iodine activity is retained in the suppression pool. Credit for iodine deposition in the suppression pool is allowed (Ref. 3) as long as suppression pool pH is maintained at or above 7. Alternative Source Term analyses credit the use of the SLC System for maintaining the pH of the suppression pool at or above 7.

Insert B1 In MODES 1, 2, and 3, the SLC System must be OPERABLE to ensure that offsite doses remain within 10 CFR 50.67 (Ref. 4) limits following a LOCA involving significant fission product releases. The SLC System is designed to maintain suppression pool pH at or above 7 following a LOCA to ensure that iodine will be retained in the suppression pool water (Ref. 3).

Insert C Due to radioactive decay, these Functions are only required to isolate secondary containment during fuel handling accidents involving handling recently irradiated fuel (i.e., 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 D Also due to radioactive decay, these Functions are only required to be OPERABLE during fuel handling accidents involving handling recently irradiated fuel (i.e., 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 E The combined leakage rate limit for all MSIV leakage paths is < 144 scfh when tested at

_>25 psig. Additionally, the leakage rate limit through each MSIV leakage path is < 57 scfh when tested at > 25 psig. These values correspond to a combined leakage rate of 250 scfh and an individual MSIV leakage rate of 100 scfh, when tested at 48 psig.

Insert F Due to radioactive decay, secondary containment is only required to be OPERABLE during fuel handling involving handling recently irradiated fuel (i.e., 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 />).

Dresden Bases Inserts Insert G Due to radioactive decay, SCIVs are only required to be OPERABLE during fuel handling involving handling recently irradiated fuel (i.e., 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 H Due to radioactive decay, the SGT System is only required to be OPERABLE during fuel handling involving handling recently irradiated fuel (i.e., 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 I Due to radioactive decay, the CREV System is only required to be OPERABLE during fuel handling involving handling recently irradiated fuel (i.e., 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 J Due to radioactive decay, the Control Room Emergency Ventilation AC System is only required to be OPERABLE during fuel handling involving handling recently irradiated fuel (i.e., 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 K (calculated control room operator dose and doses at the exclusion area and low population zone boundaries) are below the 10 CFR 50.67 (Ref. 3) exposure guidelines.

Insert L involving handling recently irradiated fuel. Due to radioactive decay, AC electrical power is only required to mitigate fuel handling accidents involving handling recently irradiated fuel (i.e., 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 M involving handling recently irradiated fuel Insert N involving handling recently irradiated fuel (i.e., 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 0 involving handling recently irradiated fuel. Due to radioactive decay, DC electrical power is only required to mitigate fuel handling accidents involving handling recently irradiated fuel (i.e., 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 />)

Dresden Bases Inserts Insert P involving handling recently irradiated fuel Insert Q Due to radioactive decay, DC electrical power is only required to mitigate fuel handling accidents involving handling recently irradiated fuel (i.e., 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 R involving handling recently irradiated fuel Insert S Due to radioactive decay, AC and DC electrical power is only required to mitigate fuel handling accidents involving handling recently irradiated fuel (i.e., 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 T involving handling recently irradiated fuel Insert U involving handling recently irradiated fuel (i.e., 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 />

Reactor Core SLs B 2.1.1 BASES APPLICABLE 2.1.1.3 Reactor Vessel Water Level (continued)

SAFETY ANALYSES reduction in cooling capability could lead to elevated cladding temperatures and clad perforation in the event that the water level becomes < 2/3 of the core height. The reactor vessel water level SL has been established at the top of the active irradiated fuel to provide a point that can be monitored and to also provide adequate margin for effective action.

SAFETY LIMITS The reactor core SLs are established to protect the integrity of the fuel clad barrier to prevent the release of radioactive materials to the environs. SL 2.1.1.1 and SL 2.1.1.2 ensure that the core operates within the fuel design criteria. SL 2.1.1.3 ensures that the reactor vessel water level is greater than the top of the active irradiated fuel in order to prevent elevated clad temperatures and resultant clad perforations.

APPLICABILITY SLs 2.1.1.1, 2.1.1.2. and 2.1.1.3 are applicable in all MODES.

SAFETY LIMIT 2.2 VIOLATIONS Exceeding an SL may cause fuel damage and create a potential for radioactive releases in excess of 10 CFR*- 7orJ

,, *SiX Cri 7i Ti," limits (Ref. 6). Therefore, itis required I to insert a1l insertable control rods and restore compliance with the SLs within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. The 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Completion Time ensures that the operators take prompt remedial action and also ensures that the probability of an accident occurring during this period is minimal.

(continued)

Dresden 2 and 3 B 2.1.1-5 Revision 3

Reactor Core SLs B 2.1.1 BASES (continued)

REFERENCES 1. UFSAR, Section 3.1.2.2.1.

2. ANF-524(P)(A) and Supplements 1 and 2, Advanced Nuclear Fuels Corporation Critical Power Methodology for Boiling Water Reactors, (as specified in Technical Specification 5.6.5).

3. ANF-1125(P)(A) and Supplements 1 and 2, ANFB Critical Power Correlation, Advanced Nuclear Fuels Corporation, (as specified in Technical Specification 5.6.5).

4. NEDE-24011-P-A, General Electric Standard Application for Reactor Fuel (GESTAR) (as specified in Technical Specification 5.6.5)

5. ANF-1125(P)(A), Supplement 1, Appendix E, ANFB Critical Power Correlation Determination of ATRIUM-9B Additive Constant Uncertainties, Siemens Power Corporation, (as specified in Technical Specification 5.6.5))

6. 10 CFR Dresden 2 and 3 R 2-1-1-6 Revi si on 3

RCS Pressure SL B 2.1.2 B 2.0 SAFETY LIMITS (SLs)

B 2.1.2 Reactor Coolant System (RCS) Pressure SL BASES BACKGROUND The SL on reactor steam dome pressure protects the RCS against overpressurization. In the event of fuel cladding failure, fission products are,-released into the reactor coolant. The RCS then serves as the primary barrier in preventing the release of fission products into the atmosphere. Establishing an upper limit on reactor steam dome pressure ensures continued RCS integrity. According to UFSAR Sections 3.1.2.2.5, and 3.1.2.2.6 (Ref. 1). the reactor coolant pressure boundary (RCPB) shall be designed with sufficient margin to ensure that the design conditions are not exceeded during normal operation and anticipated operational occurrences (AOOs).

During normal operation and AOs, RCS pressure is limited from exceeding the design pressure by more than 10%, in accordance with Section III of the ASME Code,(Ref. 2). To ensure system integrity, all RCS components are hydrostatically tested at 125% of design pressure, in accordance with ASME Code requirements, prior to initial operation when there is no fuel in the core. Following inception of unit operation, RCS components shall be pressure tested in accordance with the requirements of ASME Code,Section XI (Ref. 3).

Overpressurization of the RCS could result in a breach of the RCPB, reducing the number of protective barriers designed to prevent radioactive releases from exceedin the limits specified in 10 CFR 100, ' eactor te Crite a" (Ref. 4). If this occurred in conjunction wit a ue cladding failure, fission products could enter the containment atmosphere.

(continued)

S67/,2~/n Suc ~-n Dresden 2 and 3 B 2.1.2-1 Revision 0

RCS Pressure SL B 2.1.2 BASES (continued)

APPLICABLE The RCS safety/relief valves and the Reactor Protection SAFETY ANALYSES System Reactor Vessel Steam Dome Pressure-High Function have settings established to ensure that the RCS pressure SL will not be exceeded.

The RCS pressure SL has been selected such that it is at a 11: pressure below which it can be shown that the inteqni4ty of the system is not endangered. The reactor pressure vessel is designed to Section III of the ASME, Boiler and Pressure Vessel Code, 1963 Edition, including Addenda through the summer of 1964 and Code Case Interpretations applicable on February 8, 1965 (Ref. 5), which permits a maximum pressure transient of 110%, 1345 psig, of design pressure 1250 psig.

The SL of 1345 psig, as measured in the reactor steam dome, is equivalent to 1375 psig at the lowest elevation of the RCS. The RCS is designed to the USAS Power Piping Code, Section B31.1, 1967 Edition (Ref. 6), and ASME, Boiler and Pressure Vessel Code,Section I. 1965 Edition, including Addenda winter 1966 (Ref. 7) for the reactor recirculation piping, which permits a maximum pressure transient of 120%

of a design pressure of 1175 psig for suction piping and 1325 psig for discharge piping. The RCS pressure SL is selected to be the lowest transient overpressure allowed by the applicable codes.

SAFETY LIMITS The maximum transient pressure allowable in the RCS pressure vessel under the ASME Code,Section III, is 110% of design pressure. The maximum transient pressure allowable in the RCS piping, valves, and fittings is 120% of the design pressure of 1175 psig for suction piping. The most limiting of these allowances is the 110% of the RCS pressure vessel design pressure; therefore, the SL on maximum allowable RCS pressure is established at 1345 psig as measured at the reactor steam dome.

APPLICABILITY SL 2.1.2 applies in all MODES.

SAFETY LIMIT 2.2 6- 5%qeelden -Source 7re-m,r, VIOLATIONS Exceeding the RCS pressureSL may cause RCS failure create a otential for iv andf eleases in excess of 10 CFR , "R ctor S e Criter a,")limits (Ref. 4).

(continued)

Dresden 2 and 3 B 2.1.2-2 Revision 0

RCS Pressure SL B 2.1.2 BASES SAFETY LIMIT 2.2 (continued)

VIOLATIONS Therefore, it is required to insert all insertable control rods and restore compliance with the SL within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. The 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Completion Time ensures that the operators take prompt remedial action and also assures that the probability of an accidentoccurring during this period is minimal.

REFERENCES 1. UFSAR Sections 3.1.2.2.5. and 3.1.2.2.6.

2. ASME, Boiler and Pressure Vessel Code, Section III, Article NB-7000.

3. ASME, Boiler and Pressure Vessel Code, Section XI, Article IWB-5000.

4. 10 CFR*or67

5. ASME, Boiler and Pressure Vessel Code,Section III, 1963 Edition, Addenda summer of 1964 and Code Case Interpretations applicable on February 8, 1965.

6. ASME, USAS, Power Piping Code, Section B31.1, 1967 Edition.

7. ASME, Boiler and Pressure Vessel Code, Section I, 1965 Edition, Addenda winter 1966.

Dresden 2 and 3 B 2.1.2-3 Revision 0

Rod Pattern Control B 3.1.6 BASES (continued)

REFERENCES 1. UFSAR, Section 15.4.10.

2. XN-NF-80-19(P)(A), Volume 1, Supplement 2, Section 7.1 Exxon Nuclear Methodology for Boiling Water Reactors Neutronics Methods for Design and Analysis, (as specified in Technical Specification 5.6.5).

3. NEDE-24011-P-A, "GE Standard Application for Reactor Fuel," (as specified in Technical Specification 5.6.5).

4. Letter from T.A. Pickens (BWROG) to G.C. Lainas (NRC),

"Amendment 17 to General Electric Licensing Topical Report NEDE-24011-P-A." BWROG-8644. August 15, 1986.

5. NFSR-0091, Benchmark of CASMO/MICROBURN BWR Nuclear Design Methods, Commonwealth Edison Topical Report, (as specified in Technical Specification 5.6.5).

6. NUREG-0979, Section 4.2.1.3.2. April 1983.

7. NUREG-0800, Section 15.4.9. Revision 2, July 1981.

8. NEDO-21778-A, "Transient Pressure Rises Affected Fracture Toughness Requirements for Boiling Water Reactors," December 1978.

9. NEDO-10527, "Rod Drop Accident Analysis for Large BWRs," (including Supplements 1 and 2), March 1972.

10. ASME. Boiler and Pressure Vessel Code.

11. 10 CFR

12. NEDO-21231, "Banked Position Withdrawal Sequence,"

January 1977.

Dresden 2 and 3 B 3.1.6-5 Revision 0

SLC System B 3.1.7 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.7 Standby Liquid Control (SLC) System BASES BACKGROUND The SLC System is designed to provide the capability of bringing the reactor, at any time in a fuel cycle, from full power and minimum control rod inventory (whichi.-is at the peak of the xenon transient) to a subcritical condition with the reactor in the most reactive, xenon free state without taking credit for control rod movement. The SLC System satisfies the requirements of 10 CFR 50.62 (Ref. 1) on anticipated transient without scram.

The SLC System consists of a boron solution storage tank.

two positive displacement pumps, two explosive valves that are provided in parallel for redundancy, and associated piping and valves used to transfer borated water from the storage tank to the reactor pressure vessel (RPV). The borated solution is discharged near the bottom of the core shroud, where it then mixes with the cooling water rising through the core. A smaller tank containing demineralized water is provided for testing purposes.

APPLICABLE The SLC System is manually initiated from the main control SAFETY ANALYSES room, as directed by the emergency operating procedures, if the operator determines the reactor cannot be shut down, or kept shut down, with the control rods. The SLC System is used in the event that enough control rods cannot be inserted to accomplish shutdown and cooldown in the normal manner. The SLC System injects borated water into the reactor core to add negative reactivity to compensate for all of the various reactivity effects that could occur during plant operations. To meet this objective, it is necessary to inject a quantity of boron, which produces a concentration of 600 ppm of natural boron, in the reactor coolant at 68 0 F. To allow for potential leakage and imperfect mixing in the reactor system, an amount of boron equal to 25% of the amount cited above is added (Ref. 2).

The volume versus concentration limits in Figure 3.1.7-1 and the temperature versus concentration limits in Figure 3.1.7-2 are calculated such that the required concentration is achieved accounting for dilution in the RPV with reactor water level at the high alarm point, including the water volume in the shutdown cooling piping, the (continued)

Dresden 2 and 3 B 3.1.7-1 Revision 0

SLC System B 3.1.7 BASES APPLICABLE recirculation loop piping, and portions of other piping SAFETY ANALYSES systems which connect to the RPV below the high alarm point.

(continued) This quantity of borated solution represented is the amount that is above the bottom of the boron solution storage tank.

However, no credit is taken for the portion of the tank volume that cannot be injected.

The SLC System satisfies rite on 4 of e 3

-i 10 CFR 50.36(c)(2)(ii).

LCO The OPERABILITY of the SLC System provides backup capability for reactivity control independent of normal reactivity control provisions provided by the control rods. The OPERABILITY of the SLC System is based on the conditions of the borated solution in the storage tank and the availability of a flow path to the RPV, including the OPERABILITY of the pumps and valves. Two SLC subsystems are required to be OPERABLE; each -contains an OPERABLE pump, an explosive valve, and associated piping, valves, and instruments and controls to ensure an OPERABLE flow path.

With one subsystem inoperable the requirements of 10 CFR 50.62 (Ref. 1) cannot be met, however, the remaining subsystem is still capable of shutting down the unit.

APPLICABILITY In MODES 1 and 2. shutdown capability is required. In MODES 3 and 4. control rods are not able to be withdrawn since the reactor mode switch is in shutdown and a control rod block is applied. This provides adequate controls to ensure that the reactor remains subcritical. In MODE 5.

only a single control rod can be withdrawn from a core cell rA/~&T I containing fuel assemblies. Demonstration of adequate SDM (LCO 3.1.1, "SHUTDOWN MARGIN (SDM)") ensures that the reactor will not become critical. Therefore, the SLC System is not required to be OPERABLE when only a single control rod can be withdrawn.

ACTIONS A.1 If one SLC subsystem is inoperable, the inoperable subsystem must be restored to OPERABLE status within 7 days. In this condition, the remaining OPERABLE subsystem is adequate to (continued)

Dresden 2 and 3 B 3-1.7-2 Revision 0

SLC System B 3.1.7 BASES ACTIONS A.1 (continued) shutdown the unit. However, the overall capability is reduced since the remaining OPERABLE subsystem cannot meet the requirements of Reference 1. The 7 day Completion Time is based on the availability of an OPERABLE subsystem capable of shutting down the reactof' and the low probability of a Design Basis Accident (DBA) or severe transient occurring concurrent with the failure of the Control Rod Drive (CRD) System to shut down the reactor.

B.1 If both SLC subsystems are inoperable, at least one subsystem must be restored to OPERABLE status within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. The allowed Completion Time of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> is considered acceptable given the low probability of a DBA or transient occurring concurrent with the failure of the control rods to shut down the reactor.

G=' C.1 any-7 1/"DE Z/ Aauta-s 36/ir If any Required Action and associated Completion Time is not ivs'*T Qare met, the plant must be brought to a MODE in which the LCO "does not apply. To achieve this status, the plant must be brought to MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The allowed Completion ime 12 hou is reasonable, based on operating k, exrrinceto rach experience, to reach,0 from full power conditions in an

'/ orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.1.7.1. SR 3.1.7.2, and SR 3.1.7.3 REOUIREMENTS SR 3.1.7.1 through SR 3.1.7.3 are 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Surveillances verifying certain characteristics of the SLC System (e.g..

the volume and temperature of the borated solution in the storage tank), thereby ensuring SLC System OPERABILITY without disturbing normal plant operation. These Surveillances ensure that the proper borated solution volume and temperature, including the temperature of the pump suction piping, are maintained. Maintaining a minimum specified borated solution temperature is important in ensuring that the boron remains in solution and does not C' (continued)

Dresden 2 and 3 B 3.1.7-3 Revision 0

SLC System B 3.1.7 BASES SURVEILLANCE SR 3.1.7.8 and SR 3.1.7.9 (continued)

REQUIREMENTS should be alternated such that both complete flow paths are tested every 48 months at alternating 24 month intervals.

The Surveillance may be performed in separate steps to prevent injecting boron into the RPV. An acceptable method for verifying'-flow from the pump to the RPV is-to~pump demineralized water from a test tank through one SLC subsystem and into the RPV. The 24 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has shown these components usually pass the Surveillance when performed at the 24 month Frequency:

therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

Demonstrating that all heat traced piping between the boron solution storage tank and the suction inlet to the injection pumps is unblocked ensures that there is a functioning flow path for injecting the sodium pentaborate solution. An

-acceptable method for verifying that the suction piping is unblocked is to pump from the storage tank to the storage tank.

The 24 month Frequency is acceptable since there is a low probability that the subject piping will be blocked due to precipitation of the boron from solution in the heat traced piping. This is especially true in light of the temperature verification of this piping required by SR 3.1.7.3.

However, if, in performing SR 3.1.7.3, it is determined that the temperature of this piping has fallen below the specified minimum, SR 3.1.7.9 must be performed once within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the piping temperature is restored to within the limits of Figure 3.1.7-2.

REFERENCES 1. 10 CFR 50.62.

2. UFSAR, Section 9.3.5.3.

Dresden 2 and B 3.1.7-6 Revision 0

SDV Vent and Drain Valves B 3.1.8 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.8 Scram Discharge Volume (SDV) Vent and Drain Valves BASES BACKGROUND The SDV vent and drain valves are normally open and discharge any accumulated water in the SDV to ensure that sufficient volume is available at all times to allow a complete scram. During a scram, the SDV vent and drain valves close to contain reactor water. The SDV is a volume of header piping that connects to each hydraulic control unit (HCU) and drains into an instrument volume. There are two SDVs (headers) and two instrument volumes, each receiving approximately one half of the control rod drive (CRD) discharges. Each instrument volume has a drain line with two valves in series. Each header is connected to a common vent line via two valves in series. The header piping is sized to receive and contain all the water discharged by the CRDs during a scram. The design and functions of the SDV are described in Reference 1.

APPLICABLE The Design Basis Accident and transient analyses assume all SAFETY ANALYSES of the control rods are capable of scramming. The acceptance criteria for the SDV vent and drain valves are that they operate automatically to:

a. Close during scram to limit the amount of reactor coolant discharged so that adequate core cooling is maintained and offsite doses remain within the limits of 10 CFR(j. 2); and

b. Open on scram reset to maintain the SDV vent and drain path open so that there is sufficient volume to accept the reactor coolant discharged during a scram.

Isolation of the SDV can also be accomplished by manual closure of the SDV valves. Additionally, the discharge of reactor coolant to the SDV can be terminated by scram reset or closure of the HCU manual isolation valves. For a boundin leakag ca e, the offsite doses are well within the limits of 10 CF 1IR (Ref. 2), and adequate core cooling is maintained (Ref. 3). The SDV vent and drain valves allow continuous drainage of the SDV during normal plant operation (continued)

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SDV Vent and Drain Valves B 3.1.8 BASES SURVEILLANCE SR 3.1.8.3 (continued)

REOUIREMENTS bounding leakage case evaluated in the accident analysis (Ref. 3). Similarly, after receipt of a simulated or actual scram reset signal, the opening of the SDV vent and drain valves is verified. The LOGIC SYSTEM FUNCTIONAL TEST in

.,, .1 LCO 3.3.1.1 and'the scram time testing of control rods in LCO 3.1.3, "Control Rod OPERABILITY." overlap this Surveillance to provide complete testing of the assumed safety function. The 24 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has shown these components usually pass the Surveillance when performed at the 24 month Frequency; therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

REFERENCES 1. UFSAR, Section 4.6.3.3.2.8.

2. 10 CFR

3. NUREG-0803, "Generic Safety Evaluation Report Regarding Integrity of BWR Scram System Piping,"

August 1981.

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LHGR B 3.2.3 B 3.2 POWER DISTRIBUTION LIMITS B 3.2.3 LINEAR HEAT GENERATION RATE (LHGR)

BASES BACKGROUND The LHGR is a measure of the heat generation rate of a fuel rod in a fuel assembly at any axial location. Limits on LHGR are specified to ensure that fuel design limits are not exceeded anywhere in the core during normal operation, (i.e., steady state). Exceeding the LHGR limit could potentially result in fuel damage and subsequent release of radioactive materials. Fuel design limits are specified to ensure that fuel system damage, fuel rod failure, or inability to cool the fuel does not occur during the normal operations and anticipated operating conditions identified in References 1 and 2.

APPLICABLE The analytical methods and assumptions used in evaluating SAFETY ANALYSES the fuel system design are presented in References 1 and 2.

The fuel assembly is designed to ensure (in conjunction with the core nuclear and thermal hydraulic design, plant equipment, instrumentation, and protection system) that fuel damage will not result in the release of radioactive materials in excess of the guidelines of 10 CFR. Parts 20 00 50 nd 1 . A mechanism that could cause fuel damage n rma operations and operational transients and that is considered in fuel evaluations is a rupture of the fuel a 0, rod cladding caused by strain from the relative expansion of the U02 pellet.

A value of 1% plastic strain of the fuel cladding has been defined as the limit below which fuel damage caused by overstraining of the fuel cladding is not expected to occur (Ref. 3).

Fuel design evaluations have been performed and demonstrate that the 1% fuel cladding plastic strain design limit is not exceeded during continuous operation with LHGRs up to the operating limit specified in the COLR. The analysis also includes allowances for short term transient excursions above the operating limit while still remaining within the AOO limits, plus an allowance for densification power spiking.

For Unit 2. flow-dependent LHGR limits, LHGRFAC(F), were designed to assure adherence to all fuel thermal-mechanical (continued)

Dresden 2 and 3 B 3.2.3-1 Revision 4

Primary Containment Isolation Instrumentation Primary Containment Isolation Instrumentation B 3.3.6.1 BASES APPLICABLE l.a. Reactor Vessel Water Level-Low Low (continued)

SAFETY ANALYSES, LCO, and (reference leg) and the pressure due to the actual water APPLICABILITY level (variable leg) in the vessel. Four channels of Reactor Vessel Water Level-Low Low Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function.

The Reactor Vessel Water Level-Low Low Allowable Value is chosen to be the same as the ECCS Reactor Vessel Water Level -Low Low Allowable Value (LCO 3.3.5.1) to ensure that the MSLs isolate on a potential loss of coolant acciden (LOCA) to prevent offsite doses from exceeding 10 CFR 1 limits.

This Function isolates the Group 1 valves.

1.b. Main Steam Line Pressure-Low Low MSL pressure indicates that there may be a problem with the turbine pressure regulation, which could result in a low reactor vessel water level condition and the RPV cooling down more than 100°F/hr if the pressure loss is allowed to continue. The Main Steam Line Pressure-Low Function is directly assumed in the analysis of the pressure regulator failure (Ref. 5). For this event, the closure of the MSIVs ensures that the RPV temperature change limit (100°F/hr) is not reached. In addition, this Function supports actions to ensure that Safety Limit 2.1.1.1 is not exceeded. (This Function closes the MSIVs prior to pressure decreasing below 785 psig, which results in a scram due to MSIV closure, thus reducing reactor power to < 25% RTP.)

The MSL low pressure signals are initiated from four pressure switches that are connected to the MSL header directly downstream of the main steam equalizing header.

The switches are arranged such that, even though physically separated from each other, each switch is able to detect low MSL pressure. Four channels of Main Steam Line Pressure-Low Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function.

(continued)

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Primary Containment Isolation Instrumentation B 3.3.6.1 BASES APPLICABLE I.d. Main Steam Line Flow-Hiqh (continued)

SAFETY ANALYSES, LCO, and the RPV water level decreases too far, fuel damage could APPLICABILITY occur. Therefore, the isolation is initiated on high flow to prevent or minimize core damage. The Main Steam Line Flow-High Function is directly assumed in the analysis of the main steam line break (MSLB) (Ref. 6). The isolation action, along with the scram function~of the Reactor Protection System (RPS), ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46 and offsite doses do not exceed the 10 CFR it The MSL flow signals are initiated from 16 differential pressure switches that are connected to the four MSLs (the differential pressure switches sense differential pressure across a flow restrictor). The differential pressure switches are arranged such that, even though physically separated from each other, all four connected to one MSL would be able to detect the high flow. Four channels of Main Steam Line Flow-High Function for each MSL (two channels per trip system) are available and are required to be OPERABLE so that no single instrument failure will preclude detecting a break in any individual MSL.

The Allowable Value is chosen to ensure that offsite dose limits are not exceeded due to the break.

This Function isolates the Group I valves.

I.e. Main Steam Line Tunnel Temperature-High Main steam line tunnel temperature is provided to detect a leak in the RCPB in the steam tunnel and provides diversity to the high flow instrumentation. Temperature is sensed in four different areas of the steam tunnel above each main steam line. The isolation occurs when a very small leak has occurred in any one of the four areas. If the small leak is allowed to continue without isolation, offsite dose limits may be reached. However, credit for these instruments is not taken in any transient or accident analysis in the UFSAR. since bounding analyses are performed for large breaks, such as MSLBs.

(continued)

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Primary Containment Isolation Instrumentation B 3.3.6.1 BASES APPLICABLE I.e. Main Steam Line Tunnel Temperature-High (continued)

SAFETY ANALYSES, LCO, and Main steam line tunnel temperature signals are initiated APPLICABILITY from temperature switches located in the four areas being monitored. Even though physically separated from each other, any temperature switch in any of the four areas is able to detect a leak. Therefore, sixteen channels of Main Steam Line Tunnel Temperature-High Function are available,,

but only eight channels (two channels in each of the four trip strings) are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function.

The Main Steam Line Tunnel Temperature-High Allowable Value is chosen to detect a leak of less than 1% rated steam flow.

These Functions isolate the Group 1 valves.

Primary Containment Isolation 2.a. Reactor Vessel Water Level-Low Low RPV water level indicates that the capability to cool the fuel may be threatened. The valves whose penetrations communicate with the primary containment are isolated to limit the release of fission products. The isolation of the primary containment on low RPV water level supports actions to ensure that offsite dose limits of 10 CFR 1 re not exceeded. The Reactor Vessel Water Level-Low Function associated with isolation is implicitly assumed in the UFSAR analysis as these leakage paths are assumed to be isolated post LOCA. S .

Reactor Vessel Water Level-Low signals are initiated from differential pressure transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel. Four channels of Reactor Vessel Water Level-Low Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function.

(continued)

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Primary Containment Isolation Instrumentation B 3.3.6.1 BASES APPLICABLE 2.a. Reactor Vessel Water Level-Low (continued)

SAFETY ANALYSES, LCO, and The Reactor Vessel Water Level-Low Allowable Value was APPLICABILITY chosen to be the same as the RPS Reactor Vessel Water Level-Low scram Allowable Value CLCO 3.3.1.1). since isolation of these valves is not critical to orderly plant shutdown.

This Function isolates the Group 2 and 3 valves.

2.b. Drywell Pressure-High High drywell pressure can indicate a break in the RCPB inside the primary containment. The isolation of some of the primary containment isolation valves on high drywell pressure su orts actions to ensure that offsite dose limits of 10 CFRR are not exceeded. The Drywell Pressure-High Function. associated with isolation of the primary containment, is implicitly assumed in the UFSAR accident analysis as these leakage paths are assumed to be isolated post LOCA.

High drywell pressure signals are initiated from pressure switches that sense the pressure in the drywell. Four channels of Drywell Pressure-High per Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function.

The Allowable Value was selected to be the same as the ECCS Drywell Pressure-High (LCO 3.3.5.1) and RPS Drywell Pressure-High (LCO 3.3.1.1) Allowable Values, since this may be indicative of a LOCA inside primary containment.

This Function isolates the Group 2 valves.

2.c. Drywell Radiation-High High drywell radiation indicates possible gross failure of the fuel cladding. Therefore, when Drywell Radiation-High is detected, an isolation is initiated to limit the release of fission products. However, this Function is not assumed in any accident or transient analysis in the UFSAR because other leakage paths (e.g.. MSIVs) are more limiting.

(continued)

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7 Primary Containment Isolation Instrumentation B 3.3.6.1 "BASES APPLICABLE 5.a. SLC System Initiation (continued)

SAFETY ANALYSES, LCO, and Two channels o the SLC System Initiation Function are APPLICABILITY available and are required to be OPERABLE o y in MODES 1 and 2. since these are the only MODES where he reactor can be critical these MODES are consistent with the e/2lr-ed/s also Applicability orthe SLC System (LCO 3.1.7).

be d /ek.1 6 //? There is no Allowable Value associated with this Function

/*ES NO lZ, a/d since the channels are mechanically actuated based solely on g, s,>ee/A s? the position of the SLC System initiation switch.

yS/'/llIs a This Function isolates the reactor water cleanup inboard and d yned-n&Y/0o outboard valves.

5.b. Reactor Vessel Water Level-Low

,olIa//aA e £7 Low RPV water level indicates that the capability to cool aLOCA LoRPwaeleeiniaethttecpblttool the fuel may be threatened. Should RPV water level decrease ank1s'ureI* a yA too far, fuel damage could result. Therefore, isolation of some interfaces with the reactor vessel occurs to isolate the potential sources of a break. The isolation of the RWCU

/a-M e A& System on low RPV water level supports actions to ensure e i,/ that the fuel peak cladding temperature remains below the A)a ler.limits of 10 CFR 50.46. The Reactor Vessel Water Level-Low Function associated with RWCU isolation is not directly assumed in the UFSAR safety analyses because the RWCU System line break is bounded by breaks of larger systems (recirculation and MSL breaks are more limiting).

Reactor Vessel Water Level -Low signals are initiated from four differential pressure transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel. Four channels of Reactor Vessel Water Level-Low Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function.

The Reactor Vessel Water Level-Low Allowable Value was chosen to be the same as the RPS Reactor Vessel Water Level-Low Allowable Value (LCO 3.3.1.1), since the capability to cool the fuel may be threatened.

This Function isolates the Group 3 valves.

(continued)

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Secondary Containment Isolation Instrumentation B 3.3.6.2 BASES BACKGROUND isolation function. For both Reactor Building Exhaust (continued) Radiation-High and Refueling Floor Radiation-High Functions, the secondary containment isolation trip system logic receives input from four channels. Two channels of Reactor Building Exhaust Radiation-High are located in each of the unit reactor building exhaust ducts and two channels of Refueling Floor Radiation-High are located where they can monitor the environment of each of the unit spent fuel pools. The output of the channels associated with Unit 2 are provided to one trip system while the output of the channels associated with Unit 3 are provided to the other trip system. The output from these channels are arranged in two one-out-of-two trip system logics for each Function to initiate the secondary containment isolation function. Any Reactor Building Exhaust Radiation-High or Refueling Floor Radiation-High channel will initiate the secondary containment isolation function. Initiating the secondary containment isolation function provides an input to both secondary containment Train A and Train B logic. Either train initiates isolation of all secondary containment isolation valves and provides a start signal to the associated SGT subsystem.

APPLICABLE The isolation signals generated by the secondary containment SAFETY ANALYSES, isolation instrumentation are implicitly assumed in the LCO, and safety analyses of Reference2 aný to initiate closure APPLICABILITY of the SCIVs and start the SGT System to limit offsite doses.

Refer to LCO 3.6.4.2, "Secondary Containment Isolation Valves (SCIVs)." and LCO 3.6.4.3. "Standby Gas Treatment (SGT) System," Applicable Safety Analyses Bases for more detail of the safety analyses.

The secondary containment isolation instrumentation satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii). Certain instrumentation Functions are retained for other reasons and are described below in the individual Functions discussion.

The OPERABILITY of the secondary containment isolation instrumentation is dependent on the OPERABILITY of the individual instrumentation channel Functions. Each Function (continued)

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Secondary Containment Isolation Instrumentation B 3.3.6.2 BASES APPLICABLE 3. 4. Reactor Building Exhaust Radiation-High and SAFETY ANALYSES, Refueling Floor Radiation-High LCO, and APPLICABILITY High reactor building exhaust radiation or refuel floor (continued) radiation is an indication of possible gross failure of the fuel cladding. The release may have originated from . the rimery containment due to a break in the RCPB n 3eei d refu ing floor e to a fuel andling ac den . en eactor Building Exhaust Radiation-HIig o efueling Floor

• 0 d, /A i17 */ Radiation-High is detected, secondary containment isolation X 1,61 C*redlle4/ and actuation of the SGT System are initiated to support iOr~z4efa / actions to limit the release of fission products as assumed in the UFSAR safety analyses (Refo. 2(5 ).4 iL;

-6(47 J The Reactor Building Exhaust Radiation-High signals are initiated from radiation detectors that are located on the ventilation exhaust duct coming from the associated reactor building. Therefore, the channels must be declared inoperable if the associated reactor building ventilation exhaust duct is isolated (i.e. a unit's Reactor Building Vent Isolation Damper and either unit's Standby Gas Treatment Reactor Building Inlet Damper 2(3)-7503 are closed). Refueling Floor Radiation-High signals are initiated from radiation detectors that are located on the refueling floor around the spent fuel storage pool. The signal from each detector is input to an individual monitor whose trip outputs are assigned to an isolation channel.

Four channels of Reactor Building Exhaust Radiation-High Function and four channels of Refueling Floor Radiation-High Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function.

The Allowable Values are chosen to promptly detect gross failure of the fuel cladding.

The Reactor Building Exhaust Radiation-High and Refueling Floor Radiation-High Functions are required to be OPERABLE in MODES 1, 2, and 3 where considerable energy exists in the RCS; thus, there is a probability of pipe breaks resulting in significant releases of radioactive steam and gas. In MODES 4 and 5, the probability and consequences of these events are low due to the RCS pressure and temperature limitations of these MODES; thus, these Functions are not required. In addition, the Functions are also required to (continued)

Dresden 2 and 3 B 3.3-6.2-6 Revision I

Secondary Containment Isolation Instrumentation B 3.3.6.2 BASES APPLICABLE 3. 4. Reactor Building Exhaust Radiation-High and SAFETY ANALYSES, Refueling Floor Radiation-High (continued)

L cU, dllU APPLICABILITY be OPERABLE during CO ALTERAT S, OPDRVs~and movement of irradiated fuel assemblies in e secondary containment.

because the capability of detecting radiation releases due

-eeen to fuel failures (due to fuel uncovery or dropped fuel assemblies) must be provided to ensure thad ffsite dose limits are not exceeded.-r ACTIONS A Note has been provided to modify the ACTIONS related to secondary containment isolation instrumentation channels.

Section 1.3, Completion Times, specifies that once a Condition has been entered, subsequent divisions, subsystems, components. or variables expressed in the Condition, discovered to be inoperable or not within limits, will not result in separate entry into the Condition.

Section 1.3 also specifies that Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for inoperable secondary containment isolation instrumentation channels provide appropriate compensatory measures for separate inoperable channels. As such, a Note has been provided that allows separate Condition entry for each inoperable secondary containment isolation instrumentation channel.

A.1 A,4q Because of the diversity of sensors available to provide isolation signals and the redundancy of the isolation design, an allowable out of service time of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> or 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> depending on the Function (12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for those Functions that have channel components common to RPS instrumentation and 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for those Functions that do not have channel components common to RPS instrumentation), hass been shown to be acceptable (Refs. 4 5 to permit restoration of any inoperable channe to OPERABLE status.

This out of service time is only acceptable provided the associated Function is still maintaining isolation capability (refer to Required Action B.1 Bases). If the inoperable channel cannot be restored to OPERABLE status (continued)

B 3.3.6.2-7 Revision 0 Dresden 2 and 3

Isolation Instrumentation Secondary Containment Secondary Containment Isolation Instrumentation B 3.3.6.2 BASES ACTIONS C.1.1, C.1.2, C.2.1, and C.2.2 (continued) maintain the secondary containment function. Isolating the associated penetration flow path(s) and starting the associated SGT subsystem (Required Actions C.1.1 and C.2.1) performs the intended function of the instrumentation and allows operation to continue. The method used to place the SGT subsystem in operation must provide for automatically reinitiating the subsystem upon restoration of power following a loss of~power to the SGT subsystem.

Alternately, declaring the associated SCIVs or SGT subsystem(s) inoperable (Required Actions C.1.2 and C.2.2) is also acceptable since the Required Actions of the respective LCOs (LCO 3.6.4.2 and LCO 3.6.4.3) provide appropriate actions for the inoperable components.

One hour is sufficient for plant operations personnel to establish required plant conditions or to declare the associated components inoperable without unnecessarily challenging plant systems.

SURVEILLANCE As noted at the beginning of the SRs, the SRs for each REQUIREMENTS Secondary Containment Isolation instrumentation Function are located in the SRs column of Table 3.3.6.2-1.

The Surveillances are modified by a Note to indicate that when a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> provided the associated Function maintains isolation capability. Upon completion of the Surveillance, or expiration of the 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken. This Note is based on the reliability analysis (Refs. 4

ýa 5) assumption of the average time required to perform channe surveillance. Thatt analysis demonstrated the 6 hour testing allowance does not significantly reduce the probability that the SCIVs will isolate the associated penetration flow paths and that thee SGT System will initiate when necessary. a (continued)

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Secondary Containment Isolation Instrumentation B 3.3.6.2 C l BASES SURVEILLANCE SR 3.3.6.2.2 (continued)

REQUIREMENTS is based on the reliability The Frequency of 92 day analysis of References and SR 3.3.6.2.3 3 of the actual Calibration of trip units provides a check if trip setpoints. The channel must be declared inoperable conservative than the trip setting is discovered to be less If the in Table 3.3.6.2-1.

the Allowable Value specified than conservative trip setting is discovered to be less setpoint methodology, but accounted for in the appropriate is still is not beyond the Allowable Value, performance Under the plant safety analysis.

within the requirements of readjusted to be these conditions, the setpoint must be than accounted for in the equal to or more conservative appropriate setpoint methodology.

reliability The Frequency of 92 dayz is b sed on the

• l*,-analysis of References (and SR 3.3.6.2.4 and SR 3.3.6.2.5 of the instrument A CHANNEL CALIBRATION is a complete check channel the loop and the sensor. This test verifies necessary the responds to the measured parameter within the channel CHANNEL CALIBRATION leaves range and accuracy. successive between adjusted to account for instrument drifts setpoint specific calibrations consistent with the plant methodology.

3.3.6.2.5 are based The Frequencies of SR 3.3.6.2.4 and SR month calibration on the assumption of a 92 day and a 24 of the interval, respectively, in the determination analysis.

magnitude of equipment drift in the setpoint (continued)

. . ii Revision 0 Dresden 2 and 3 B 3.3.6.Z-1i

Secondary Containment Isolation Instrumentation B 3.3.6.2 BASES SURVEILLANCE SR 3.3.6.2.6 REQUIREMENTS (continued) The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required isolation logic for a specific channel. The system functional testing performed on SCIVs and the SGT System in LCO 3.6.4.2 and LCO 3.6.4.3, respectively, overlaps this Surveillance to provide complete testing of the assumed safety function.

The 24 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power.

Operating experience has shown that these components usually pass the Surveillance when performed at the 24 month Frequency.

REFERENCES 1. UFSAR, Section 6.2.3.

2. UFSAR, Section 15.6.5.

• ---*.J UFSA~zW~cton 15 .7.d3 NEDC-31677P-A, "Technical Specification Improvement Analysis for BWR Isolation Actuation Instrumentation,"

July 1990.

NEDC-30851P-A Supplement 2. "Technical Specifications Improvement Analysis for BWR Isolation Instrumentation Common to RPS and ECCS Instrumentation," March 1989.

B 3.3.6.2-12 Revision 0 Dresden 2 and 3

CREV System Isolation Instrumentation B 3.3.7.1 BASES (continued)

APPLICABLE The ability of the CREV System to maintain the habitability SAFETY ANALYSES of the control room emergency zone is explicitly assumed for certain accidents as discussed in the UFSAR safety analyses (Refs. 1, and 2). CREV System operation ensures that the radiation exposure of control room personnel, through the duration of any one of the postulated accidents, does not exceed the limits set by GDC of 10 C 50, Appen CREV System instrumentation satisfies Criterion 3 oof 10 CFR 50.36(c)(2)(ii). /0 0 LCO High reactor building ventilation exhaust radiation is an indication of possible gross failure of the fuel cladding.

The release may have originated from the primary containment due to a break in he RCPB or he refuelVg floor duýýo a

(:-fueji-fandling accent. When high reactor building ventilation exhaust radiation is alarmed in the control room, the CREV System is manually initiated in the isolation/pressurization mode and required dampers are closed since this condition could result in radiation exposure to control room personnel.

The Reactor Building Ventilation System-High High Radiation Alarm Function signals are initiated from radiation detectors that are located in the ventilation exhaust ducting coming from the reactor building and refueling zones. The signals from each detector are input to individual monitors whose trip outputs are assigned to a control room alarm. Two channels of Reactor Building Ventilation System-High High Radiation Alarm Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the alarm function.

The Allowable Value was selected to promptly detect gross failure of the fuel cladding and to ensure protection of control room personnel. Each channel must have its setpoint set within the specified Allowable Value in SR 3.3.7.1.3.

The actual setpoint is calibrated consistent with applicable setpoint methodology assumptions. Nominal trip setpoints are specified in the setpoint calculations. The nominal setpoints are selected to ensure that the setpoints do not exceed the Allowable Value between successive CHANNEL (continued)

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CREV System Isolation Instrumentation B 3.3.7.1 BASES LCO CALIBRATIONS. Operation with a trip setpoint less (continued) conservative than the nominal trip setpoint, but within its Allowable Value, is acceptable. A channel is inoperable if its actual trip setpoint is not within its required Allowable Value.

Trip setpoints are those predetermined values of output at wh*i~h an action should take place. The setpoints are compared to the actual process parameter (e.g., reactor building ventilation exhaust radiation), and when the measured output value of the process parameter exceeds the setpoint, the associated device (e.g.,trip unit) changes state. The analytic limits are derived from the limiting values of the process parameters obtained from the safety analysis. The trip setpoints are determined from the analytic limits, corrected for defined process, calibration, and instrument errors. The Allowable Values are then determined, based on the trip setpoint values, by accounting for the calibration based errors. These calibration based errors are limited to reference accuracy, instrument drift, errors associated with measurement and test equipment, and calibration tolerance of loop components. The trip setpoints and Allowable Values determined in this manner provide adequate protection because instrument uncertainties, process effects, calibration tolerances, instrument drift, and severe environment errors (for channels that must function in harsh environments as defined by 10 CFR 50.49) are accounted for and appropriately applied for the instrumentation.

APPLICABILITY The Reactor Building Ventilation System-High High Radiation Alarm Function is required to be OPERABLE in MODES 1, 2, and 3 and during movement of.irradiated fuel assemblies in the secondary containment,

• A ýATlS, and operations with a potential for draining the reactor vessel (OPDRVs),

to ensure that control room personnel can be protected during a LOCA, fuel handling event, or vessel draindown event. During MODES 4 and 5, when these specifie U conditions are not in progress (e.g..COP ALTER ION ), the probabilit of a LOCA or fuel damage is low; thus, the

- F*unctions not required.ue:

(continued)

Dresden 2 and 3 B 3.3.7.1-3 Revision 0

RCS Specific Acti vi ty B 3.4.6 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.6 RCS Specific Activity BASES BACKGROUND During circulation, the reactor coolant acquires radioactive materials due to release of fission products from fuel leaks into the reactor coolant and activation of corrosion products in the reactor coolant. These radioactive materials in the reactor coolant can plate out in the RCS, and, at times, an accumulation will break away to spike the normal level of radioactivity. The release of coolant during a Design Basis Accident (DBA) could send radioactive materials into the environment.

Limits on the maximum allowable level of radioactivity in the reactor coolant are established to ensure that in the event of a release of any radioactive material to the environment during a DBA, radiation doses are maintained within the limits of 10 CFR 1 (-Ref. 1).

7 This LCO contains iodine specific activity limits. The iodine isotopic activities per gram of reactor coolant are expressed levels in terms of a DOSE EQUIVALENT 1-131.

allowable are intended to limit the 2 hourThe radiation dose to an individual at he site boundary to a small fraction of the 10 CFR limit.

APPLICABLE Analytical methods and assumptions involving radioactive SAFETY ANALYSES material in the primary coolant are presented in the UFSAR (Ref. 2). The specific activity in the reactor coolant (the source term) is an initial condition for evaluation of the consequences of an accident due to a main steam line break (MSLB) outside containment. No fuel damage is postulated in the MSLB accident, and the release of radioactive material to the environment is assumed to end when the main steam isolation valves (MSIVs) close completely.

EZ~6 This MSLB release forms the basis for determinien offsite and control room doses (Ref. 2). The limits on the specific activit of the rimary coolant ensure that the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> th oid a whol ody doseerat the site boundary, resulting (continued)

Dresden 2 and 3 B 3.4.6-1 Revision 0

RCS Specific Activity B 3.4.6 BASES APPLICABLE from an MSLB outside containment during steady state SAFETY ANALYSES operation, will not exceed(the dose guidelines of (continued) 10 CFR The limits on the specific- ctivity of the primary coolant also ensure the th id ose to control room operators, resulting from a MSLB outside containment during SOs 7 steady state oper aion will not exceed the limits( .;

10 C 50, App dix A (Ref. 3).

6f /T # , The limit on specific activity is a value from a parametric evaluation of typical site locations. This limit is conservative because the evaluation considered more restrictive parameters than for a specific site, such as the location of the site boundary and the meteorological conditions of the site.

RCS specific activity satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

LCO Thie sp - IF n*1O-Te lvi is limited to < 0.2 pCi/gm DOSE

{EQUIVALE.NT 1-1311.- This limit ensures the source term assme innv*

the nssafe-st y analysis

, +÷ for" the MSLB is not exceeded, so any release of radioactivity to the environment during an MSLB is less than a small fraction of the 10 CFR2limits

~~ and 9;/9FRýI-19 of 1_ FR 50. App 'dix AC(R . 3).

IAPPLICABILITY In MODE 1. and MODES 2 and 3 with any main steam line not isolated, limits on the primary coolant radioactivity are Slzapplicable

• since there is an escape path for release of

,g a',ze 1 *radioactive material from the primary coolant to the

/d._* *Aenvironment in the event of an MSLB outside of primary

,>i~r,~/7,'G,* *containment.

In MODES 2 and 3 with the main steam lines isolated, such limits do not apply since an escape path does not exist.

In MODES 4 ana 5, no limits are required since the reactor is not pressurized and the potential for leakage is reduced.

ACTIONS A.1 and A.2 When the reactor coolant specific activity exceeds the LCO DOSE EQUIVALENT 1-131 limit, but is : 4.0 pCi/gm, samples must be analyzed for DOSE EQUIVALENT 1-131 at least once (continued)

Dresden 2 and 3 B 3.4.6-2 Revision 0

Activity RCS Specific RCS Specific Activity B 3.4.6 BASES ACTIONS A.1 and A.2 (continued) every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. In addition, the specific activity must be restored to the LCO limit within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. The Completion Time of once every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is based on the time needed to take and analyze a sample. The 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> Completion Time to restore the activity level provides a reasonable time for temporary coolant'rectivity increases (iodine spikes.or crud bursts) to be cleaned up with the normal processing systems.

A Note to the Required Actions of Condition A excludes the MODE change restriction of LCO 3.0.4. This exception allows entry into the applicable MODE(S) while relying on the ACTIONS even though the ACTIONS may eventually require plant shutdown. This exception is acceptable due to the significant conservatism incorporated into the specific activity limit, the low probability of an event which is limiting due to exceeding this limit, and the ability to restore transient specific activity excursions while the plant remains at, or proceeds to power operation.

B.1, B.2.1, B.2.2.1 and B.2.2.2 a'as -dene SIf the DOSE EQUIVALENT 1-131 cannot be restore to 5 0.

pCi/gm within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, or if at any tim it is > 4.0 PCi/gm, it must be determined at least once every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and all the main steam lines must be solated within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. Isolating the main steam lines precludes the possibility of releasing radioact've material to the environment in an amount that ore than a small fraction GDC 19f 10 C 50)

Mof the requirements of 10 CFR

?9iJ2/-"eJ d1 S*Z*7 Appe ix A . 3) during a postulated MSLB accident.

ddgse 4 'M7-s**ternatively, the plant can be placed in MODE 3 within

ý*-12 hours and in MODE 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. This option is provided for those instances when isolation of main steam lines is not desired (e.g., due to the decay heat loads).

In MODE 4. the requirements of the LCO are no longer applicable.

The Completion Time of once every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is the time needed to take and analyze a sample. The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Completion Time is reasonable, based on operating experience, to isolate the main steam lines in an orderly manner and without (continued)

B 3.4.6-3 Revision 0 Dresden 2 and 3

RCS Specific Activity B 3.4.6 BASES ACTIONS B.1, B.2.1, B.2.2.1, and B.2.2.2 (continued) challenging plant systems. Also, the allowed Completion Times for Required Actions B.2.2.1 and B.2.2.2 for placing the unit in MODES 3 and 4 are reasonable, based on operating experience, to achieve the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. .LV SURVEILLANCE SR 3.4.6.1 REQUIREMENTS This Surveillance is performed to ensure iodine remains within limit during normal operation. The 7 day Frequency is adequate to trend changes in the iodine activity level.

This SR is modified by a Note that requires this Surveillance to be performed only in MODE I because the level of fission products generated in other MODES is much less.

REFERENCES 1. 10 CFR.

2. UFSAR, Section 15.6.4.

Dresden 2 and 3 B 3.4.6-4 Revision 0

Primary Containment B 3.6.1.1 BASES BACKGROUND This Specification ensures that the performance of the (continued) primary containment, in the event of a Design Basis Accident (DBA). meets the assumptions used in the safety analyses of References 1 and 2. SR 3.6.1.1.1 leakage rate requirements are in conformance with 10 CFR 50, Appendix J, Option B (Ref. 3), as modified by approved exemptions.

APPLICABLE The safety design basis for the primary containment is that SAFETY ANALYSES it must withstand the pressures and temperatures of the limiting DBA without exceeding the design leakage rate.

The DBA that postulates the maximum release of radioactive material within primary containment is a LOCA. In the analysis of this accident, it is assumed that primary containment is OPERABLE such that release of fission products to the environment is controlled by the rate of primary containment leakage.

Analytical methods and assumptions involving the primary containment are presented in References 1 and 2. The safety analyses assume a nonmechanistic fission product release following a DBA, which forms the basis for determination of offsite doses. The fission product release is, in turn, based on an assumed leakage rate from the primary containment. OPERABILITY of the primary containment ensures that the leakage rate assumed in the safety analyses is not exceeded.

The maximum allowable akage rate for the primary containment (L,) is 1 by wei ht of the containment air per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at th esi n sis 0 A pte alcu containment pressure a of 48 psig.

Primary containment satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii).

LCO Primary containment OPERABILITY is maintained by limiting leakage to

• 1.0 La, except prior to the first startup after performing a required Primary Containment Leakage Rate Testing Program leakage test. At this time, applicable leakage limits must be met. In addition, the leakage from the drywell to the suppression chamber must be limited to (continued)

Dresden 2 and 3 B 3.6.1.1-2 Revision 0

Primary Containment Air Lock B 3.6.1.2 BASES BACKGROUND containment leakage rate to within limits in the event of a (continued) DBA. Not maintaining air lock integrity or leak tightness may result in a leakage rate in excess of that assumed in the safety analysis.

APPLICABLE The DBA that postulates the maximum release of radioactive SAFETY ANALYSES material within primary containment is a LOCA. In the analysis.of..-this accident, it is assumed that primary containment is OPERABLE, such that release of fission products to the environment is controlled by the rate of primary containment leakage. The primary containment i designed with a maximum allowable leakage rate (L,) of Z f the containment air mass per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at sýign asis 0 eakcalcute containment pressureW(

of 48 psig RThis al owable leakage rate forms the basis for the acceptance criteria imposed on the SRs associated with the air lock.

Primary containment air lock OPERABILITY is also required to minimize the amount of fission product gases that may escape primary containment through the air lock and contaminate and pressurize the secondary containment.

The primary containment air lock satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii).

LCO As part of the primary containment pressure boundary, the air lock safety function is related to control of containment leakage following a DBA. Thus, the air lock structural integrity and leak tightness are essential to the successful mitigation of such an event.

The primary containment air lock is required to be OPERABLE.

For the air lock to be considered OPERABLE, the air lock interlock mechanism must be OPERABLE, the air lock must be in compliance with the Type B air lock leakage test, and both air lock doors must be OPERABLE. The interlock allows only one air lock door to be opened at a time. This provision ensures that a gross breach of primary containment does not exist when primary containment is required to be OPERABLE. Closure of a single door in the air lock is (continued)

Dresden 2 and 3 B 3.6.1.2-2 Revision 0

PCIVs B 3.6.1.3 BASES SURVEILLANCE SR 3.6.1.3.6 REQUIREMENTS (continued) Verifying that the isolation time of each MSIV is within the specified limits is required to demonstrate OPERABILITY.

The isolation time test ensures that the MSIV will isolate in a time period that does not exceed the times assumed in the DBA and transient analyses. This ensures that the calculated radio9ogical consequences of these events remain within"1YO *CR limits. The Frequencyof this SR is in accordance with he requirements of the Inservice Testing Program.

5o~

SR 3.6.1.3.7 Automatic PCIVs close on a primary containment isolation signal to prevent leakage of radioactive material from primary containment following a DBA. This SR ensures that each automatic PCIV will actuate to its isolation position on a primary containment isolation signal. The LOGIC SYSTEM FUNCTIONAL TEST in LCO 3.3.6.1. "'Primary Containment Isolation Instrumentation," overlaps this SR to provide complete testing of the safety function. The 24 month Frequency was developed considering it is prudent that this Surveillance be performed only during a unit outage since isolation of penetrations would eliminate cooling water flow and disrupt the normal operation of many critical components. Operating experience has shown that these components usually pass this Surveillance when performed at the 24 month Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

SR 3.6.1.3.8 This SR requires a demonstration that each reactor instrumentation line excess flow check valve (EFCV) is OPERABLE by verifying that the valve actuates to the isolation position on an actual or simulated instrument line break condition. This test is performed by blowing down the instrument line during an inservice leak or hydrostatic test and verifying a distinctive "click" when the poppet valve seats or a quick reduction in flow. This SR provides assurance that the instrumentation line EFCVs will perform as designed. The 24 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned (continued)

Dresden 2 and 3 B 3.6.1.3-13 Revision 0

PCIVs B 3.6.1.3 BASES SURVEILLANCE SR 3.6.1.3.8 (continued)

REQUIREMENTS transient if the Surveillance were performed with the reactor at power. Operating experience has shown that these components usually pass this Surveillance when performed at the 24 month Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

SR 3.6.1.3.9 The TIP shear isolation valves are actuated by explosive charges. An in place functional test is not possible with this design. The explosive squib is removed and tested to provide assurance that the valves will actuate when required. The replacement charge for the explosive squib shall be from the same manufactured batch as the one fired or from another batch that has been certified by having one of the batch successfully fired. The Frequency of 24 months on a STAGGERED TEST BASIS is considered adequate given the administrative controls on replacement charges and the frequent checks of circuit continuity (SR 3.6.1.3.4). Other administrative controls, such as those that limit the shelf life and operating life, as applicable, of the explosive charges must be followed.

SR 3.6.1.3.10 The analyses in References 2 and 3 are based on leakage that is less than the specified leakage rate. The leakage rate of each main steam isolation valve path is assumed to be the maximum pathway leakage (leakage through the worse of the two isolation valves). If both isolation valves in the penetration are closed the actual leakage rate is the lesser leakage rate of the two valves. This method of quantifying leakage is only to be used for this SR (i.e., Appendix J maximum pathway leakage limits are to be quantified in accordance with t riry Containment Leakage Rate Testing Program). /Th combined leaka ethrough all MSl eakage pa *s must < 46 scfh whe ~dtested at > 25 i. i ensures that MSI kage is proper y e or ins determining the overall primary containment leakage rate.

The Frequency is required by the Primary Containment Leakage Rate Testing Program.

MSIV leakage is considered part of L, (continued)

Dresden 2 and 3 B 3.6.1.3-14 Revision 0

Secondary Containment B 3.6.4.1 B 3.6 CONTAINMENT SYSTEMS B 3.6.4.1 Secondary Containment BASES BACKGROUND The function of the secondary containment is to contain, dilute, and hold up fission products that may leak from primary containment following 2 Design Basis Accident (DBA).

In conjunction with operation of the Standby Gas Treatment (SGT) System and closure of certain valves whose lines penetrate the secondary containment, the secondary containment is designed to reduce the activity level of the fission products prior to release to the environment and to isolate and contain fission products that are released during certain operations that take place inside primary containment, when primary containment is not required to be OPERABLE, or that take place outside primary containment.

The secondary containment is a structure that completely encloses both primary containments and those components that may be postulated to contain primary system fluid. This structure forms a control volume that serves to hold up and dilute the fission products. It is possible for the pressure in the control volume to rise relative to the environmental pressure (e.g.. due to pump and motor heat load additions). To prevent ground level exfiltration while allowing the secondary containment to be designed as a conventional structure, the secondary containment requires support systems to maintain the control volume pressure' at less than the external pressure. Requirements for these systems are specified separately in LCO 3.6.4.2. "Secondary Containment Isolation Valves (SCIVs)," and LCO 3.6.4.3, "Standby Gas Treatment (SGT) System."

• ~I~s APPLICABLE Th ee a two p cipal accident& for wIT-credit ji~s taken SAFETY ANALYSES for s?5con"'aary conanment OPERABILITY. TN e are loss of coo jt.L444eaL (LOCA) (Ref. 1 a fel han 'ng c2). e The perRorms no-Asi active function in response to ea of tes imiting even1t; however, its leak tightness is required to ensure that the release of radioactive materials from the primary containment is restricted to those leakage paths and (continued)

Dresden 2 and 3 B 3.6.4.1-1 Revision 0

Secondary Containment B 3.6.4.1 BASES analysis APPLICABLE associated leakage rates assumed in the accident the secondary SAFETY ANALYSES and that fission products entrapped within containment structure will be treated by the SGT System (continued) prior to discharge to the environment.

of Secondary containment satisfies Criterion 3 10 CFR 50.36(c)(2)(ii).

a control volume LCO An OPERABLE secondary containment provides into which fission products that bypass or leak from primary from the reactor coolant containment, or are released pressure boundary components located in secondary to release containment, can be diluted and processed prior For the secondary containment to be to the environment. to adequate leak tightness considered OPERABLE, it must have be established and ensure that the required vacuum can be closed maintained, the hatches and blowout panels must welds, bellows, or and sealed, the sealing mechanisms (e.g.,

O-rings) associated with each secondary containment penetration must be OPERABLE (such that secondary all inner and containment leak tightness can be maintained),

access or all outer doors in each secondary containment opening must be closed.

to a fission product APPLICABILITY In MODES 1, 2, and 3. a LOCA could lead secondary release to primary containment that leaks to containment. Therefore, secondary containment OPERABILITY that is required during the same operating conditions require primary containment OPERABILITY.

of the In MODES 4 and 5. the probability and consequences LOCA are reduced due to the pressure and temperature limitations in these MODES. Therefore, maintaining secondary containment OPERABLE is not required in MODE 4 other situations or 5 to ensure a control volume, except for releases of radioactive material can for which significant a o tial be postulated, such as during operations with (OPDRVs , ring RE rE eeta for drainnn the reactor vessel ERATJ S), or during movement of irradiated fuel assemblies in the secondary containment.

continued)

C A 1-9 Revision 0 D J. - -

Dresden 2 and 3

Secondary Containment B 3.6.4.1 BASES (continued)

ACTIONS A.1 If secondary containment is inoperable, it must be restored to OPERABLE status within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Completion Time provides a period of time to correct the problem that is commensurate with the importance of maintaining secondary also containment during MODES 1. 2. and 3. This time period ensures that the probability of an accident (requiring secondary containment OPERABILITY) occurring.during periods where secondary containment is inoperable is minimal.

B.1 and B.2 If secondary containment cannot be restored to OPERABLE must status within the required Completion Time, the plant LCO does not apply. To be brought to a MODE in which the be brought to at least achieve this status, the plant must The MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and to MODE 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

allowed Completion Times are reasonable, based on operating full experience, to reach the required plant conditions from power conditions in an orderly manner and without challenging plant systems.

Movement of irradiated fuel assemblies in the secondary containment, CO ALTERATIUNS.and OPDRVs can be postulated to caus fission product release to the secondary containment. In such cases, the secondary containment is the only barrier to release of fission products he environment. OFR__LTERATIOfg" an d movement of irradiated fuel assemblies mt be immediately\suspended if the secondary containment is inoperable. -Aere ere,

ý oenson h e acti Su s p en suisp _i:sshall not preclude completing 7ýE an action that involves moving a component to a safe position. Also, action must be immediately initiated to suspend OPDRVs to minimize the probability of a vessel draindown and subsequent potential for fission product release. Actions must In i OPDRVs are suspended.

Required Action C.1 ha4 een mo ied by a Note stating that LCO 3.0.3 is not applicable. If moving irradiated fuel assemblies while in MODE 4 or 5, LCO 3.0.3 would not (continued)

B 3.6.4.1-3 Revision 0 Dresden 2 and 3

Secondary Containment B 3.6.4.1 BASES ACTIONS C.1 . n 3(continued) o specify any action. If movin irradiated fuel assemblies and 1.2 while in MODE 1, 2, or 3. the fuel movement is independent of reactor operations. Therefore. in either case. inability to suspend movement ofkirradiated fuel assemblies would not be a sufficient reason to requre c or shut own.

SURVEILLANCE SR 3.6.4.1.1 REQUIREMENTS This SR ensures that the secondary containment boundary is sufficiently leak tight to preclude exfiltration under expected wind conditions. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Frequency of this SR was developed based on operating experience related to secondary containment vacuum variations during the applicable MODES and the low probability of a DBA occurring.

Furthermore, the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Frequency is considered adequate in view of other indications available in the control room, including alarms, to alert the operator to an abnormal secondary containment vacuum condition.

SR 3.6.4.1.2 and SR 3.6.4.1.4 Verifying that one secondary containment access door in each access opening is closed and each equipment hatch is closed and sealed ensures that the infiltration of outside air of such a magnitude as to prevent maintaining the desired negative pressure does not occur. Verifying that all such openings are closed provides adequate assurance that exfiltration from the secondary containment will not occur.

In this application, the term "sealed" has no connotation of leak tightness. In addition, for equipment hatches that are floor plugs, the "sealed" requirement is effectively met by gravity. Maintaining secondary containment OPERABILITY requires verifying one door in the access opening is closed.

An access opening contains one inner and one outer door. In some cases a secondary containment barrier contains multiple inner or multiple outer doors. For these cases, the access openings share the inner door or the outer door, i.e., the access openings have a common inner door or outer door. The intent is to not breach the secondary containment at any (continued)

Dresden 2 and 3 B 3.6.4.1-4 Revision 0

Secondary Containment B 3.6.4.1 BASES SURVEILLANCE SR 3.6.4.1.3 (continued)

REQUIREMENTS addition to the requirements of LCO 3.6.4.3. either SGT subsystem will perform this test. The inoperability of the SGT System does not necessarily constitute a failure of this Surveillance relative to secondary containment OPERABILITY.

Operating experience has shown the secondary containment boundary usually passes the Surveillance when performed at the 24 month Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

REFERENCES 1. UFSAR, Section 15.6.5.

B 3.6.4.1-6 Revision 0 Dresden 2 and 3

SCIVs B 3.6.4.2 B 3.6 CONTAINMENT SYSTEMS B 3.6.4.2 Secondary Containment Isolation Valves (SCIVs)

BASES BACKGROUND The function of the SCIVs. in combination with other accident mitigation systems, is to limit fission product release during and f.ovfiinq ostulated Design Basis 1, Accidents (DBAs) (Relý I an( ). Secondary containment isolation within the time limits specified for those isolation valves designed to close automatically ensures that fission products that leak from primary containment following a DBA, or that are released during certain operations when primary containment is not required to be OPERABLE or take place outside primary containment, are maintained within the secondary containment boundary.

The OPERABILITY requirements for SCIVs help ensure that an adequate secondary containment boundary is maintained during and after an accident by minimizing potential paths to the environment. These isolation devices consist of either passive devices or active (automatic) devices. Manual valves, de-activated automatic valves secured in their closed position (including check valves with flow through the valve secured), and blind flanges are considered passive devices.

Automatic SCIVs (i.e.. dampers) close on a secondary containment isolation signal to establish a boundary for untreated radioactive material within secondary containment following a DBA or other accidents.

Other penetrations required to be closed during accident conditions are isolated by the use of valves in the closed position or blind flanges.

APPLICABLE The SCIVs must be OPERABLE to ensure the secondary SAFETY ANALYSES containment barrier to fission product releases is /S established. The -pa accidentjfor which the j secondary containment boundary is required a loss of coolan accident (Ref. 1) an fu-e ing a en Re ) The secondary conb-ainment performs no activ*

fuTion in response toit h of t se limiting event. but the boundary establishedby SCV required to ensure that (continued)

Dresden 2 and 3 B 3.6.4.2-1 Revision 0

SCIVs B 3.6.4.2 BASES APPLICABLE leakage from the primary containment is processed by the SAFETY ANALYSES Standby Gas Treatment (SGT) System before being released to (continued) the environment.

Maintaining SCIVs OPERABLE with isolation times within limits ensures that fission products will remain trapped inside secondary containment so that they can be treated by the SGT System prior to discharge to the environment.

SCIVs satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).

LCO SCIVs form a part of the secondary containment boundary.

The SCIV safety function is related to control of offsite radiation releases resulting from DBAs.

The power operated, automatic, isolation valves are considered OPERABLE when their isolation times are within limits and the valves actuate on an automatic isolation signal. The valves covered by this LCO. along with their associated stroke times, are listed in the Technical Requirements Manual (Ref.

The normally closed manual SCIVs are considered OPERABLE when the valves are closed and blind flanges are in place, or open under administrative controls. These passiv isolation valves or devices are listed in Reference APPLICABILITY In MODES 1; 2, and 3, a DBA could lead to a fission product release to the primary containment that leaks to the secondary containment. Therefore, the OPERABILITY of SCIVs is required.

In MODES 4 and 5. the probability and consequences of these events are reduced due to pressure and temperature limitations in these MODES. Therefore, maintaining SCIVs OPERABLE is not required in MODE 4 or 5. except for other situations under which significant radioactive releases can be postulated, such as during operations with a otential nill* (OPDRVs) reactor vessel R _r Oin for the draiP for drainin A RAT S. or during movement ofoirradiated fuei _ rE I assemblies in the secondary containment.

(continued)

JJAJSe 2 6?r Dresden 2 and 3 B 3.6.4.2-2 Revision 0

SCIVs B 3.6.4.2 BASES ACTIONS B.1 (continued)

The Condition has been modified by a Note stating that Condition B is only applicable to penetration flow paths with two isolation valves. This clarifies that only Condition A is entered if one SCIV is inoperable in each of two penetrations.

C.1 and C.2 If any Required Action and associated Completion Time cannot be met, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and to MODE 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

If any Required Action and associated Completion Time are not met, the plant must be placed in a condition in which LCO does not the movement apply. If applicable, OR LTERATIOD and)

.the ofiirradtated fuel assemblies in the secon-Tary3 containment must be immediately sus ended. Sus ensi of

- e actl eshall not preclude completion of movement i of a component to a safe position. Also, if applicable.

actions must be immediately initiated to suspend OPDRVs in order to minimize the probability of a vessel draindown and the subsequent potential for fission product release.

Actions must continue until OPDRVs are suspended.

Required Action D.1 has been modified by a Note stating that LCO 3.0.3 is not applicable. If moving irradiated fuel assemblies while in MODE 4 or 5. LCO 3.0.3 would not specify any action. If moving fuel while in MODE 1. 2. or 3, the fuel movement is independent of reactor operations.

Therefore, in either case, inability to suspend movement of 1\_.ýirradiated fuel assemblies would not be a sufficient reason to require a reactor shutdown.

(continued)

B 3.6.4.2-5 Revision 0 Dresden 2 and 3

SCIVs B 3.6.4.2 BASES SURVEILLANCE SR 3.6.4.2.2 REQUIREMENTS (continued) Verifying that the isolation time of each power operated, automatic SCIV is within limits is required to demonstrate OPERABILITY. The isolation time test ensures that the SCIV will isolate in a time period less than or equal to that assumed in the safety analyses. The Frequency of this SR is 92 days.

SR 3.6.4.2.3 Verifying that each automatic SCIV closes on a secondary containment isolation signal is required to prevent leakage of radioactive material from secondary containment following a DBA or other accidents. This SR ensures that each automatic SCIV will actuate to the isolation position on a secondary containment isolation signal. The LOGIC SYSTEM FUNCTIONAL TEST in LCO 3.3.6.2, "Secondary Containment Isolation Instrumentation." overlaps this SR to provide complete testing of the safety function. While this Surveillance can be performed with the reactor at power, operating experience has shown these components usually pass the Surveillance when performed at the 24 month Frequency.

which is based on the refueling cycle. Therefore, the Frequency was concluded to be acceptable from a reliability, standpoint.

REFERENCES 1. UFSAR. Section 15.6.5.

cUFSA. =15.3.

lSection

""T4n Technical Requirements Manual.

Dresden 2 and 3 B 3.6.4.2-7 Revision 0

SGT System B 3.6.4.3 BASES BACKGROUND The demister is provided to remove entrained water in the (continued) air, while the electric heater reduces the relative humidity of the airstream to less than 70% (Ref. 2). The prefilter removes large particulate matter, while the HEPA filter removes fine particulate matter and protects the charcoal from fouling. The charcoal adsorber removes gaseous elemental iodine and organic iodides, and the final HEPA filter collects any carbon fines exhausted from the charcoal adsorber.

The SGT System automatically starts and operates in response to actuation signals indicative of conditions or an accident that could require operation of the system. Following initiation, the primary charcoal filter train inlet damper opens, the cooling damper closes, the associated fan starts, and the fan discharge damper opens. When sufficient flow develops, the heater turns on and the flow control damper begins modulating to control system flow and maintain a negative pressure in the secondary containment. If, either a low flow or heater off condition exists for the primary subsystem after 20 seconds, the primary subsystem is tripped and the standby SGT subsystem starts.

omiti.ate h APPLICABLE The design basis for the SGT System /is SAFETY ANALYSES conse uences of a loss of coolant ccident and el han ing acc ts_(Ref 3{a*). For t se analyzed eve , t sGT System is assumea to b an y initiated 40 to reduce, via filtration and adsorption, the ra loac ve material released to the environment.

The SGT System satisfies 10 CFR 50.36(c)(2)(ii).

LCO Following a DBA, a minimum of one SGT subsystem is required to maintain the secondary containment at a negative pressure with respect to the environment and to process gaseous releases. Meeting the LCO requirements for two OPERABLE subsystems ensures operation of at least one SGT subsystem in the event of a single active failure.

APPLICABILITY In MODES 1. 2. and 3, a DBA could lead to a fission product release to primary containment that leaks to secondary containment. Therefore, SGT System OPERABILITY is required during these MODES.

(continued)

B 3.6.4.3-2 Revision 0 Dresden 2 and 3

SGT System B 3.6.4.3 BASES APPLICABILITY In MODES 4 and 5. the probability and consequences of these (continued) events are reduced due to the pressure and temperature limitations in these MODES. Therefore, maintaining the SGT System in OPERABLE status is not required in MODE 4 or 5, except for other situations under which significant releases of radioactive material can be postulated, such as during operation wit ntial f aining the reactor vessel (OPDRVs ing C ALTER ONSor during movement of irradiaO wuel assemb the, secondary containment..

ACTIONS A.1 ./7 With one SGT subsystem inoperable, the inoperable subsystem must be restored to OPERABLE status in 7 days. In this condition, the remaining OPERABLE SGT subsystem is adequate to perform the required radioactivity release control function. However, the overall system reliability is reduced because a single failure in the OPERABLE subsystem could result in the radioactivity release control function not being adequately performed. The 7 day Completion Time is based on consideration of such factors as the availability of the OPERABLE redundant SGT System and the low probability of a DBA occurring during this period.

B.1 and 8.2 If the SGT subsystem cannot be restored to OPERABLE status within the required Completion Time in MODE 1, 2, or 3, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and to MODE 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

During movement of irradiated fuel assemblies in the rsecondary containmentrfnq Oa ALTE ONS or during OPDRVs, when Required Action A.1 cannot be completed within the required Completion Time, the OPERABLE SGT subsystem should immediately be placed in operation. This action (continued)

B 3.6.4.3-3 Revision 0 Dresden 2 and 3

SGT System B 3.6.4.3 BASES ACTIONS C.1, C.2.1, C. and .3 (continued) 6 C2D.2ensures that the remaining subsystem is OPERABLE, that no

_failures that could prevent automatic actuation will occur, and that any other failure would be readily detected.

rieCl) An alternative to Required Action C.1 is to immediately

__________fsuspend activities that represent a potential for releasing radioactive material .to the secondary containment, thus placing the plant in a condition that minimizes risk. If applicable, COR LTERATI an .movement of irradiated fuel embli ust immediately-be suspended. Suspension of is- ael/vT th eacti ies must not preclude completion of movement of acomponent o a safe position. Also, if applicable, actions must immediately be initiated to suspend OPDRVs in order to minimize the probability of a vessel draindown and re*en*

subsequent potential for fission product release. Actions must continue until OPDRVs are suspended.

The Required Actions of Condition C have been modified by a Note stating that LCO 3.0.3 is not applicable. If moving irradiated fuel assemblies while in MODE 4 or 5, LCO 3.0.3 would not specify any action. If moving irradiated fuel assemblies while in MODE 1, 2, or 3, the ue movement is independent of reactor operations. Therefore, in either case, inability to suspend movement of4irradiated fuel assemblies would not be a sufficient reason to require a reactor shutdown.

D.1 If both SGTS subsystems are inoperable in MODE 1, 2, or 3, the SGT system may not be capable of supporting the required radioactivity release control function. Therefore, one SGT subsystem must be restored to OPERABLE status within I hour.

The 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Completion Time provides a period of time to correct the problem that is commensurate with the importance of supporting the required radioactivity release control function in MODES 1, 2. and 3. This time period also ensures that the probability of an accident (requiring the SGT System) occurring during periods where the required radioactivity release control function may not be maintained is minimal.

(continued)

Dresden 2 and 3 B 3.6.4.3-4 Revision 0

SGT System B 3.6.4.3 BASES ACTIONS E.1 and E.2 (continued)

If one SGT subsystem cannot be restored to OPERABLE status within the required Completion Time in MODE 1. 2, or 3, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and to MODE 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

F.1,F an 3 n When two SGT subsystems are inoperable, if applicable,O e R L AATlONndndmmovement of rradiated fuel assemblies in A 1secondary containment must im a e y be suspended.

Suspension of h act -iep shall not preclude completion o movement of a component to a safe position. Also, if applicable, action must immediately be initiated to suspend OPDRVs in order to minimize the probability of a vessel -"

draindown and subsequent potential for fission product release. Actions must continue until OPDRVs are suspended.

Required Action F.1 has been modified by a Note stating that LCO 3.0.3 is not applicable. If moving# rradiated fuel assemblies while in MODE 4 or 5, LCO 3.0. no specify any action. If movingkirradiated fuel assemblies while inin MODE 1, 2, or 3. the fue movement is in epen ent o reactor operations. Therefore, in either case, inability to suspend movement ofkirradiated fuel assemblies would not be aa sufficient reason to require a reactor shut own.

(continued)

B 3.6.4.3-5 Revision 0 Dresden 2 and 3

SGT System B 3.6.4.3 BASES (continued)

SURVEILLANCE SR 3.6.4.3.1' REQUI REMENTS Operating (from the control room using the manual initiation switch) each SGT subsystem for > 10 continuous hours ensures that both subsystems are OPERABLE and that all associated controls are functioning properly. It also ensures that blockage, fan or motor failure, or excessive vibration can be detected for corrective action. Operation with the he-aters on (automatic heater cycling, to maintain temperature) for > 10 continuous hours every 31 days eliminates moisture on the adsorbers and HEPA filters. The 31 day Frequency was developed in consideration of the known reliability of fan motors and controls and the redundancy available in the system.

SR 3.6.4.3.2 t1 This SR verifies that the required SGT filte testing is performed in accordance with the Ventilation Filter Testing Program (VFTP). The SGT System filter test are in accordance with Regulatory Guide 1.52 (Ref.). The VFTP includes testing HEPA filter performance, charcoal adsorber efficiency, minimum system flow rate, and the physical properties of the activated charcoal (general use and following specific operations). Specific test frequencies and additional information are discussed in detail in the VFTP.

SR 3.6.4.3.3 This SR verifies that each SGT subsystem starts on receipt of an actual or simulated initiation signal. While this Surveillance can be performed with the reactor at power, operating experience has shown that these components usually pass the Surveillance when performed at the 24 month Frequency. The LOGIC SYSTEM FUNCTIONAL TEST in LCO 3.3.6.2.

"Secondary Containment Isolation Instrumentation," overlaps this SR to provide complete testing of the safety function.

Therefore, the Frequency was found to be acceptable from a reliability standpoint.

(continued)

B 3.6.4.3-6 Revision 0 Dresden 2 and 3

SGT System B 3.6.4.3 BASES (continued)

REFERENCES 1. UFSAR, Section 3.1.2.4.12.

2. UFSAR, Section 6.5.3.2.

3. UFSAR, Section 15.6.5.

Regulatory Guide 1.52. Rev. 2.

Dresden 2 and 3 B 3.6.4.3-7 Revision 0

CREV System CREV System B 3.7.4 BASES BACKGROUND The CREV System is designed to maintain the n ro room (continued) emergency zone environment for a 30 dacontinuous occupancy after a DBA without exceeding 5 rem h~ e b doseei equival t to any *>7rt of thg body). The CREV System will pressurize the control room emergency zone to about 0.125 inches water gauge to minimize infiltration of air from adjacent zones. CREV System operation in maintaining control room habitability is discussed in the-UFSAR, Sections 6.4. 9.4, and 15.6.5 (Refs. 1, 2. and 3.

respectively).

APPLICABLE The ability of the CREV System to maintain the habitability SAFETY ANALYSES of the control room emergency zone is an explicit assumption for the safety analyses presented in the UFSAR, Sections 6.4 and 15.6.5 (Refs. I and 3, respectively). The isolation/

pressurization mode of the CREV System is assumed to_ operate following a DBA loss of coolant accident6fu hand

-*C Xc i Teýmain steam line break, a co ro

• a QL 5. The radiological dose-sT-o control room personnel as a result of the DBA loss of coolant accident are summarized in Reference 3.

The CREV System satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii).

LCO The CREV System is required to be OPERABLE. Total system failure could result in exceeding a dose of 5 rem to the control room operators in the event of a DBA.

The CREV System is considered OPERABLE when the individual components necessary to control operator exposure are OPERABLE. The system is considered OPERABLE when its associated:

a. AFU is OPERABLE,

b. Train B air handling unit (fan portion only) is OPERABLE, including the ductwork, to maintain air circulation to and from the control room emergency zone; and

c. Emergency outside air ventilation intake is OPERABLE.

(continued)

Dresden 2 and 3 B 3.7.4-2 Revision 0

CREV System B 3.7.4 BASES LCO The AFU is considered OPERABLE when a booster fan is (continued) OPERABLE: HEPA filter and charcoal adsorbers are not excessively restricting flow and are capable of performing their filtration functions; and heater, ductwork, valves, and dampers are OPERABLE, and air circulation through the filter train can be maintained.

In add.ition, the control room emergency zone boundary must be maintained, including the integrity of the walls, floors, ceilings, ductwork, and access doors, such that the pressurization limit of SR 3.7.4.4 can be met. However, it is acceptable for access doors to be open for normal control room emergency zone entry and exit and not consider it to be a failure to meet the LCO.

APPLICABILITY In MODES 1, 2, and 3, the CREV System must be OPERABLE to control operator exposure during and following a DBA, since the DBA could lead to a fission product release.

In MODES 4 and 5, the probability and consequences of a DBA are reduced because of the pressure and temperature limitations in these MODES. Therefore, maintaining the CREV System OPERABLE is not required in MODE 4 or 5, except for the following situations under which si nificant radioactive releases can be postulated: /

a. During movement o lirradiated fuel assemblies in the secondary containment; -4* 'd bDur ý' CORE AIRATIONS and*

,. During operations with potential for draining the reactor vessel (OPDRVs).

ACTIONS A.1 With the CREV System inoperable in MODE 1. 2, or 3, the inoperable CREV System must be restored to OPERABLE status within 7 days. The 7 day Completion Time is based on the low probability of a DBA occurring during this time period.

(continued)

Dresden 2 and 3 B 3.7.4-3 Revision 0

CREV System B 3.7.4 BASES ACTIONS B.1 and B.2 (continued)

In MODE 1. 2, or 3. if the inoperable CREV System cannot be restored to OPERABLE status within the associated Completion Time, the unit must be placed in a MODE that minimizes risk.

To achieve this status, the unit must be placed in at least MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in MODE 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times are reasorable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.

C.11( C , an'* '. 3 LCO 3.0.3 is not applicable while in MODE 4 or 5. However, since irradiated fuel movement can occur in MODE 1, 2, or 3.

the Required Actions of Condition C are modified by a Note "indicating that LCO 3.0.3 does not apply. If moving irradiated fuel assemblies while in MODE 1, 2, or 3. the fuel movement is independent of reactor operations.

Entering LCO 3.0.3 while in MODE 1, 2. or 3 would require the unit to be shutdown, but would not require immediate sus ension of movement o irradiated fuel assemblies. The NOTE to the AT ."LCO 3.0.3 is not applicable," ensures that the actions for immediate suspension ofirradiated fuel assem y movement are not postpone u- entry into LCO 3.0.3.

With the CREV System inoperable, during movement of irradiated fuel assemblies in the secondary containment u ng COR ALTERA NS or during OPDRVs, action must te taken immediately to suspend activities that present a potential for releasing radioactivity that might require the CREV System to be placed in the isolation/pressurization mode of operation. This places the unit in a condition that minimizes risk.

If applicable, CALTE IlONS movement of irradiated- '

fuel assemblies in the secondary containment must be suspended immediately. Suspension of he activ ies shall not preclude completion of movement of a component to a safe position. Also, if applicable, action must be initiated (continued)

Dresden 2 and 3 B 3.7.4-4 Revision 0

CREV System B 3.7.4 BASES ACTIONS C.l . an .3ý (continued)

'imm ediately to suspend OPDRVs to minimize the probability of Sýa vessel draindown and the subsequent potential for fission product release. Action must continue until the OPDRVs are suspended.

SURVEILLANCE SR 3.7.4.1 REQUIREMENTS This SR verifies that the CREV System in a standby mode starts from the control room and continues to operate. This SR includes initiating flow through the HEPA filters and charcoal adsorbers. Standby systems should be checked periodically to ensure that they start and function properly. As the environmental and normal operating conditions of this system are not severe, testing the system once every month provides an adequate check on this system.

Monthly heater operation for > 10 continuous hours, during system operation dries out any moisture that has accumulated in the charcoal as a result of humidity in the ambient air.

Furthermore, the 31 day Frequency is based on the known reliability of the equipment.

SR 3.7.4.2 This SR verifies that the required CREV testing is performed in accordance with Specification 5.5.7. "Ventilation Filter Testing Program (VFTP)." The CREV filter tests are in accordance with Regulatory Guide 1.52 (Ref. 4). The VFTP includes testing HEPA filter performance, charcoal adsorber efficiency, system flow rate, and the physical properties of the activated charcoal (general use and following specific operations). Specific test frequencies and additional information are discussed in detail in the VFTP.

SR 3.7.4.3 This SR verifies that on a manual initiation from the control room, the CREV System filter train starts and the isolation dampers close. Operating experience has shown that these components normally pass the SR when performed at the 24 month Frequency. Therefore, the Frequency was found to be acceptable from a reliability standpoint.

(continued)

Dresden 2 and 3 B 3.7.4-5 Revision 0

Control Room Emergency Ventilation AC System B 3.7.5 BASES APPLICABLE generator supported switchgear. Train B Control Room HVAC SAFETY ANALYSES is normally in the standby condition and is used for (continued) accident mitigation. Train A Control Room HVAC is nonsafety related and is in operation during normal conditions. The Train B refrigeration condensing unit, normally served by the Service Water System, can be provided with cooling water from the Unit 2 Containment Cooling Service Water (CCSW)

Systcm. The Control Room Emergency Ventilation AC System is designed in accordance with Seismic Category I requirements.

The Control Room Emergency Ventilation AC System is capable of removing sensible and latent heat loads from the control room emergency zone, including consideration of equipment heat loads and personnel occupancy requirements to ensure equipment OPERABILITY.

The Control Room Emergency Ventilation AC System satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii).

LCO The Control Room Emergency Ventilation AC System is required to be OPERABLE. Total system failure could result in the equipment operating temperature exceeding limits.

The Control Room Emergency Ventilation AC System is considered OPERABLE when the individual components necessary to maintain the control room emergency zone temperature are OPERABLE. These components include the cooling coils, fans, chillers, compressors, ductwork, dampers, and associated instrumentation and controls. In addition, during conditions in MODES other than MODES 1, 2, and 3 when the Control Room Emergency Ventilalion AC System is required to be OPERABLE e.g,/, during COK ALTERATI S , the necessary portions of the Unit 2 CCSW System and Ultimate Heat Sink capable of providing cooling to the refrigeration condensing unit are part of the OPERABILITY requirements covered by this LCO.

APPLICABILITY In MODE 1, 2, or 3, the Control Room Emergency Ventilation AC System must be OPERABLE to ensure that the control room emergency zone temperature will not exceed equipment OPERABILITY limits following control room emergency zone isolation.

(continued)

Quad Cities 1 and 2 B 3.7.5-2 Revision 0

Control Room Emergency Ventilation AC System B 3.7.5 BASES APPLICABILITY In MODES 4 and 5, the probability and consequences of a (continued) Design Basis Accident are reduced due to the pressure and temperature limitations in these MODES. Therefore, maintaining the Control Room Emergency Ventilation AC System OPERABLE is not required in MODE 4 or 5, except for the following situations under which significant radioactive releases can be postulated:

a. During movement ofirradiated fuel assemblies in the secondary containment; d rb. *1Zuring CORF7LTERATION na During operations with a potential for draining the reactor vessel (OPDRVs).

ACTIONS A.1 With the Control Room Emergency Ventilation AC System inoperable in MODE 1. 2, or 3. the system must be restored to OPERABLE status within 30 days. The 30 day Completion Time is based on the low probability of an event occurring requiring control room emergency zone isolation and the availability of alternate nonsafety cooling methods.

B.1 and B.2 In MODE 1. 2, or 3, if the inoperable Control Room Emergency Ventilation AC System cannot be restored to OPERABLE status within the associated Completion Time, the unit must be placed in a MODE that minimizes risk. To achieve this status, the unit must be placed in at least MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in MODE 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.

(continued)

Quad Cities 1 and 2 B 3.7.5-3 Revision 0

Control Room Emergency Ventilation AC System 8 3.7.5 BASES ACTIONS (continued)

LCO 3.0.3 is not applicable while in MODE 4 or 5. However,

,Jnseirradiated fuel movement can occur in MODE 1, 2, or 3,

" the Required Actions of Condition C are modified by a Note indicating that LCO 3.0.3 does not apply. If moving p-irradiated fuel assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations. ,-'

Entering LCO 3.0.3 while in MODE 1, 2, or 3 would require the unit to be shutdown, but would not require immediate suspension of movement of'irradiated fuel assemblies. The Note to the ACTIONS, "LCO 3.0.3 is not applicable," ensures that the actions for immediate suspension ofqirradiated fuel assembly movement are not postponed due to entry into LCO 3.0.3.

With the Control Room Emergency Ventilation AC System inoperable during movement ofwirradiated fuel as.=e.blies in the secondary containmen , ng -0 T S or during OPDRVs, action must De taken immediateTy to suspend activities that present a potential for releasing radioactivity that might require isolation of the control room. This places the unit in a condition that mind.izes , -

risk.

t If applicable, CO LTERATIO anmovement hof t 'irradiate m zes fuel assemblies in the secondary containment m be suspended immediately. Suspension of th acti ies shall not preclude completion of movement of a component to a safe position. Also, if applicable, action must be initiated immediately to suspend OPDRVs to minimize the probability of a vessel draindown and subsequent potential for fission product release. Action must continue until the OPDRVs are suspended.

SURVEILLANCE SR 3.7.5.1 REQUIREMENTS This SR verifies that the heat removal capability of the system is sufficient to remove the control room emergency zone heat load assumed in the safety analyses. The SR consists of a combination of testing and calculation. The 24 month Frequency is appropriate since significant degradation of the Control Room Emergency Ventilation AC System is not expected over this time period.

(continued)

Quad Cities 1 and 2 B 3.7.5-4 Revision 0

Main Condenser Offgas B 3.7.6 B 3.7 PLANT SYSTEMS B 3.7.6 Main Condenser Offgas BASES BACKGROUND During unit operation, steam from the low pressure turbine is exhausted directly into the main condenser. Air and noncondensible gases are collected in the main condenser, then exhausted through the steam jet air ejectors (SJAEs) to the Main Condenser Offgas System. The offgas from the main condenser normally includes radioactive gases.

The Main Condenser Offgas System has been incorporated into the unit design to reduce the gaseous radwaste emission.

This system uses a catalytic recombiner to recombine radiolytically dissociated hydrogen and oxygen. The gaseous mixture is cooled by the offgas condenser; the water and condensibles are stripped out by the offgas condenser and moisture separator. The radioactivity of the remaining gaseous mixture (i.e., the offgas recombiner effluent) is monitored downstream of the moisture separator prior to entering the holdup line.

APPLICABLE The main condenser offgas gross gamma activity rate is an SAFETY ANALYSES initial condition of the Main Condenser Offgas System failure event, discussed in Reference 1. The analysis assumes a gross failure in the Main Condenser Offgas System that results in the rupture of the Main Condenser Offgas System pressure boundary. The gross gamma activity rate is controlled to ensure that, during the event, the calculated offsite doses will be well within the limits of 10 CFR 1 (Ref. 2).

The main condenser offgas limits satisfy Criterion 2 of 10 CFR 50.36(c)(2)(ii).

LCO To ensure compliance with the assumptions of the Main Condenser Offgas System failure event (Ref. 1), the fission product release rate should be consistent with a noble gas release to the reactor coolant of 100 pCi/MWt-second after decay of 30 minutes. The LCO is established consistent with this requirement (2527 MWt x 100 pCi/MWt-second =

252,700 pCi/second).

(continued)

Dresden 2 and 3 B 3.7.6-1 Revision 0

Main Condenser Offgas B 3.7.6 BASES (continued)

SURVEILLANCE SR 3.7.6.1 REOUIREMENTS This SR, on a 31 day Frequency, requires an isotopic analysis of a representative offgas sample (taken at the recombiner outlet or the SJAE outlet if the recombiner is bypassed) to ensure that the required limits are satisfied.

The noble gases to be sampled are Xe-133, Xe-135, Xe-138, Kr-85M, Kr-87, and Kr-88. If the measured rate of radioactivity increases significantly as indicated by the main condenser air ejector noble gas activity monitor (by

> 50% after correcting for expected increases due to changes in THERMAL POWER), an isotopic analysis is also performed within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after the increase is noted, to ensure that the increase is not indicative of a sustained increase in the radioactivity rate. The 31 day Frequency is adequate in view of other instrumentation that continuously monitor the offgas, and is acceptable, based on operating experience.

This SR is modified by a Note indicating that the SR is not required to be performed until 31 days after any main steam line is not isolated and the SJAE is in operation. Only in this condition can radioactive fission gases be in the Main Condenser Offgas System at significant rates.

REFERENCES 1. Letter E-DAS-015-O0 from D.A. Studley (Scientech-NUS) to T. Leffler (ComEd), dated January 24, 2000.

2. 10 CFR Dresden 2 and 3 B 3.7.6-3 Revision 0

Spent Fuel Storage Pool Water Level B 3.7.8 B 3.7 PLANT SYSTEMS B 3.7.8 Spent Fuel Storage Pool Water Level BASES BACKGROUND The minimum water level in the spent fuel storage pool meets the assumptions of iodine decontamination factors following a fuel handling accident.

A general description of the spent fuel storage pool design is found in the UFSAR, Section 9.1.2 (Ref. 1). The assumptions of the fuel handling accident are found in Reference 2.

APPLICABLE The water level above the irradiated fuel assemblies is an SAFETY ANALYSES explicit assumption of the fuel handling accident. A fuel handling

__ radiloia accident is evalua e jto ensure that-the cosqec~acltd1hole body and thyr*i ds* s at the exclu~son area Zlow and *Ipulationn zone

-'e7'_*//eundaries) Zare 25% of 10 CFR l18((Ref. 3)) eexposur S*,,*Z~~guidelie NvG-80(es4*fd 0 CFR 50, 1"- A dendix A, GDC 19 5) and less th Ythe.

imits. (Ref. 6). A fuel handling accidenL could release a Trac ion o the fission product inventory by breaching the fuel r cladding as discussed in Regulato ide (Ref.

The fuel handling accident is evaluated for the dropping of an irradiated fuel assembly onto the reactor core. The water level in the spent fuel storage pool provides for absorption of water soluble fission product gases and transport delays of soluble and insoluble gases that must pass through the water before being released to the secondary containment atmosphere. This absorption and transport delay reduces the potential radioactivity of the release during a fuel handling accident.

The spent fuel storage pool water level satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

LCO The specified water level preserves the assumptions of the fuel handling accident analysis (Ref. 2). As such, it is the minimum required for fuel movement within the spent fuel storage pool.

(continued)

Dresden 2 and 3 B 3.7.8-1 Revision 0

Pool Water Level Spent Fuel Storage Spent Fuel Storage Pool Water Level B 3.7.8 BASES (continued)

REFERENCES 1. UFSAR, Section 9.1.2.

2. Letter E-DAS-00-046 from D.A. Studley (Scientech) to Robert Tsai (ComEd), "Submittal of Calculation in Support of Improved Tech. Spec. Program," dated February 14, 2000.

3. 10.FR

4. NUREG-0800, S ction'15.7.4. vision 1. July 19 NUREG-080 , Section 6.4 R ision 2, July 1 1.

6. 10 CF 50, Appendix A. GDC 19.

Regulatory Guide .25 March 2.

Dresden 2 and 3 B 3.7.8-3 Revision 0

AC Sources-Operating B 3.8.1 BASES SURVEILLANCE SR 3.8.1.21 (continued)

REQUI REMENTS Requirement. Exceptions are noted to the opposite unit SRs of LCO 3.8.1. SR 3.8.1.9 and SR 3.8.1.20 are excepted since only one opposite unit offsite circuit and DG is required by the given unit's Specification. SR 3.8.1.13, SR 3.8.1.18, and SR 3.8.1.19 are excepted since these SRs test the opposite unit's ECCS initiation signal, which is not needed for the AC electrical power sources to be OPERABLE on the given unit.

The Frequency required by the applicable opposite unit SR also governs performance of that SR for the given unit.

As Noted, if the opposite unit is in MODE 4 or 5, or moving irradiated fuel assemblies in the secondary containment, the following opposite unit SRs are not required to be performed: SR 3.8.1.3, SR 3.8.1.10 through SR 3.8.1.12, and SR 3.8.1.14 through SR 3.8.1.17. This ensures that a given unit SR will not require an opposite unit SR to be performed, when the opposite unit Technical Specifications exempts performance of an opposite unit SR (however, as stated in the opposite unit SR 3.8.2.1 Note 1, while performance of an SR is exempted, the SR must still be met).

REFERENCES 1. UFSAR, Section 3.1.2.2.8.

2. UFSAR, Section 8.2.

3. UFSAR, Chapter 6.

4. UFSAR, Chapter 15.

5. Generic Letter 84-15, July 2, 1984.

6. Regulatory Guide 1.93, Revision 0. December 1974.

7. UFSAR, Section 3.1.2.2.9.

8. Regulatory Guide 1.9, Revision 3, July 1993.

9. Regulatory Guide 1.108, Revision 1, August 1977.

(continued)

Dresden 2 and 3 B 3.8.1-33 Revision 0

AC Sources -Shutdown B 3.8.2 B 3.8 ELECTRICAL POWER SYSTEMS B 3.8.2 AC Sources-Shutdown BASES BACKGROUND A description of the AC sources is provided in the Bases for LCO 3.8.1, "AC Sources-Operating."

APPLICABLE The OPERABILITY of the minimum AC sources during MODES 4 SAFETY ANALYSES and 5, and during movement ofkirradiated fuel assemblies in the secondary containment ensures that:

a. The facility can be maintained in the shutdown or refueling condition for extended periods;

b. Sufficient instrumentation and control capability is available for monitoring and maintaining the unit status; and

c. Adequate AC electrical power is provided to mitigate events postulated during shutdown, such as an inadvertent draindown of the vessel or a fuel handling accident.

In general, when the unit is shutdown the Technical Specifications requirements ensure that the unit has the capability to mitigate the consequences of postulated accidents. However, assuming a single failure and concurrent loss of all offsite or loss of all onsite power is not required. The rationale for this is based on the fact that many Design Basis Accidents (DBAs) that are analyzed in MODES 1, 2, and 3 have no specific analyses in MODES 4 and 5. Worst case bounding events are deemed not credible in MODES 4 and 5 because the energy contained within the reactor pressure boundary, reactor coolant temperature and pressure, and corresponding stresses result in the probabilities of occurrences significantly reduced or eliminated, and minimal consequences. These deviations from DBA analysis assumptions and design requirements during shutdown conditions are allowed by the LCO for required systems.

During MODES 1, 2. and 3, various deviations from the analysis assumptions and design requirements are allowed within the ACTIONS. This allowance is in recognition that (continued)

Dresden 2 and 3 B 3.8.2-1 Revision 0

AC Sources-Shutdown B 3.8.2 BASES LCO assuming a loss of the offsite circuit. Together, (continued) OPERABILITY of the required offsite circuit and DG ensures the availability of sufficient AC sources to operate the plant in a safe manner and to mitigate the consequences of postulated events during shutdown (e.g., fuel handling accidents and reactor vessel draindown).

S/The qualifiedcef-fsite circuit(s) must be capable, of aintaining rated frequency and voltage while connected to their respective ESS bus(es), and of accepting required loads during an accident. Qualified offsite circuits are those that are described in the UFSAR and are part of the licensing basis for the unit. The offsite circuit from the 138 kV or 345 kV switchyard consists of the incoming breakers and disconnects to the 22 or 32 reserve auxiliary transformer (RAT), associated 22 or 32 RAT, and the respective circuit path including feeder breakers to 4160 kV ESS buses required by LCO 3.8.8. Another qualified circuit is provided by the bus tie between the corresponding ESS buses of the two units.

The required DG must be capable of starting, accelerating to rated speed and voltage, connecting to its respective 4160 V ESS bus on detection of bus undervoltage, and accepting required loads. This sequence must be accomplished within 13 seconds. Each DG must also be capable of accepting required loads within the assumed loading sequence intervals, and must continue to operate until offsite power can be restored to the 4160 V ESS buses. These capabilities are required to be met from a variety of initial conditions such as DG in standby with engine hot and DG in standby with engine at ambient conditions. Additional DG capabilities must be demonstrated to meet required Surveillances. Proper sequencing of loads, including tripping of nonessential loads, is a required function for DG OPERABILITY. The necessary portions of the DG Cooling Water and Ultimate Heat Sink System capable of providing cooling to the required DG is also required.

It is acceptable for divisions to be cross tied during shutdown conditions, permitting a single offsite power circuit to supply all required divisions.

(continued)

Dresden 2 and 3 B 3.8.2-3 Revision 0

Shutdown AC Sources -

AC Sources -Shutdown B 3.8.2 BASES (continued) 4 and 5 APPLICABILITY The AC sources are required to be OPERABLE in MODES the and during movement o irradiated fuel assemblies in secondary con a rrnht to provide assurance that:

recei a. Systems providing adequate coolant inventory makeup are available for the irradiated fuel assemblies in the core in case of an inadvertent draindown of the reactor vessel;

b. Systems needed to mitigate a fuel handling accident are available;

c. Systems necessary to mitigate the effects of events that can lead to core damage during shutdown are available; and

d. Instrumentation and control capability is available for monitoring and maintaining the unit in a cold shutdown condition or refueling condition.

in AC power requirements for MODES 1, 2, and 3 are covered LCO 3.8.1.

However, ACTIONS LCO 3.0.3 is not applicable while in MODE 4 or 5.

in MODE 1.

since irradiated fuel assembly movement can occur 2, or 3, the ACTIONS have been modified by a Note stating that LCO 3.0.3 is not applicable. If movin irradiated fuel assemblies while in MODE 4 or 5, LCO 3.0.3 would not specify r/ any action. If movin irradiated fuel assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations. Entering LCO 3.0.3 while in MODE 1, 2, or 3 would require the unit to be shutdown, but would not require immediate suspension of movement of irradiated fuel assemblie-s. The Note to the ACTIONS, "LCO 3.0.3 is not applicable," ensures that the actions for immediate

-sension of irradiated fuel assembly movement are not postponed due to entry into LCO 3.0.3.

A.1 not An offsite circuit is considered inoperable if it is If two or more available to one required 4160 V ESS bus.

one division 4160 V ESS buses are required per LCO 3.8.8, (continued)

B 3.8.2-4 Revision 0 Dresden 2 and 3

AC Sources -Shutdown B 3.8.2 BASES ACTIONS A.1 (continued) with offsite power available may be capable of supporting sufficient required features to allow continuation of CORE ALTERATIONS, fuel movement, and operations with a potential for draining the reactor vessel. By the allowance of the option to declare required features inoperable that are not can be,,

rr e / powered from offsite power, appropriate restrictions implemented in accordance with the required feature(s) LCOs' ACTIONS. Required features remaining powered from a qualified offsite circuit, even if that circuit is considered inoperable because it is not powering other required features, are not declared inoperable by this Required Action. For example, if both Division 1 and 2 ESS buses are required OPERABLE by LCO 3.8.8, and only the Division I ESS buses are not capable of being powered from offsite power, then only the required features powered from Division 1 ESS buses are required to be declared inoperable.

A.2.1, A.2.2, A.2.3, A.2.4. 8.1, B.2, B.3, and B.4 With the required offsite circuit not available to all required divisions, the option still exists to declare all required features inoperable per Required Action A.1. Since this option may involve undesired administrative efforts.

the allowance for sufficiently conservative actions is made.

With the required DG inoperable, the minimum required diversity of AC power sources is not available. It is, ALTERATIONS, movement of re e(o'/ therefore, required to suspend in CORE the secondary containment, and "irradiated fuel assemblies activities that could result in inadvertent draining of the reactor vessel.

Suspension of these activities shall not preclude completion of actions to establish a safe conservative condition.

These actions minimize the probability of the occurrence of postulated events. It is further required to immediately initiate action to restore the required AC sources and to continue this action until restoration is accomplished in order to provide the necessary AC power to the plant safety systems.

(continued)

B 3.8.2-5 Revision 0 Dresden 2 and 3

AC Sources- Shutdown B 3.8.2 BASES ACTIONS A.2.1, A.2.2, A.2.3, A.2.4, B.1, B.2, B.3, and B.4 (continued)

The Completion Time of immediately is consistent with the required times for actions requiring prompt attention. The restoration of the required AC electrical power sources should be completed as quickly as possible in order to minimize the time duringwhich the plant safety systems may be without sufficient power.

Pursuant to LCO 3.0.6, the Distribution System ACTIONS would not be entered even if all AC sources to it are inoperable, resulting in de-energization. Therefore, the Required Actions of Condition A have been modified by a Note to indicate that when Condition A is entered with no AC power to any required ESS bus, ACTIONS for LCO 3.8.8 must be immediately entered. This Note allows Condition A to provide requirements for the loss of the offsite circuit whether or not a division is de-energized. LCO 3.8.8 provides the appropriate restrictions for the situation involving a de-energized division.

SURVEILLANCE SR 3.8.2.1 REQUIREMENTS SR 3.8.2.1 requires the SRs from LCO 3.8.1 that are necessary for ensuring the OPERABILITY of the AC sources in other than MODES 1. 2, and 3 to be applicable. SR 3.8.1.9 is not required to be met since only one offsite circuit is required to be OPERABLE. SR 3.8.1.20 is excepted because starting independence is not required with the DG(s) that is not required to be OPERABLE. SR 3.8.1.21 is not required to be met because the opposite unit's DG is not required to be OPERABLE in MODES 4 and 5. and during movement of irradiated fuel assemblies in secondary containment. efer to the corresponding Bases for LCO 3.8.1 for a discussion of each SR.

This SR is modified by two Notes. The reason for Note 1 is to preclude requiring the OPERABLE DG(s) from being paralleled with the offsite power network or otherwise rendered inoperable during the performance of SRs, and to preclude de-energizing a required 4160 V ESS bus or disconnecting a required offsite circuit during performance of SRs. With limited AC sources available, a single event (continued)

Dresden 2 and 3 B 3.8.2-6 Revision 0

DC Sources- Shutdown B 3.8.5 B 3.8 ELECTRICAL POWER SYSTEMS B 3.8.5 DC Sources -Shutdown BASES BACKGROUND A description of the DC sources is provided in the Bases for LCO 3.8.4, "DC Sources-Operating."

APPLICABLE The initial conditions of Design Basis Accident and SAFETY ANALYSES transient analyses in the UFSAR, Chapter 6 (Ref. 1) and Chapter 15 (Ref. 2). assume that Engineered Safety Feature systems are OPERABLE. The DC electrical power system provides normal and emergency DC electrical power for the diesel generators (DGs), emergency auxiliaries, and control and switching during all MODES of operation and during movement of *rradiated fuel assemblies in the secondary containment.

The OPERABILITY of the DC subsystems is consistent with the e initial assumptions of the accident analyses and the requirements for the supported systems' OPERABILITY.

The OPERABILITY of the minimum DC electrical )power sources during MODES 4 and 5 and during movement off irradiated fuel assemblies in the secondary containment ensures that:

a. The facility can be maintained in the shutdown or refueling condition for extended periods;

b. Sufficient instrumentation and control capability is available for monitoring and maintaining the unit status; and

c. Adequate DC electrical power is provided to mitigate events postulated during shutdown, such as an inadvertent draindown of the vessel or a fuel handling accident In general, when the unit is shut down, the Technical Specifications requirements ensure that the unit has the capability to mitigate the consequences of postulated accidents. However, assuming a single failure and concurrent loss of all offsite or all onsite power is not required. The rationale for this is based on the fact that (continued)

B 3.8.5-1 Revision 0 Dresden 2 and 3

DC Sources -Shutdown B 3.8.5 BASES LCO by LCO 3.8.8. "Distribution Systems-Shutdown." This (continued) requirement ensures the availability of sufficient DC electrical power sources to operate the unit in a safe manner and to mitigate the consequences of postulated events

( A-/ e ~ during shutdown (e.g., fuel handling accidentsyand vertent reactor vesse draindown . ne associated alternate 125 VDC electrical power subsystem may be used to satisfy the requirements of the 125, VDC subsystems.

APPLICABILITY The DC electrical power sources required to be OPERABLE in MODES 4 and 5 and during movement of irradiated fuel assemblies in the secondary containment provide assura ce that: r

a. Required features to provide adequate coolant inventory makeup are available for the irradiated fuel assemblies in the core in case of an inadvertent draindown of the reactor vessel;

b. Required features needed to mitigate a fuel handling accident are available:

c. Required features necessary to mitigate the effects of events that can lead to core damage during shutdown are available; and 1*T d. Instrumentation and control capability is available for monitoring and maintaining the unit in a cold shutdown condition or refueling condition.

The DC electrical power requirements for MODES 1, 2. and 3 are covered in LCO 3.8.4.

ACTIONS LCO 3.0.3 is not applicable while in MODE 4 or 5. However, since irradiated fuel assembly movement can occur in MODE 1.

2. or 3, the ACTIONS have been modified by a Note stating that LCO 3.0.3 is not applicable. If movingirradiated fuel assemblies while in MODE 4 or 5, LCO 3.0.3 would not specify an action. If irradiated fuel assemblies while in MODE 1, 2. or 3. the fuel movement is independent of reactor operations. Entering LCO 3.0.3 while in MODE 1, 2, or 3 (continued)

B 3.8.5-3 Revision 0 Dresden 2 and 3

DC Sources -Shutdown B 3.8.5 BASES ACTIONS would require the unit to be shutdown, but would not require (continued) immediate suspension of movement of irradiated fuel

• assemblies. The Note to-1"" ý1ne 'CITS, "LCO 3.0.3 is not applicable," ensures that the actions for immediate sus Sension of irradiated fuel assembly movement are not postponed due to entry into LC0 3.0.3.

rece/(y A.1. A.2.1, A.2.2, A.2.3., and A.2.4 By allowance of the option to declare required features inoperable with associated DC electrical power subsystem(s) inoperable, appropriate restrictions are implemented in accordance with the affected system LCOs' ACTIONS. However.

in many instances, this option may involve undesired administrative efforts. Therefore, the allowance for sufficiently conservative actions is made (i.e., to suspend CORE ALTERATIONS, movement o4 irradiated fuel assemblies in the secondary containmen , and any activities that could result in inadvertent draining of the reactor vessel).

Suspension of these activities shall not preclude completion of actions to establish a safe conservative condition.

These actions minimize the probability of the occurrence of postulated events. It is further required to immediately initiate action to restore the required DC electrical power subsystems and to continue this action until restoration is accomplished in order to provide the necessary DC electrical power to the plant safety systems.

The Completion Time of immediately is consistent with the required times for actions requiring prompt attention. The restoration of the required DC electrical power subsystems should be completed as quickly as possible in order to minimize the time during which the plant safety systems may be without sufficient power.

SURVEILLANCE SR 3.8.5.1 REQUIREMENTS SR 3.8.5.1 requires all Surveillances required by SR 3.8.4.1 through SR 3.8.4.9 to be applicable. Therefore, see the corresponding Bases for LCO 3.8.4 for a discussion of each SR.

(continued)

Dresden 2 and 3 B 3.8.5-4 Revision 0

Distribution Systems -Shutdown B 3.8.8 B 3.8 ELECTRICAL POWER SYSTEMS B 3.8.8 Distribution Systems-Shutdown BASES BACKGROUND A description of the AC and DC electrical power distribution systems is provided in the Bases for LCO 3.8.7, "Distribution Systems -Operating."

APPLICABLE The initial conditions of Design Basis Accident and and SAFETY ANALYSES transient analyses in the UFSAR, Chapter 6 (Ref. 1)

Engineered Safety Feature (ESF)

Chapter 15 (Ref. 2), assume systems are OPERABLE. The AC and DC electrical power distribution systems are designed to provide sufficient ensure capacity, capability, redundancy. and reliability to power to ESF systems so that the availability of necessary and containment design the fuel, Reactor Coolant System, limits are not exceeded.

The OPERABILITY of the AC and DC electrical power distribution system is consistent with the initial assumptions of the accident analyses and the requirements for the supported systems' OPERABILITY.

power The OPERABILITY of the minimum AC and DC electrical during sources and associated power distribution subsystems MODES 4 and 5, and during movement ofirradiated fuel assemblies in the secondary containmenteures th

a. The facility can be maintained in the shutdown or refueling condition for extended periods;

b. Sufficient instrumentation and control capability is available for monitoring and maintaining the unit status; and nf

_rs* S c. Adequate power is provided to mitigate events postulated during shutdown, such as an inadvertent draindown of the vessel or a fuel handling accident.

The AC and DC electrical power distribution systems satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).

(continued)

B 3.8.8-1 Revision 0 Dresden 2 and 3

Distribution Systems- Shutdown B 3.8.8 BASES (continued)

LCO Various combinations of subsystems, equipment, and components are required OPERABLE by other LCOs, depending on the specific plant condition. Implicit in those requirements is the required OPERABILITY of necessary support features. This LCO explicitly requires energization of the portions of the electrical distribution system, including the opposite unit Division 2 electrical distribution subsystem, necessary to support OPERABILITY of Technical Specifications required systems, equipment, and components-both specifically addressed by their own LCO, and implicitly required by the definition of OPERABILITY.

Maintaining these portions of the distribution system energized ensures the availability of sufficient power to operate the plant in a safe manner to mitigate the consequences of postulated events during shutdown (e.g.,

fuel handling accidents and inadver react essel draindown).

APPLICABILITY The AC and DC electrical power distribution subsystems required to be OPERABLE in MODES 4 and 5 and during movement o irradiated fuel assemblies in the secondary containment provide assurance that:

a. Systems to provide adequate coolant inventory makeup are available for the irradiated fuel in the core in case of an inadvertent draindown of the reactor

_ vessel;

b. Systems needed to mitigate a fuel handling accident "are available:

c. Systems necessary to mitigate the effects of events that can lead to core damage during shutdown are available; and

d. Instrumentation and control capability is available for monitoring and maintaining the unit in a cold shutdown condition or refueling condition.

The AC and DC electrical power distribution subsystem requirements for MODES 1, 2, and 3 are covered in LCO 3.8.7.

(continued)

Dresden 2 and 3 B 3.8.8-2 Revision 0

Distribution Systems -Shutdown B 3.8.8 BASES (continued)

ACTIONS LCO 3.0.3 is not applicable while in MODE 4 or 5. However, since irradiated fuel assembly movement can occur in MODE 1, 2, or 3, the ACTIONS have been modified by a Note stating that LCO 3.0.3 is not applicable. If movinUirradiated fuel assemb1ies while in MODE 4-or 5, LCO 3.0.3 would not specify any action. If moving irradiated fuel assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations. Entering.LCO 3.0.3 while in MODE 1,2., or 3 would require the unit to be shutdown, but would not require immediate suspension of movement ofwirradiated fuel assemblies. The Note to the ACTIONS, "LCO 3.0.3 is not sapplicable," ensures that the actions for immediate suspension ofdirradiated fuel assembly movement are not postponed due to entry into LCO 3.0.3.

A.I. A.2.1. A.2.2. A.2.3. A.2.4. and A.2.55 Although redundant required features may require redundant divisions of electrical power distribution subsystems to be OPERABLE, one OPERABLE distribution subsystem division may be capable of supporting sufficient required features to allow continuation of CORE ALTERATIONS, fuel movement, and operations with a potential for draining the reactor vessel.

By allowing the option to declare required features associated with an inoperable distribution subsystem inoperable, appropriate restrictions are implemented in accordance with the affected distribution subsystem LCO's Required Actions. In many instances this option may involve undesired administrative efforts. Therefore, the allowance for sufficiently conservative actions is made, (i.e.. to suspend CORE ALTERATIONS, movement of irradiated fuel assemblies in the secondary containment, and any activities that could result in inadvertent drain' of the reactor vessel).

Suspension of these activities shall not preclude completion of actions to establish a safe conservative condition.

These actions minimize the probability of the occurrence of postulated events. It is further required to immediately initiate action to restore the required AC and DC electrical power distribution subsystems and to continue this action until restoration is accomplished in order to provide the necessary power to the plant safety systems.

(continued)

Dresden 2 and 3 B 3.8.8-3 Revision 0

RPV Water Level-Irradiated Fuel B 3.9.6 B 3.9 REFUELING OPERATIONS B 3.9.6 Reactor Pressure Vessel (RPV) Water Level-Irradiated Fuel BASES BACKGROUND The movement of irradiated fuel assemblies within the RPV requires a minimum water level of 23 ft above the top of the RPV flange. During refueling, this maintains a sufficient water level in the reactor vessel cavity and spent fuel pool. Sufficient water is necessary to retain iodine fission product activity in the water in the event of a fuel handling accident (Refs. 1 and 2). Sufficient iodine activity wou beLderetained to limit offsite doses from the accident to(<5 o19CFR limits, as provided by the guidance of Reference APPLICABLE During movement of irradiated fuel assemblies, the water SAFETY ANALYSES level in the RPV is an initial condition design parameter in analysis of a fuel handling ancident in1).containment A minimum postulatedy Regulato~r Guid Yli3(Ref.

1 water level of 23' ft (R u ory osi ton C.1.*e of Rf '

Sa11nw**_decoamination factor-of 0 (Reg latory Posj it*on*

7 O- *-C.1 0tfRýR. 1)'to be used in the acci ent analysis or Y. i 1~i This-relates to the assumption that9 of the9 (feSth i) released total r iodine fuel from the pellet to cladding assembly rods is retained by the water.

gap of all

/*--*/*-*rfTe uelp 4et to cladding/gap is assum-ea] otan1*

d ln*jte tota Ifuel rod iodin~e/nventory (Ref.l.

Analysis of the fuel handling accident inside containment is described in Reference 2. With a minimum water level of 23 ft and a minimum decay time of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to fuel handling, the analysis and test programs demonstrate that the iodine release due to a postulated fuel handling accident is adequately captured by the water and that offsit doses are maintained within allowable limits (Ref.y. While the worst case assumptions include the dropping of the irradiated fuel assembly being handled onto the reactor core, the possibility exists of the dropped assembly striking the RPV flange and releasing fission products. Therefore, the minimum depth for water coverage to ensure acceptable radiological consequences is specified from the RPV flange. Since the worst case event results in (continued)

Dresden 2 and 3 R 3.9.6-1 Revision 0

RPV Water Level- Irradiated Fuel B 3.9.6 BASES APPLICABLE failed fuel assemblies seated in the core, as well as the SAFETY ANALYSES dropped assembly, dropping an assembly on the RPV flange (continued) will result in reduced releases of fission gases.

RPV water level satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

LCO A minimum water level of 23 ft above the top of the RPV flange is required to ensure that the radiological consequences of a postulated fuel handling accident are within acceptable limits, as provided by the guidance of Reference APPLICABILITY LCO 3.9.6 is applicable when moving irradiated fuel assemblies within the RPV. The LCO minimizes the possibility of a fuel handling accident in containment that is beyond the assumptions of the safety analysis. If irradiated fuel is not present within the RPV, there can be no significant radioactivity release as a result of a postulated fuel handling accident. Requirements for handling of new fuel assemblies or control rods (where water depth to the RPV flange is not of concern) are covered by LCO 3.9.7. "RPV Water Level-New Fuel or Control Rods."

Requirements for fuel handling accidents in the spent fuel storage pool are covered by LCO 3.7.8, "Spent Fuel Storage Pool Water Level."

ACTIONS A.l If the water level is < 23 ft above the top of the RPV flange, all operations involving movement of irradiated fuel assemblies within the RPV shall be suspended immediately to ensure that a fuel handling accident cannot occur. The suspension of irradiated fuel movement shall not preclude completion of movement of a component to a safe position.

(continued)

Dresden 2 and 3 B 3.9.6-2 Revision 0

RPV Water Level-Irradiated Fuel B 3.9.6 BASES (continued)

SURVEILLANCE SR 3.9.6.1 REQUIREMENTS Verification of a minimum water level of 23 ft above the top of the RPV flange ensures that the design basis for the postulated fuel handling accident analysis during refueling operations is met. Water at the required level limits the consequences of damaged fuel rods, which are postulated to rl). - result from a fuel hapdling accident in containment-,,

(Ref. 2).

The Frequency of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is based on engineering judgment and is considered adequate in view of the large volume of water and the normal procedural controls on valve positions, which make significant unplanned level changes unlikely.

REFERENCES I. Regulatory Guide1.2 Marcha23.,92.

2. UFSAR, Section 15.7.3.

1REG-0800 Sect on1 .4.

10C FR Dresden 2 and 3 B 3.9.6-3 Revision 0

I RPV Water Level -New Fuel or Control Rods B 3.9.7 B 3.9 REFUELING OPERATIONS B 3.9.7 Reactor Pressure Vessel (RPV) Water Level -New Fuel or Control Rods BASES BACKGROUND The movement of new fuel assemblies or handling of control rods within the RPV when fuel assemblies seated within the reactor vessel.,are irradiated requires a minimumwater level of 23 ft above the top of irradiated fuel assemblies seated within the RPV. During refueling, this maintains a sufficient water level above the irradiated fuel.

Sufficient water is necessary to retain iodine fission product activity in the water in the event of a fuel handling accident (Refs. I and 2). Sufficient iodine activity would be retained to limit offsite doses from the accident to 125 Io FR( limits, as provided by the guidance of Referencep" S, APPLICABLE During movement of new fuel assem ie or handling of SAFETY ANALYSES control rods over irradiated uel assemblies, the water level in the RPV is an initial condition design parameter in the analysis of a fuel handlin cident in containment postulated by Regulator Guide 1 5 (Ref. 1). A minimU 20o water level of 23 ft Re ator ition C.1 of Ref. I allows n factor of 00 (Reg atory Pos* ion C.l. of R . 1) to be used in the acci ent 0 analysis for iodin . is relates to the assumption that of the to a iodine released from the pellet to cladding ns- gap 4'*-of all thtk-~53fue1 assembly rods is retained by the

• wate. /7e T`0e7pe11*K to cladding ga ils assumed to /-..

Cc_ontaiK1O% of thezt/otal fuel rod io hne inventory (R . 1).

Analysis of the fuel handling accident inside containment is described in Reference 2. With a minimum water level of 23 ft and a minimum decay time of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to fuel handling, the analysis and test programs demonstrate that

~the iodine release due to a postulated fuel handling accident is adequately captured by the water and that offsit doses are maintained within allowable limits (Re ). The related assumptions include the worst case dropping of an irradiated fuel assembly onto the reactor core loaded with irradiated fuel assemblies.

RPV water level satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

(continued)

Dresden 2 and 3 B 3.9.7-1 Revision 0

RPV Water Level -New Fuel or Control Rods B 3.9.7 BASES (continued)

LCO A minimum water level of 23 ft above the top of irradiated fuel assemblies seated within the RPV is required to ensure that the radiological consequences of a postulated fuel handling accident are within Acceptable limits, as provided by the guidance of Reference APPLICABILITY LCO.3..9.7 is applicable when moving,,Ttew fuel assemblies or handling control rods (i.e., movement with other than the normal control rod drive) when irradiated fuel assemblies are seated within the RPV. The LCO minimizes the possibility of a fuel handling accident in containment that is beyond the assumptions of the safety analysis. If irradiated fuel is not present within the RPV, there can be no significant radioactivity release as a result of a postulated fuel handling accident. Requirements for fuel handling accidents in the spent fuel storage pool are covered by LCO 3.7.8. "Spent Fuel Storage Pool Water Level."

Requirements for handling irradiated fuel over the RPV are covered by LCO 3.9.6, "Reactor Pressure Vessel (RPV) Water Level -Irradiated Fuel."

ACTIONS A.1 If the water level is < 23 ft above the top of irradiated fuel assemblies seated within the RPV, all operations involving movement of new fuel assemblies and handling of control rods within the RPV shall be suspended immediately to ensure that a fuel handling accident cannot occur. The suspension of fuel movement and control rod handling shall not preclude completion of movement of a component to a safe position.

SURVEILLANCE SR 3.9.7.1 REQUIREMENTS Verification of a minimum water level of 23 ft above the top of irradiated fuel assemblies seated within the RPV ensures that the design basis for the postulated fuel handling accident analysis during refueling operations is met. Water at the required level limits the consequences of damaged fuel rods, which are postulated to result from a fuel handling accident in containment (Ref. 2).

(continued)

Dresden 2 and 3 B 3.9.7-2 Revision 0

RPV Water Level -New Fuel or Control Rods B 3.9.7 BASES SURVEILLANCE SR 3.9.7.1 (continued)

REOUIREMENTS The Frequency of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is based on engineering judgment and is considered adequate in view of the large volume of water and the normal procedural controls on valve positions.

which make significant unplanned level changes unlikely.

REFERENCES 1. Regulatory Guidef.1."'

2. UFSAR, Section 15.7.3.

3. NUREG-0800 Section 15..*4 Dresden 2 and 3 B 3.9.7-3 Revision 0