Text

Proposed Change to Technical Specifications Regarding Full Scope Adoption of Alternate Source Term
	ML041600619
Person / Time
Site:	Indian Point
Issue date:	06/02/2004
From:	Dacimo F; Entergy Nuclear Northeast
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	NL-04-068
	Download: ML041600619 (89)
	v • d • e

Entergy Nuclear Northeast

__ Indian Point Energy Center 450 Broadway, GSB

- tEntegP RO. Box 249 Buchanan, NY 10511-0249 Tel 914 734 6700 Fred Dacimo Site Vice President Administration June 2, 2004 Indian Point Unit No. 3 Docket No. 50-286 NL-04-068 Document Control Desk U.S. Nuclear Regulatory Commission Mail Stop O-P1-17 Washington, DC 20555-0001

Subject:

Proposed Change to Technical Specifications Regarding Full Scope Adoption of Alternate Source Term

Reference:

1. Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Plants," dated July 1, 2000.

2. NRC Safety Evaluation, "Issuance of Indian Point 3 Amendment 215 for Selective Adoption of Alternate Source Term," dated March 17, 2003.

3. NRC Generic Letter 2003-01, "Control Room Habitability," dated June 12, 2003.

4. Entergy letter NL-03-129, "NRC Generic Letter 2003-01, Control Room Habitability, 60-day Response," dated August 6, 2003.

Dear Sir:

Pursuant to 10CFR50.90, Entergy Nuclear Operations, Inc (Entergy) is submitting a license amendment request for Indian Point 3 (IP3). This application for amendment proposes a full scope adoption of the alternate source term (AST) methodology for design basis accident dose consequence evaluations in accordance with 10 CFR 50.67. The analysis methodology used is based on the guidance provided in Reference 1.

NRC previously approved (Reference 2) a selective adoption of the AST methodology at IP3 for the fuel handling accident (FHA). The FHA analysis is being revised, as described in this application, to incorporate updated assumptions regarding control room ventilation parameters and to bound a planned increase in rated thermal power for IP3. Details regarding analysis assumptions, methods, and results for the affected design basis accidents, including the revised FHA analysis, are provided in Attachment 1II.

~AIDE

NL-04-068 Docket No. 50-286 Page 2of 2 Entergy has evaluated the proposed license amendment in accordance with 10 CFR 50.91 (a)(1) using the criteria of 10 CFR 50.92 (c) and Entergy has determined that this proposed change involves no significant hazards considerations (Attachment I). Proposed changes to the Technical Specifications, summarized below, are provided in Attachment II:

Revise definition of Dose Equivalent I-131 to reflect the new dose methodology.

Revise the Ventilation Filter Testing Program (Specification 5.5.10) to remove filter testing requirements (high efficiency particulate and charcoal adsorbers) for the Fuel Storage Building Emergency Ventilation System and the Containment Purge System.

Additional Technical Specification changes supported by the adoption of AST will be submitted at a future time.

In accordance with 10CFR50.91, a copy of this applicabion and the associated attachments are being submitted to the designated New York State official.

Entergy requests approval of this license amendment request by March 2005 with a 30-day implementation period. Approval of this amendment request is required to support Entergy's request (submitted separately) for a proposed increase in rated thermal power for Indian Point 3.

This amendment request also supports preparation of a final response to Generic Letter 2003-01 (Reference 3) as stated in the 60-day response letter (Reference 4) for IP3.

There are no new commitments identified in this submittal. If you have any questions or require additional information, please contact Mr. Kevin Kingsley at (914) 734-6695.

I declare under penalty of perjury that the foregoing is true and correct. Executed on ff LQq FrdR. Dacirno Site Vice President Indian Point Energy Center cc:

Mr. Patrick D. Milano, Sr Project Manager Resident Inspector's Office Project Directorate I Indian Point Unit 3 Division of Licensing Project Management U.S. Nuclear Regulatory Commission U.S. Nuclear Regulatory Commission P.O. Box 337 Mail Stop 0 8 C2 Buchanan, NY 10511-0337 Washington, DC 20555-0001 Mr. Peter R. Smith Mr. Hubert J. Miller President, NYSERDA Regional Administrator, Region 1 17 Columbia Circle U.S. Nuclear Regulatory Commission Albany, NY 12203 475 Allendale Road King of Prussia, PA 19406-1415 Mr. Paul Eddy New York State Dept. of Public Service 3 Empire State Plaza Albany, NY 12223-6399

ATTACHMENT I TO NL-04-068 ANALYSIS OF PROPOSED LICENSE AMENDMENT REQUEST AND RELATED TECHNICAL SPECIFICATIONS CHANGES FOR ADOPTION OF ALTERNATE SOURCE TERM METHODOLOGY ENTERGY NUCLEAR OPERATIONS, INC.

INDIAN POINT NUCLEAR GENERATING UNIT NO. 3 DOCKET NO. 50-286

Attachment I to NL-04-068 Docket 50-286 Page 1 of 9

1. DESCRIPTION This letter is a request to amend Operating License DPR-64, Docket No. 50-286 for Indian Point Nuclear Generating Unit No. 3.

This amendment seeks to revise several pages of Sections 1.1 and 5.5 pertaining to the implementation of alternate source term (AST) methodology in accordance with 10CFR50.67. This submittal applies to all design basis accidents associated with dose, as identified in Section 4 below.

Upon issuance of this proposed amendment, the source term analysis for Indian Point Unit 3 will be complete and consistent with a revised Rated Thermal Power (RTP) of 3216 MW and bounding for the current RTP of 3067.4 MW.

2. PROPOSED CHANGE

a. The Indian Point 3 Technical Specifications Definition of DOSE EQUIVALENT 1-131 currently says:

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

The Definition of DOSE EQUIVALENT 1-131 is revised to state:

"DOSE EQUIVALENT I-131 shall be that amount of I -131 (curies) that alone would produce the same committed effective dose equivalent (CEDE) dose as the quantity and isotopic mixture of I -130, I -131, I -132,1 -133, I -134 and I -135 actually present. The CEDE dose conversion factors used for this calculation shall be those listed in Table 2.1 of EPA Federal Guidance Report No. 11,

'Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion and Ingestion,' 1988."

b. Indian Point 3 Technical Specifications Section 5.5.10 "Ventilation Filter Testing Program (VFTP)" shall delete the phrases "Fuel Storage Building Emergency Ventilation System" and "and Containment Purge System" from the first paragraph. Secondly, the reference to the Fuel Storage Building Emergency Ventilation System in Item (2) shall be removed. Finally, Item (3) ("every 18 months for the Containment Purge System") shall be deleted entirely.

c. Indian Point 3 Technical Specifications Section 5.5.10, Table a, includes removal efficiencies for HEPA filters in four ventilation systems. The removal efficiency requirements for (1) the Fuel Storage Building Emergency Ventilation System

Attachment I to NL-04-068 Docket 50-286 Page 2 of 9 and (2)the Containment Purge System HEPA filters are to be removed from this Technical Specification.

d. Indian Point 3 Technical Specifications Section 5.5.10, Table b, includes removal efficiencies for charcoal adsorbers in four ventilation systems. The removal efficiency requirements for (1) the Fuel Storage Building Emergency Ventilation System and (2) the Containment Purge System charcoal adsorbers are to be removed from this Technical Specification.

e. Indian Point 3 Technical Specifications Section 5.5.10, Table c, includes methyl iodide removal efficiencies for charcoal adsorbers in four ventilation systems.

The methyl iodide removal efficiency requirements for (1) the Fuel Storage Building Emergency Ventilation System and (2) the Containment Purge System charcoal adsorbers are to be removed from this Technical Specification.

f. Indian Point 3 Technical Specifications Section 5.5.10, Table d, includes pressure drop testing for filtration trains in three systems. This requirement for the Fuel Storage Building Emergency Ventilation System is being removed from the Technical Specifications.

3. BACKGROUND This License Amendment Request is the second submittal to incorporate the methodology of the Alternate Source Term (AST), as defined in Regulatory Guide 1.183.

On June 5, 2002, Entergy requested implementation of the AST methodology as applied exclusively to the fuel handling analysis (Reference 1). The SER for this analysis was issued in March 2003 as IP3 Technical Specifications Amendment 215.

The Indian Point 3 licensing basis for all remaining dose-related design basis accidents in Chapter 14 of the FSAR (Reference 6) are currently based on the methodologies and assumptions derived from TID-14844 (Reference 2). 10CFR50.67 allows licensees to revise the current source term used in radiological analyses. Regulatory Guide 1.183 (Reference 4) provides methods and assumptions that may be used in adopting an alternate source term for use in evaluating the radiological consequences of various hypothetical accident scenarios. This License Amendment Request (LAR) proposes to adopt an alternate source term in accordance with 10CFR50.67, using the guidance of Regulatory Guide 1.183. Affected sections of the FSAR will be revised in accordance with 10CFR50.71 to reflect the new analyses' assumptions, methods and results as compared to the regulatory acceptance criteria.

Implementation of the alternate source term methodology supports the following proposed changes to the Technical Specifications:

Revised Definition of Dose Equivalent Iodine Dose Equivalent Iodine (DEI) is a parameter that is derived from a combination of the measured activities of the five isotopes I -131,1 -132,1 -133,1 -134 and I -135. It is used routinely throughout the Technical Specifications and is currently based upon the standard definition as established in TID-14844 for thyroid dose. The revised definition, which is standard for Alternate Source Term methodology, is a combination of activities

Attachment I to NL-04-068 Docket 50-286 Page 3 of 9 from the six isotopes I -130,1 -131,1 -132,1 -133,1 -134 and I -135. The revised DEI definition also varies from the TID-14484 in that it is based on committed effective dose equivalent (CEDE) dose instead of thyroid dose.

The limits for DEI, as established in Technical Specification 3.4.16 (RCS Specific Activity), are unchanged under the new definition.

Elimination of Filter Testing for the Containment Purge System Section 5.5 of the Technical Specifications includes testing requirements for the charcoal and HEPA filters in the Containment Purge System. As noted in Attachment l1l, no credit is taken for the dose-mitigating effects of either charcoal or HEPA filters in this system, for any of the radiological accidents occurring in Containment. Therefore the filtration testing requirements are no longer necessary.

Elimination of Filter Testing Requirements for the Fuel Storage Building Emergency Ventilation System Section 5.5 of the Technical Specifications includes testing requirements for the charcoal and HEPA filters in the Fuel Storage Building Emergency Ventilation System. As noted in Attachment Ill, no credit is taken for the dose-mitigating effects of either charcoal or HEPA filters in this system. The FHA analysis presumes fuel assembly drop in Containment without Containment Integrity established and with no credit for filtration.

This bounds the FHA in the Fuel Storage Building. Therefore the filtration testing requirements are no longer necessary.

4. TECHNICAL ANALYSIS This submittal seeks to implement the AST methodology, as allowed by Regulatory Guide 1.183 (Reference 4) and 10CFR50.67 (Reference 5) for the remainder of design basis accidents associated with radiation dose. These are:

Large-Break LOCA (LB-LOCA)

Small-Break LOCA (SB-LOCA)

Steam Generator Tube Rupture (SGTR)

Locked Rotor

Rod Ejection

Main Steam Line Break (MSLB)

Gas Decay Tank Rupture

Volume Control Tank Rupture

Holdup Tank Rupture In addition, the Fuel Handling Accident, as originally evaluated in Reference 1, has been revised for consistency with the other nine design basis accidents, as noted in Item 8 below.

The assumptions supporting these calculations are equivalent to those currently in the 1P3 design basis, with the following exceptions:

Attachment I to NL-04-068 Docket 50-286 Page 4 of 9

1. The rated thermal power (RTP) assumed for all accidents is 3216 MWt, plus 2% uncertainty. This bounds the currently licensed RTP of 3067.4 MWt, plus 0.6% uncertainty, and supports a planned power uprate for IP3.

2. No credit is taken for charcoal or HEPA filtration capability in the Containment Purge System. This is consistent with assumptions in the previous AST analysis for the fuel handling accident. Accordingly, this submittal seeks the removal of filtration requirements from the IP3 Technical Specifications' Ventilation Filter Testing Program (VFTP) in Section 5.5.

3. No credit is taken for charcoal filtration capability in the Containment Building Fan Cooler Units. Furthermore, except for SB-LOCA and Rod Ejecton analyses, no credit is taken for HEPA filtration. However, in order to ensure consistency with equipment Environmental Qualifications (EQ) requirements and post-accident accessibility calculations, the test requirements for FCU HEPA and charcoal are retained.

The analysis considers possible future retirement of the Spray Additive Tank (SAT) and, where appropriate, provides a separate set of dose calculations assuming that pH is controlled via passive trisodium phosphate baskets rather than the SAT. These alternate calculations, which credit neither HEPA nor charcoal filtration in the FCUs, are provided for information only. Entergy does not at this time seek retirement of the SAT.

4. Two separate Control Room Heating, Ventilation and Air Conditioning (CR-HVAC) System configurations are analyzed for dose results in all accident scenarios. One configuration assumes a filtered recirculation flow rate of 1000 cfm and a filtered makeup flow rate of 400 cfm. The other assumes no filtered recirculation and a filtered makeup flow rate of 1500 cfm (i.e., all air to the CR is from outside the building). The dose reported for each accident scenario is the higher of the doses resultant from these two configurations.

This approach was used to establish the bounding dose, since it may not be apparent which CR-HVAC configuration is limiting due to the dynamic nature of AST analyses. Entergy will establish the preferred configuration for CR-HVAC operation based on tracer gas test results (see item 5) and the analysis results provided in this amendment request.

5. The unfiltered inleakage flow rate into the CR is assumed to be < 700 cfm.

The prior analysis for the fuel handing accident used a conservatively high value of 1800 cfm. Tracer Gas Testing Is currently scheduled for August 2004 to confirm the acceptability of this revised assumption.

6. The minimum hot leg switchover time is being reduced from 14 hours1.62037e-4 days 0.00389 hours 2.314815e-5 weeks 5.327e-6 months to 6.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months , in order to minimize the risk of boron precipitation in the reactor coolant. The 6.5-hour assumption is conservative with respect to the current LOCA analysis based on a 14-hour assumption and is consistent with a new LOCA analysis that supports a proposed power increase for IP3.

7. A penalty has been applied to Containment Spray System (CSS) flow rate to provide margin for future results from CSS pump testing.

Attachment I to NL-04-068 Docket 50-286 Page 5 of 9

8. The Fuel Handling Accident (FHA) analysis, which was approved and issued as License Amendment 215, has been modified for consistency with the remainder of the AST accidents, regarding control room ventilation systems flow rates.

These are the significant assumptions applicable to the AST analysis. Details of assumptions that apply to the individual accident scenarios are included in the report (Reference 3), which is included as Attachment II.

The calculated dose results, reported in units of Total Effective Dose Equivalent (TEDE) for all accidents, are summarized on the table below:

Control Room Site Boundary Low Population Accident Report Dose Dose Zone Dose Scenario Section (versus Limit) (versus Limit) (versus Limit)

LB-LOCA 2.0 4.4 rem 23.4 rem 11.2 rem (5 rm (25 rem) (25 rem)

SB-LOCA 7.0 2.2 rem 11.0 rem 5.5 rem (5 rem) (25 rem) (25 rem)

SGTR (with 3.0 2.2 rem 4.9 rem 1.9 rem pre-accident iodine spike) (5 rem) (25 rem) (25 rem)

SGTR (with 3.0 0.9 rem 1.9 rem 0.8 rem accident-initiated iodine spike) (2.5 rem) (2.5 rem)

Locked Rotor 4.0 2.5 rem 1.1 rem 1.4 rem (5 rem) (2.5 rem) (2.5 rem)

Rod Ejection 5.0 0.9 rem 4.4 rem 2.2 rem (5 rem) (6.3 rem) (6.3 rem)

MSLB (with 6.0 0.6 rem 0.2 rem 0.3 rem pre-accident iodine spike) (5 rem) (25 rem) (25 rem)

MSLB (with 6.0 2.1 rem 0.5 rem 0.8 rem accident-initiated iodine spike) (5 rem) (2.5 rem) (2.5 rem)

GDT Rupture 8.0 0.1rem 0.4 rem 0.2 rem (5 rem) (0.5 rem) (0.5 rem)

VCT Rupture 9.0 0.08 rem 0.42 rem 0.16 rem (5 rem) (0.5 rem) (0.5 rem)

HU Tank 10.0 0.1 rem 0.38 rem 0.14 rem Rupture (5 rem) (0.5 rem) (0.5 rem)

FHA 11.0 1.4 rem 5.7 rem 2.1 rem (5 rem) (6.3 rem) (6.3 rem)

The table demonstrates that dose limits are met for all accidents.

Attachment I to NL-04-068 Docket 50-286 Page 6 of 9

5. REGULATORY ANALYSIS 5.1 No Significant Hazards Consideration Entergy Nuclear Operations, Inc, has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing upon the three standards set forth in 10CFR50.92, 'Issuance of Amendment," as discussed below:

1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No The proposed change involves the reanalysis of design basis radiological accidents in Containment and the Fuel Storage Building. The new analyses, based on the Alternate Source Term (AST), in accordance with 10CFR50.67, will replace the existing analyses that are based on the methodologies of TID-14844. As a result of the new analyses, changes to the Technical Specifications are proposed which take credit for the new analysis results.

The proposed changes to the Technical Specifications modify requirements regarding filter testing for a variety of systems (i.e., Containment Purge, Fuel Storage Building Emergency Ventilation). The analyses do not credit charcoal or HEPA filtration for dose mitigation. The proposed changes reflect the plant configuration that will support implementation of the AST analyses.

The AST analysis follows the guidance of the NRC Regulatory Guide 1.183 and uses the acceptance criteria of the NRC Standard Review Plan (NUREG-0800) for offsite doses and General Design Criteria for Control Room personnel. The accident analyses conservatively assume that the Containment Building and the Fuel Storage Building, including ventilation filtration systems for those buildings, do not diminish or delay the assumed fission product release.

The proposed changes also revise the definition of Dose Equivalent Iodine (DEI) to be consistent with the assumptions of the analyses. The limits for DEI do not change as a result of the implementation of the AST analyses.

The change from the original source term to the new proposed AST is a change in analysis method and assumptions and has no effect on accident initiators or causal factors that contribute to the probability of occurrence of previously analyzed accidents.

Use of AST to analyze the dose effect of design basis accidents shows that regulatory acceptance criteria for the new methodology continue to be met. Changing the analysis methodology does not change the sequence or progression of the accident scenario.

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

Attachment I to NL-04-068 Docket 50-286 Page 7 of 9

2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The changes proposed in this license amendment request involve the use of a new analysis methodology and related regulatory acceptance criteria. In addition, certain changes to plant ventilation systems can be made based on the analysis results, using the new methodology. Use of a new analysis method does not impact the design or operation of plant systems or components and new accident scenarios would therefore not be created. The proposed changes to air ventilation and filtration systems do not adversely affect plant equipment used to protect plant safety limits or the way in which that plant equipment is operated or maintained. As a result, no new failure modes are being introduced that could lead to different accidents.

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

3. Does the proposed change involve a significant reduction in a margin of safety?

Response: No.

The existing dose analysis methodology and assumptions demonstrate that the dose consequences for all design basis accidents are within regulatory limits for whole body and thyroid doses as established in I OCFR1 00 (except for the Fuel Handling Analysis, which is already based on the AST methodology). The alternate dose analysis methodology and assumptions also demonstrate that the dose consequences of these accidents are within the regulatory requirements established for the new methodology.

The limits applicable to the alternate analysis are established in IOCFR50.67 in conjunction with the Total Effective Dose Equivalent (TEDE) acceptance directed in Regulatory Guide 1.183. The acceptance criteria for both dose analysis methods have been developed for the purpose of evaluating design basis accidents to demonstrate adequate protection of public health and safety. An acceptable margin of safety is inherent in both types of acceptance criteria.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, Entergy Nuclear Operations, Inc, concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 10CFR50.92 (c), and, accordingly, a finding of uno significant hazards" is justified.

5.2 Applicable Regulatory Requirements I Criteria The proposed changes have been evaluated to determine compliance with applicable regulatory requirements.

The revised analyses for the design basis accidents identified in Section 4 are based on 10CFR50.67 and use the regulatory guidance of Regulatory Guide 1.183 and Standard

Attachment I to NL-04-068 Docket 50-286 Page 8 of 9 Review Plan (NUREG 0800) Section 15.0.1. The analyses demonstrate compliance with these regulatory guides and criteria. Use of the new analysis method replaces 10CFR100 as the applicable dose acceptance criteria for all design basis accidents (the AST methodology was previously implemented for the Fuel Handling Accident in Technical Specifications Amendment 215).

GDC 19 requires that holders of an operating license using an alternative source term under IOCFR50.67 shall meet the requirements of the criterion by ensuring the radiation exposures to Control Room occupants shall not exceed 5 rem TEDE. The analysis provided to support the requested changes demonstrates that this requirement is met.

Entergy has determined that the proposed changes do not require any exemptions or relief from regulatory requirements. The proposed use of an alternate source term to evaluate the consequences of a design basis accident results in a change to the existing licensing basis analysis described in the FSAR. In accordance with 10CFR50.71, Entergy will update the FSAR to reflect the proposed new analysis method. The changes to the Technical Specifications incorporate assumptions used in the new analysis.

5.3 Environmental Considerations The proposed amendment does not involve (i) a significant hazards consideration, or (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10CFR52.22(c)(9). Therefore, pursuant to IOCFR51.22(b), no environmental assessment need be prepared in connection with the proposed amendment.

6. PRECEDENCE The NRC has approved similar applications regarding full scope adoption of the alternate source term for Surry Units 1 and 2 (Dockets 50-280 and 281), approved March 8, 2002) and Kewaunee (Docket 50-305), approved March 17, 2003.

7. REFERENCES

1) Letter IPN-02-044, Entergy to USNRC, "Proposed Changes to Technical Specifications: Selective Adoption of Altemate Source Term and Incorporation of Generic Changes: TSTF-51, TSTF-68 and TSTF-312," June 5, 2002

2) J. J. DiNunnio et al, "Calculation of Distance Factors for Power and Test Reactor Sites," USAEC TID-14844, U.S. Atomic Energy Commission, 1962.

3) Westinghouse Report, "Licensing Report for the Radiological Consequences of Accidents Using Alternative Source Term Methodology (Regulatory Guide 1.183) for the Indian Point 3 Nuclear Power Plant," dated May 24, 2004.

Attachment I to NL-04-068 Docket 50-286 Page 9 of 9

4) Regulatory Guide 1.183, "AItemative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," July 2000

5) NRC Final Rule 10CFR50.67, issued in Federal RegisterVol 64, No. 246, pages 71990-72002, December 23, 1999

6) Indian Point 3 Final Safety Analysis Report Chapter 14 ("Safety Analysis")

ATTACHMENT 11 TO NL-04-068 MARKUP OF TECHNICAL SPECIFICATION PAGES FOR PROPOSED ADOPTION OF ALTERNATE SOURCE TERM METHODOLOGY Page 1.1-3 Definitions Page 5.0-21 Ventilation Filter Testing Program Page 5.0-22 Ventilation Filter Testing Program (continued)

Page 5.0-23 Ventilation Filter Testing Program (continued)

Page 5.0-24 Ventilation Filter Testing Program (continued)

Page 5.0-25 Ventilation Filter Testing Program (continued)

[NOTE: There are no Bases associated with these Technical Specification pages]

ENTERGY NUCLEAR OPERATIONS, INC.

INDIAN POINT NUCLEAR GENERATING UNIT NO. 3 DOCKET NO. 50-286

Definitions 1.1 1.1 Definitions (continued)

DOSE EQUIVALENT 1-131 DOSE EQUIVALENT I-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, I-132, I-133, 1-134, and I-135 actually present.

The thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID-14844, AEC, 1962, "Calculation of Distance Factors DELETE, replace for Power and Test Reactor Sites," or those listed in with Insert A, below Table E-7 of Regulatory Guide 1.109, Rev.1, NRC, 1977, or ICRP 30, Supplement to Part 1, page 192-212, Table titled, "Committed Dose Equivalent in Target Organs or Tissues per Intake of Unit Activity".

E - AVERAGE E shall be the average (weighted in proportion to the DISINTEGRATION ENERGY concentration of each radionuclide in the reactor coolant at the time of sampling) of the sum of the average beta and gamma energies per disintegration (inMeV) for isotopes, other than iodines, with half lives > 10 minutes, making up at least 95% of the total noniodine activity in the coolant.

La The maximum allowable primary containment leakage rate, L,, shall be 0.1% of primary containment air weight per day at the calculated peak containment pressure (Pa).

LEAKAGE LEAKAGE shall be:

a. Identified LEAKAGE

1. LEAKAGE, such as that from pump seals or valve packing (except for leakage into closed systems and reactor coolant pump (RCP) seal water injection or leakoff), that is captured and conducted to collection systems or a sump or collecting tank; (Leakage into closed systems is leakage that can be accounted for and contained by a INSERT A DOSE EQUIVALENT I-131 shall be that amount of 1-131 (curies) that alone would produce the same committed effective dose equivalent (CEDE) dose as the quantity and isotopic mixture of 1-130, I-131, I-132, I-133, I-134 and I-135 actually present. The CEDE dose conversion factors used for this calculation shall be those listed in Table 2.1 of EPA Federal Guidance Report No. 11,

'Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion and Ingestion,' 1988.

(continued)

INDIAN POINT 3 1.1 - 3 Amendment 205

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.10 Ventilation Filter Testing Proaram (VFTP)

This program provides controls for implementation of required testing of the ventilation filter function for the Fuel Storagc Building Emergency Ventilation System, Control Room Ventilation SystemT and Containment Fan Cooler Units, and Cntainmnet Purge Sytc m.

Applicable tests described in Specifications 5.5.10.a, 5.5.10.b, 5.5.10.c and 5.5.10.d shall be performed:

1) After 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of charcoal adsorber use since the last test; and,

2) Every 24 months for the Fuel Storage Building Emergency Ventilation System, Control Room Ventilation System, and Containment Fan Cooler Units; and,

3) Every 18 menths for the Containment Purge System; and,

4) 3 After each complete or partial replacement of the HEPA filter train or charcoal adsorber filter; and, i) 4 After any structural maintenance on the system housing that could alter system integrity; and,

6) 5 After significant painting, fire, or chemical release in any ventilation zone communicating with the system while it is in operation.

SR 3.0.2 is applicable to the Ventilation Filter Testing Program.

(continued)

INDIAN POINT 3 5.0-21 Amendment 205

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.10 Ventilation Filter Testing Proaram (VFTP) (continued)

a. Demonstrate for each system that an inplace test of the high efficiency particulate air (HEPA) filters shows the specified penetration and system bypass leakage when tested in accordance with the referenced standard at the flowrate specified below.

Ventilation Removal Flowrate System Efficiency (cfm) Reference Standard lIDELETE I

4. 4 Fuel Storage Building Emergency 80% to 120%

Ventilation of design System accident Regulatory Guide 1.52, Rev 2,

_ ^.

80% to 120%

Control Room of design Ventilation accident Regulatory Guide 1.52, Rev 2, system 2 99% rate Sections C.5.a and C.5.c 80% to 120%

of design Containment Fan accident Regulatory Guide 1.52, Rev 2, Cooler Units 2 99% rate Sections C.5.a and C.5.c 90% to 110%

Containment of design Purge System operating rat- Regulatory Guide 1.52, Rev 2, 2 99% Sections C.5.a and C.S.c (continued)

INDIAN POINT 3 5.0 - 22 Amendment 20 5

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.10 Ventilation Filter Testing Program (VFTP) (continued)

b. Demonstrate for each system that an inplace test of the charcoal adsorber shows the specified penetration and system bypass leakage when tested in accordance with the referenced standard at the flowrate specified below.

Ventilation Removal system Efficiency Flowrato (cfm) Reference Standard I D ELETE I

_ __ _ _ _ _ _III_ I.

Fuel Storage Building Emergency 80% to 120% of Ventilation design Regulatory Guide 1.52, Rev 2, System 2 99% accident rate Sections C.5.a and C.5.d 80% to 120% of Control Room design Ventilation accident rate Regulatory Guide 1.52, Rev 2, System k 99% Sections C.5.a and C.5.d

4 4.

80% to 120% of Containment Fan design Regulatory Guide 1.52, Rev 2, lIDELETE I Cooler Units 2 99% accident rate Sections C.5.a and C.5.d

- I I 90% to 110% of Containment design Regulatory Guide 1.52, Rev 2, Purge System 2 99% operating rate Sections C.5.a and C.5.d (continued)

INDIAN POINT 3 5.0 - 23 Amendment 205

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.10 Ventilation Filter Testing Proaram (VFTP) (continued)

c. Demonstrate for each system that a laboratory test of a sample of the charcoal adsorber shows the methyl iodide removal efficiency specified below when tested in accordance with ASTM D3803-1989, subject to clarification below, at a temperature of 86F and a relative humidity of 95%.

Methyl iodide ASTM D3803-1989 Ventilation System removal Clarification efficiency I DELETE I (%):

Fuel Storage Building 2 90 59 ft/min face velocity Emergency Ventilation System Control Room Ventilation 2 95.5 78 ft/min face velocity System Containment Fan Cooler 2 85 59 ft/min face velocity lDELETE l Units Unt Containment Purge System 2 90 31 ft/min face velocity

+ I Note: For the 1" beds, the Control Room Ventilation System methyl iodide removal efficiency is verified greater than or equal to 93%

rather than 95.5% at a face velocity of 50 ft/min under the above requirements. This is done prior to fuel movement in Refuel Outage 12 and every 6 months after Refuel Outage 12 until the end of Refuel Outage 13 or the 2" beds are installed.

(continued)

INDIAN POINT 3 5.0 - 24 Amendment 219

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.10 Ventilation Filter Testing Program (VFTP) (continued)

d. Demonstrate for each system that the pressure drop across the combined HEPA filters, the demisters and prefilters (if installed),

and the charcoal adsorbers is less than the value specified below when tested at the flowrate specified below.

Ventilation System Delta P

_ __ (inches wg) Flowrate (cfm)

I DELETE I I 4 Fuel Storage Building Emergency Ventilation System a 90% of design accident 6 rate l 1 Control Room Ventilation System i 90% of design accident 6 rate Containment Fan Cooler units k 90% of design accident 6 rate (continued)

INDIAN POINT 3 5.0 - 25 Amendment 20 5

ATTACHMENT III TO NL-04-068 LICENSING REPORT FOR THE RADIOLOGICAL CONSEQUENCES OF ACCIDENTS USING ALTERNATE SOURCE TERM METHODOLOGY FOR THE INDIAN POINT 3 NUCLEAR POWER PLANT ENTERGY NUCLEAR OPERATIONS, INC.

INDIAN POINT NUCLEAR GENERATING UNIT NO. 3 DOCKET NO. 50-286

Licensing Report for the Radiological Consequences of Accidents Using Alternative Source Term Methodology (Regulatory Guide 1.183) for the Indian Point Unit 3 Nuclear Power Plant May 24, 2004

Table of Contents 1.0 Radiological Consequences Utilizing Alternative Source Terms ........................................ 3 2.0 Large Break Loss of Coolant Accident Radiological Analysis ........................................... 7 3.0 Steam Generator Tube Rupture Radiological Analysis ............................................. 13 4.0 Locked Rotor Radiological Analysis ............................................. 16 5.0 Rod Ejection Radiological Analysis ............................................. 18 6.0 Steam Line Break Radiological Analysis ............................................. 22 7.0 Small Break LOCA Radiological Analysis ............................................. 25 8.0 Gas Decay Tank Rupture Radiological Analyses ............................................. 28 9.0 Volume Control Tank Rupture ............................................. 30 10.0 Holdup Tank Failure ............................................. 32 11.0 Fuel Handling Accident ............................................. 34 12.0 Conclusions ............................................. 37 13.0 References ............................................. 38 Table 1: Committed Effective Dose Equivalent Dose Conversion Factors .39 Table 2: Effective Dose Equivalent Dose Conversion Factors .40 Table 3: Offsite Breathing Rates and Atmospheric Dispersion Factors .41 Table 4: Control Room Parameters .42 Table 5: Core Total Fission Product Activities .43 Table 6: RCS Coolant Concentrations .44 Table 7: Nuclide Decay Constants .45 Table 8: Iodine Chemical Species .46 Table 9: Fission Product Release Timing ........................................................ 46 Table 10: Core Fission Product Release Fractions .47 Table 11: Nuclide Groups .48 Table 12: Assumptions Used for Large Break LOCA Dose Analysis .49 Table 13: Steam Generator Tube Rupture Thermal Hydraulic Results .51 Table 14: Assumptions Used for SGTR Dose Analysis .52 Table 15: Iodine Specific Activities .54 Table 16: Iodine Spike Appearance Rates .54 Table 17: Assumptions Used for Locked Rotor Dose Analysis .55 Table 18: Assumptions Used for Rod Ejection Dose Analysis .57 Table 19: Assumptions Used for Steam Line Break Dose Analysis .60 Table 20: Assumptions Used for Small Break LOCA Dose Analysis .62 Table 21: Assumptions Used for Gas Decay Tank Rupture Dose Analysis .65 Table 22: Assumptions Used for Volume Control Tank Rupture Dose Analysis .66 Table 23: Assumptions Used for Holdup Tank Failure Dose Analysis .67 Table 24: Assumptions Used for FHA Analysis .68 Table 25: Core Source Term 84 Hours after Shutdown .69 Page 2 of 69

1.0 Radiological Consequences Utilizing Alternative Source Terms 1.1 Introduction The Indian Point Unit 3 Nuclear Power Plant licensing basis for the radiological consequences analyses for Chapter 14 of the FSAR is currently based on methodologies and assumptions that are derived from TID-14844 (Reference 1) and other early guidance.

Regulatory Guide (RG) 1.183 (Reference 2) provides guidance on application of alternative source terms (AST) in revising the accident source terms used in design basis radiological consequences analyses, as allowed by 10CFR50.67 (Reference 3). This includes modeling consistent with NUREG- 1465, "Accident Source Terms for Light-Water Nuclear Power Plants" (Reference 4). The alternative source term methodology as established in RG 1.183 (based on NUREG-1465) is being used to calculate the offsite and control room radiological consequences for Indian Point 3 to support the control room habitability program. The following FSAR chapter 14 radiological consequences analyses are analyzed: LBLOCA, SBLOCA, steam generator tube rupture (SGTR),

locked rotor, rod ejection, main steamline break (MSLB), gas decay tank (GDT) rupture, volume control tank (VCT) rupture, holdup tank rupture, and fuel handling accident (FHA). Each accident and the specific input assumptions are described in detail in subsequent sections in this report.

The FHA was analyzed previously using the AST methodology and an SER was issued in March 2003. The FHA radiological consequences analysis has been revised to reflect a more conservative source term model and to incorporate changes in the control room HVAC flow rates and allowable inleakage.

1.2 Common Analysis Inputs and Assumptions The assumptions and inputs described in this section are common to analyses discussed in this report. The accident specific inputs and assumptions are discussed in Sections 2 through 10.

The total effective dose equivalent (TEDE) doses are determined at the site boundary (SB) for the limiting two-hour interval, at the low population zone (LPZ) and to control room personnel (CR) for the duration of the event. The interval for determining control room doses may extend beyond the time when the releases are terminated. This accounts for the additional dose to the operators in the control room, which will continue for as long as the activity is circulating within the control room envelope.

The TEDE dose is equivalent to the committed effective dose equivalent (CEDE) from inhalation and the deep dose equivalent (DDE) from external exposure. Effective dose equivalent (EDE) is used in lieu of DDE in determining the contribution of external dose to the TEDE consistent with RG 1.183 guidance. The dose conversion factors (DCFs) used in determining the CEDE dose are from Reference 5 and are given in Table 1. The dose conversion factors used in determining the EDE dose are from Reference 7 and are listed in Table 2.

The offsite breathing rates and the offsite atmospheric dispersion factors used in the offsite radiological calculations are provided in Table 3.

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Parameters modeled in the control room personnel dose calculations are provided in Table 4. These parameters include the normal operation flowrates, the emergency operation flowrates, control room volume, filter efficiencies and control room operator breathing rates. In the analyses presented in this report, the control room is modeled as a discrete volume. The atmospheric dispersion factors calculated for release of activity from the release point to the control room intake are used to determine the activity available at the intake. The inflow (filtered and unfiltered) to the control room and the control room recirculation flow are used to calculate the activity introduced to the control room and cleanup of activity from that flow.

The core inventory for the Indian Point 3 alternative source term analysis was determined using fuel management data on fuel regions, burnups and enrichments. The ORIGEN2 computer code (Reference 6) was used to generate the core inventory. Consideration was given to normal fuel management variations from cycle-to-cycle. The core fission product activity is provided in Table 5 for all nuclides. Reactor coolant system and volume control tank radiation sources were determined using ORIGEN2 as a basis for core inventory. The build-up of fission product activities in the reactor coolant system, volume control tank liquid and vapor phases is calculated. Values selected are the maximums that occur during the fuel cycle from startup through the equilibrium cycle.

The core and reactor coolant system activities in Tables 5 and 6 are based on a core power of 3216 MWt, plus 2.0% uncertainty. The activities in Tables 5 and 6 include an additional 4% increase to address fuel management variations. Decay constants for each nuclide are calculated using the half-lives of Reference 8 and are provided in Table 7.

1.3 Dose Calculation Models 1.3.1 Offsite Dose Calculation Models The TEDE dose is calculated for the worst 2-hour period at the EAB. At the LPZ the TEDE dose is calculated up to the time all releases are terminated. The TEDE doses are obtained by combining the CEDE doses and the EDE doses.

Offsite inhalation doses (CEDE) are calculated using the following equation.

DCEDE DCF (IAR )j (BR )j (X /Q )J]

where:

DCEDE = CEDE dose via inhalation (rem)

DCFj = CEDE dose conversion factor via inhalation for isotope i (rem/Ci) (Table 1)

(LAR)\j = Integrated activity of isotope i released during the time interval j (Ci)

(BR)j = Breathing rate during time interval j (m3 /sec) (Table 3)

(X/Q)j = Atmospheric dispersion factor during time interval j (sec/M3 ) (Table 3)

Page 4 of 69

Offsite external exposure (EDE) doses are calculated using the following equation:

=

DE E D i (AR) 1 ( /Q* )]

where:

DEDE = External exposure dose via cloud submersion (rem)

DCFj = EDE dose conversion factor via external exposure for isotope i (rem-m3 /Ci-sec) (Table 2)

(IAR)1 j = Integrated activity of isotope i released during the time interval j (Ci)

(x/Q)j = Atmospheric dispersion factor during time interval j (sec/rn3) (Table 3) 1.3.2 Control Room Dose Calculation Models CEDE (doses due to inhalation) and EDE (doses due to external exposure) are calculated for 30 days in the control room.

The control room is modeled as a discrete volume. The atmospheric dispersion factors calculated for the transfer of activity to the control room intake are used to determine the activity available at the control room intake. The inflow (filtered and unfiltered) to the control room and the filtered recirculation flow are used to calculate the concentration of activity in the control room. Control room parameters used in the analyses are presented in Table 4. Control room atmospheric dispersion factors used in each analysis are provided in the input assumption table for that accident.

Control room inhalation doses are calculated using the following equation:

DCEDE = E[DCFj (xConcij * (BR)j * (OF)j1 where:

DCEDE = CEDE dose via inhalation (rem)

DCFj = CEDE dose conversion factor via inhalation for isotope i (rem/Ci) (Table 1)

Concij = concentration in the control room of isotope i, during time interval j, calculated dependent upon inleakage, filtered inflow, filtered recirculation, total outflow and CR volume (Ci-sec/m3)

(BR)j = breathing rate during time interval j (m3 /sec) (Table 4)

(OF)j = occupancy factor during time interval j (Table 4)

Control room external exposure doses due to activity in the control room volume are calculated using the following equation:

Page 5 of 69

DDE =(-)*EDCF EConc j *(OF)J where:

DEDE = external exposure dose via cloud submersion (rem)

GF = geometry factor, calculated based on Reference 11, using the equation:

GF =- 173 , where V is the control room volume in ft3 DCFi = EDE dose conversion factor via external exposure for isotope i (rem.m 3 /Ci sec)

(Table 2)

Concij = concentration in the control room of isotope i, during time interval j, calculated dependent upon inleakage, filtered inflow, filtered recirculation, total outflow and CR volume (Ci-sec/m 3 )

(OF)Q = occupancy factor during time interval j (Table 4) 1.3.3 Control Room Direct Dose from External Sources Direct dose to the control room operators from activity in containment was calculated using a source based on ORIGEN2 with the timing of releases specified by RG 1.183..

The source in the containment was projected to the control room, through control room shielding in place, by point kernel methods.

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2.0 Large Break Loss of Coolant Accident Radiological Analysis An abrupt failure of the main reactor coolant pipe is assumed to occur and it is assumed that the emergency core cooling features fail to prevent the core from experiencing significant degradation (i.e., melting). This sequence cannot occur unless there are multiple failures, and thus goes beyond the typical design basis accident that considers a single active failure. Activity from the core is released to the containment and from there released to the environment by means of containment leakage and leakage from the emergency core cooling system.

2.1 Comparison of RG 1.183 Source Term Methodology to TID-14844 The reanalysis of the LBLOCA offsite and control room doses for Indian Point 3 uses the following RG 1.183 source term characteristics in place of those identified in TID-14844 and Regulatory Guide 1.4 (Reference 12):

Iodine chemical species

Fission product release timing

Fission product release phases through early in-vessel

Fission product release fractions

Fission product groups A comparison of RG 1.183 to TID- 14844 is provided in Tables 8 through 10.

2.2 Input Parameters and Assumptions The input parameters and assumptions are listed in Table 12. Activity is released from the fuel into the containment using the timing and release fractions from Tables 9 and 10.

The analysis considers the release of activity from the containment via containment leakage. Leakage of sump solution through the reactor coolant pump (RCP) leak-off line is also considered. In addition, once the external recirculation mode of the emergency core cooling system (ECCS) is established, activity in the sump solution may be released to the environment by means of leakage from ECCS equipment into the auxiliary building. The total offsite and control room doses are the sum of the doses resulting from each of the postulated release paths.

The following sections address topics of significant interest.

2.2.1 Source Term The reactor coolant activity is assumed to be released to the containment atmosphere over the first 30 seconds of the accident. However, the activity in the coolant is insignificant compared with the release from the core and is not included in the analysis.

The use of RG 1.183 source term modeling results in several major departures from the assumptions used in the existing LOCA dose analysis as reported in the FSAR.

Instead of assuming instantaneous melting of the core and release of activity to the containment, the release of activity from the core occurs over a 1.8-hour interval.

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The gap release phase occurs in the first half hour and the release from the melted fuel occurs over the next 1.3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months .

Instead of considering only the release of iodines and noble gases, a wide spectrum of nuclides is taken into consideration. Table 11 lists the nuclides being considered for the LOCA with core melt (eight groups of nuclides). Tables 9 and 10 provide a comparison between the fission product release fractions and the timing/duration of releases to the containment as assumed in TID-14844 and in RG 1.183.

Instead of the iodine being primarily in the elemental form, the iodine is mainly in the form of cesium iodide which exists as particulate and the fraction that is in the organic form is much smaller. The iodine characterization from RG 1.183 is compared in Table 8 with that from Regulatory Guide 1.4.

The other groups of nuclides (other than the iodines and the noble gases) all occur as particulates only.

For the containment leakage analysis, all activity released from the fuel is assumed to be in the containment atmosphere until removed by sprays, sedimentation, radioactive decay or leakage from the containment. For the leakage of sump solution outside of containment (ECCS leakage and RCP seal leak-off), all iodine activity released from the fuel is assumed to be in the sump solution until removed by radioactive decay or leakage.

2.2.2 Containment Modeling The containment building is modeled as two discrete volumes: sprayed and unsprayed.

The containment volume is 2.61E6 ft3 with a sprayed fraction of 80 percent of the total.

The volumes are conservatively assumed to be mixed only by the containment fan coolers. Removal of activity by the fan cooler filters is conservatively neglected in the analysis. The analysis therefore supports removal of the fan cooler filters.

The containment is assumed to leak at the design leak rate of 0.1% per day for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of the accident and then to leak at half that rate (0.05% per day) for the remainder of the 30-day period following the accident considered in the analysis.

2.2.3 Removal of Activity from the Containment Atmosphere No credit is taken for the removal of elemental and organic iodine by the charcoal filters on the fan cooler units nor is credit taken for removal of aerosols by the HEPA filters on the fan cooler units. The removal of elemental iodine from the containment atmosphere is assumed to be accomplished only by containment sprays and radioactive decay. The removal of particulates from the containment atmosphere is assumed to be accomplished by containment sprays, sedimentation and radioactive decay. The noble gases and the organic iodine are assumed to be subject to removal only by radioactive decay.

One train of the containment spray system is assumed to operate in the injection mode following the LOCA. When the RWST drains to a predetermined level, the operators switch to recirculation of the sump liquid to provide a source to the sprays. There is an 67-second delay for injection spray flow initiation. The minimum injection spray duration until the level is reached is approximately 44 minutes. The switchover from Page 8 of 69

injection to recirculation is assumed to take 3 minutes. During this 3 minutes, the analysis does not credit any spray removal in the containment. The analysis assumed that the recirculation sprays operate until 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after the accident.

2.2.3.1 Containment Spray Removal of Elemental Iodine The current Standard Review Plan (Reference 9) identifies a methodology for the determination of spray removal of elemental iodine independent of the use of spray additive. The removal rate constant is determined by:

As = 6KgTF / VD Where: Kg = Gas phase mass transfer coefficient, ft/min T = Time of fall of the spray drops, min F = Volume flow rate of sprays, ft3 /hr V = Containment sprayed volume, ft3 D = Mass-mean diameter of the spray drops, ft Parameters for Indian Point 3 in the spray injection phase are listed below:

Kg = 9.84 ft/min T= 10.0 sec F = 2200 gpm V = 2.088E6 ft3 D=0.112 cm These parameters and the appropriate conversion factors were used to calculate the elemental spray removal coefficients. The upper limit of 20 hr-' specified for this model is applied in the analysis in place of the calculated value of 22.7 he'.

The elemental iodine removal rate during recirculation spray operation can be calculated by multiplying the injection spray removal rate (22.7 hr ) by the ratio of the recirculation spray flow rate (1050 gpm) to the injection spray flow rate (2200 gpm). The recirculation spray removal rate is then 10.8 hr '. However, during recirculation, the spray solution will gradually become loaded with elemental iodine which will limit the capacity of the spray to remove airborne iodine. As the decontamination factor (DF) approaches its defined limit, the removal coefficient would be only a small fraction of its original value.

This is approximated by setting the removal coefficient at one half of the calculated value (5.4 hr-'). This was rounded down to 5.0 hi-'.

Removal of elemental iodine from the containment atmosphere is assumed to be terminated when the airborne inventory drops to 0.5 percent of the total elemental iodine released to the containment (this is a DF of 200). With the RG 1.183 source term methodology, this is interpreted as being 0.5 percent of the total inventory of elemental iodine that is released to the containment atmosphere over the duration of gap and in-vessel release phases. In the analysis, this occurs at 2.765 hours0.00885 days 0.213 hours 0.00126 weeks 2.910825e-4 months .

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2.2.3.2 Containment Spray Removal of Particulates Particulate spray removal is determined using the model described in Reference 9. The first order spray removal rate constant for particulates may be written as follows:

= 3hFE / 2Vd Where: h = Drop Fall Height F = Spray Flow Rate V = Volume Sprayed E = Single Drop Collection Efficiency d = Drop Diameter Parameters for Indian Point 3 in the spray injection phase are listed below:

h = 118.5 ft F = 2200 gpm V = 2.088E6 ft3 The E/d term depends upon the particle size distribution and spray drop size. From Reference 10, it is conservative to use 10 m' for E/d until the point is reached when the inventory in the atmosphere is reduced to 2% of its original (DF of 50). With the RG 1.183 source term methodology, this is interpreted as being 2% of the total inventory of particulate iodine that is released to the containment atmosphere over the duration of gap and in-vessel release phases.

These parameters and the appropriate conversion factors were used to calculate the particulate spray removal coefficients. A value of 4.6 hr-' was used in the analysis during the spray injection phase. The recirculation spray particulate removal rate is 2.2 hr' which corresponds to the spray flow rate during the recirculation phase (1050 gpm). The DF of 50 is reached at 3.445 hours0.00515 days 0.124 hours 7.357804e-4 weeks 1.693225e-4 months . Recirculation spray credit continues at a reduced rate (0.22 hr 1) until 4.0 hours0 days 0 hours 0 weeks 0 months . At this time, the recirculation sprays are terminated.

2.2.3.3 Sedimentation Removal of Particulates During spray operation, credit is taken for sedimentation removal of particulates in the unsprayed region. Before sprays are initiated, after sprays are terminated and during the 3-minute switchover from injection to recirculation when sprays are not credited, credit for sedimentation is taken in both the sprayed and unsprayed regions.

Based on the containment systems experiments (CSE) which examined the air cleanup experienced through natural transport processes, it was found that a large fraction of the aerosols were deposited on the floor rather than on the walls indicating that sedimentation was the dominant removal process for the test (Reference 10). The CSE tests determined that there was a significant sedimentation removal rate even with a relatively low aerosol concentration. From Reference 10, even at an air concentration of 10 pg/M3 , the sedimentation removal coefficient was above 0.3 h'. With approximately 2.0 percent of particulates remaining airborne at the end of credited spray removal, there would be more than 10,000 pg/M3 and an even higher sedimentation rate would be expected. For the analysis, the sedimentation removal coefficient is conservatively assumed to be only 0.1 Ih-. This value for sedimentation removal of particulates has been accepted by the NRC for Indian Point Unit 2, Shearon Harris and Kewaunee in their safety evaluation Page 10 of 69

reports for the application of the alternative source term methodology. It is also conservatively assumed that sedimentation removal does not continue beyond a DF of 1000.

2.2.4 Leakage of Sump Solution Outside of Containment In accordance with RG 1.183, it is assumed that the iodine is instantaneously and homogeneously mixed in the primary containment sump water at the time of release from the core.

2.2.4.1 ECCS Leakage When ECCS external recirculation is established following the LOCA, leakage is assumed to occur from ECCS equipment outside containment. The leakage goes into the auxiliary building and no filtration or holdup is credited for this release. Initially, the ECCS recirculation is internal to the containment and there is no potential for leakage outside containment. However, the switch to external recirculation occurs at 6.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months because of the need to switch from cold leg recirculation mode to hot leg recirculation mode. The ECCS leakage is modeled as 4.0 gallon/hr, which is doubled from the plant allowable leakage value of 2.0 gallon/hr consistent with RG 1.183. The leakage continues for the 30-day period following the accident considered in the analysis. The airborne fraction is modeled as 2.7% based on calculations taking into account solution pH, temperature of the leaked water, room volumes (where leakage occurs), and ventilation flows.

2.2.4.2 Reactor Coolant Pump Seal Leak-off Line During the first 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months of the LOCA event, leakage from the reactor coolant pump seal leak-off line is assumed at the rate of 1.0 gallon/hr which is doubled from the plant allowable leakage value of 0.50 gallon/hr consistent with RG 1.183. The airborne fraction is modeled as 10% consistent with RG 1.183.

2.2.5 Control Room Isolation In the event of a large break LOCA, the low pressurizer pressure SI setpoint will be reached shortly after event initiation. The SI signal causes the control room heating, ventilation and air conditioning (HVAC) to switch from the normal operation mode to the accident mode of operation. It is assumed that the SI setpoint is reached immediately at the start of the event and a conservative 60-second delay time for switching from normal to accident operating mode (recirculation with filtered fresh air intake) is modeled.

2.3 Acceptance Criteria The offsite dose limit for a LOCA is 25 rem TEDE per RG 1.183. This is the guideline value of IOCFR50.67. The limit for the control room dose is 5.0 rem TEDE per 10CFR50.67.

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2.4 Results and Conclusions The calculated doses for the large break LOCA are:

Site Boundary 23.4 rem TEDE Low Population Zone 11.2 rem TEDE Control Room 4.4 rem TEDE The control room dose from external sources (activity remaining in containment and activity outside the control room envelope) was determined to be negligible.

The acceptance criteria are met.

The site boundary dose reported is for the worst two-hour period, determined to be from 0.6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to 2.6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

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3.0 Steam Generator Tube Rupture Radiological Analysis The steam generator tube rupture (SGTR) event is separated into two analyses, a thermal and hydraulic analysis and a radiological consequences analysis. The thermal hydraulic analysis is not impacted by the alternative source term methodology, so the results from the licensing basis analysis, summarized below, are used as input to the dose analysis.

3.1 Steam Generator Tube Rupture Thermal and Hydraulic Analysis Results The major hazard associated with an SGTR event is the radiological consequences resulting from the transfer of radioactive reactor coolant to the secondary side of the ruptured steam generator (SG) and subsequent release of radioactivity to the atmosphere.

The primary thermal-hydraulic parameters which affect the calculation of offsite doses for an SGTR include the amount of reactor coolant transferred to the secondary side of the ruptured steam generator, the amount of primary to secondary break flow that flashes to steam and the amount of steam released from the ruptured steam generator to the atmosphere.

It was assumed that the break flow is terminated by operator action at 30 minutes to isolate the ruptured SG. This does not constitute a requirement that the operators demonstrate the ability to terminate break flow within 30 minutes from the start of the event and it is recognized that the operators may not be able to terminate break flow within 30 minutes for all postulated SGTR events. The purpose of the calculation is to provide conservatively high mass-transfer rates for use in the radiological consequences analysis. This was achieved by assuming a constant break flow at the equilibrium flow rate, with a constant flashing fraction that does not credit the plant cooldown, for a relatively long time period. Thirty minutes was selected for this purpose. This modeling is consistent with the SGTR analysis currently presented in Section 14.2.4 of the Updated Final Safety Analysis Report (UFSAR).

As stated in UFSAR Section 14.2.4.4, Westinghouse has performed an SGTR analysis with the operator response time increased from 30 minutes to 60 minutes. The analysis also took into account the effect of charging flow and considered the potential for SG overfill. The analysis demonstrates that the results for the 60-minute release period with a more precisely modeled mass-release analysis are bounded by the 30-minute duration constant break flow model. Thus, the analysis justifies extending the allowable time from 30 to 60 minutes for operator response in the affected Emergency Operating Procedures while the dose analysis remains conservatively based on the 30-minute release model.

Based on a primary and secondary side mass and energy balance, the break flow and atmospheric steam releases from the ruptured and intact steam generators were calculated for 30 minutes. After 30 minutes, it was assumed that steam is released only from the intact steam generators to dissipate the core decay heat and to subsequently cool the plant down to the Residual Heat Removal System (RHRS) operating conditions. It was assumed that the RHRS is capable of removing core decay heat within 29 hours3.356481e-4 days 0.00806 hours 4.794974e-5 weeks 1.10345e-5 months after the SGTR initiation, and that steam releases are terminated at that time. A primary and secondary side mass and energy (M&E) balance was used to calculate the steam release for the intact steam generators from 0 to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , from 2 to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , and from 8 to 29 hours3.356481e-4 days 0.00806 hours 4.794974e-5 weeks 1.10345e-5 months .

Page 13 of 69

The limiting tube rupture break flow, break flow flashing fraction and ruptured steam generator atmospheric steam releases from 0 to 30 minutes are provided in Table 13 along with the long-term intact steam generators steam releases for use in radiological consequences analysis. The values in Table 13 include an approximate 10% increase in mass flow rates for use in the conservative radiological analysis. Increasing the mass transfer data before performing the radiological consequences analysis allows future plant changes that result in small increases in the mass transfer rates to be evaluated without requiring the radiological analysis to be redone.

3.2 Steam Generator Tube Rupture Radiological Analysis For the SGTR, the complete severance of a single steam generator tube is assumed to occur. Due to the pressure differential between the primary and secondary systems, radioactive reactor coolant is discharged from the primary into the secondary system.

A portion of this radioactivity is released to the outside atmosphere through the main condenser, the atmospheric dump valves (ADVs) or the safety valves (MSSVs). In addition, iodine activity is contained in the secondary coolant prior to the accident and some of this activity is released to the atmosphere as a result of steaming from the SGs following the accident.

3.2.1 Input Parameters and Assumptions A summary of input parameters and assumptions is provided in Table 14.

The analysis of the SGTR radiological consequences uses the analytical methods and assumptions outlined in RG 1.183. For the pre-accident iodine spike case, it is assumed that a reactor transient has occurred prior to the SGTR and has raised the RCS iodine concentration to the technical specification limit for a transient of 60 ACi/gm of dose equivalent (DE) I-131. For the accident-initiated iodine spike case, the reactor trip associated with the SGTR creates an iodine spike in the RCS which increases the iodine release rate from the fuel to the RCS to a value 335 times greater than the release rate corresponding to a maximum equilibrium RCS concentration of 1.0 ACi/gm of DE 1-13 1.

The duration of the accident-initiated iodine spike is limited by the amount of activity available in the fuel-clad gap of fuel pins containing cladding defects. Based on having 8 percent of the iodine in the fuel-clad gap, the gap inventory would be depleted within 4.0 hours0 days 0 hours 0 weeks 0 months and the spike is terminated at that time.

The noble gas activity concentration in the RCS at the time the accident occurs is based on a one percent fuel defect level. This is approximately equal to the technical specification value of 100/E bar j+/-Ci/gm for gross radioactivity. The iodine activity concentration of the secondary coolant at the time the SGTR occurs is assumed to be equivalent to the technical specification limit of 0.1 LCi/gm of DE 1-131.

Aside from the tube rupture break flow the amount of primary to secondary SG tube leakage is assumed to be 0.3 gpm per steam generator.

An iodine partition factor in the SGs of 0.01 (curies iodine/gm steam) / (curies iodine/gm water) is used. Prior to reactor trip and concurrent loss of offsite power, an iodine removal factor of 0.01 is taken for steam released to the condenser.

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All noble gas activity carried over to the secondary side through SG tube leakage is assumed to be immediately released to the outside atmosphere.

Break flow flashing fractions and steam release rates from the intact and ruptured steam generators were calculated. The amount of break flow that flashes to steam is conservatively calculated assuming that all break flow is from the hot leg side of the break and that the primary temperatures remain constant. Activity contained in the flashed break flow is released from the steam generators without partitioning.

At 29 hours3.356481e-4 days 0.00806 hours 4.794974e-5 weeks 1.10345e-5 months after the accident, the RHR System is assumed to be placed into service for heat removal and there is no further steam release from the secondary system.

3.2.2 Control Room Isolation The low pressurizer pressure SI setpoint will be reached at -6.5 minutes (392 seconds) from event initiation. The SI signal causes the control room HVAC to switch from the normal operation mode to the accident mode of operation. It is conservatively assumed that the control room HVAC does not fully enter the accident mode of operation until 7.5 minutes after event initiation.

3.3 Acceptance Criteria The offsite dose limit for a SGTR with a pre-accident iodine spike is 25 rem TEDE per RG 1.183. This is the guideline value of IOCFR50.67. For a SGTR with an accident-initiated iodine spike, the offsite dose limit is 2.5 rem TEDE per RG 1.183.

This is 10% of the guideline value of 10CFR50.67. The limit for the control room dose is 5.0 rem TEDE per 10CFR50.67.

3.4 Results and Conclusions The SGTR accident doses are listed below.

For the pre-accident iodine spike:

Site Boundary 4.9 rem TEDE Low Population Zone 1.9 rem TEDE Control Room 2.2 rem TEDE For the accident-initiated iodine spike:

Site Boundary 1.9 rem TEDE Low Population Zone 0.8 rem TEDE Control Room 0.9 rem TEDE The acceptance criteria are met.

The site boundary doses for both the pre-accident and accident initiated iodine spikes reported are for the worst two-hour period, determined to be from 0.0 to 2.0 hours0 days 0 hours 0 weeks 0 months .

The control room doses from activity outside the control room envelope was determined to be negligible.

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4.0 Locked Rotor Radiological Analysis An instantaneous seizure of a reactor coolant pump rotor is assumed to occur which rapidly reduces flow through the affected reactor coolant loop. Fuel clad damage may be predicted to occur as a result of this accident. Due to the pressure differential between the primary and secondary systems and assumed steam generator tube leakage, fission products are discharged from the primary into the secondary system. A portion of this radioactivity is released to the outside atmosphere through either the atmospheric relief valves or safety valves. In addition, iodine activity is contained in the secondary coolant before the accident and some of this activity is released to atmosphere as a result of steaming from the steam generators following the accident.

4.1 Input Parameters and Assumptions A summary of input parameters and assumptions is provided in Table 17.

The analysis of the locked rotor radiological consequences uses the analytical methods and assumptions outlined in RG 1.183.

The current analysis of record shows that there are no rods in DNB for the locked rotor event. However, it is conservatively assumed that 5% of the fuel rods in the core suffer damage as a result of the locked rotor sufficient that all of their gap activity is released to the reactor coolant system. Eight percent of the total I-131 core activity is in the fuel-cladding gap. Ten percent of the total Kr-85 core activity is in the fuel-cladding gap.

Five percent of other iodine isotopes and other noble gases and 12 percent of the total core activity of alkali metals are assumed to be in the fuel-cladding gap. The fuel clad gap activity fractions are from RG 1.183 (Reference 2). The maximum power peaking factor of 1.7 is applied to the calculation of the gap activities, consistent with RG 1.183.

The iodine activity concentration of the primary coolant at the time the locked rotor occurs is assumed to be equivalent to the technical specification limit (ITS 3.4.16.2) of 1.0 gCi/gm of DE I-131 and is given in Table 15. The initial concentration of noble gases and alkali metals in the reactor coolant is given in Table 6 and is based on one percent defective fuel, which corresponds to the technical specification limit (ITS 3.4.16.1) of 100/E-bar.

The iodine activity concentration of the secondary coolant at the time the locked rotor occurs is assumed to be equivalent to the technical specification limit (ITS 3.7.17) of 0.1 tCi/gm of DE I-131 and is given in Table 15. The alkali metal activity concentration of the secondary coolant at the time the locked rotor occurs is assumed to be 10 percent of the primary side alkali metal concentration.

The amount of primary to secondary SG tube leakage is assumed to be equal to the technical specification limit of 1.0 gpm total (i.e., 360 gpd per SG - ITS 3.4.13). The density for this leakage is 62.4 lb.ft3 .

An iodine partition factor in the SGs of 0.01 (curies iodine/gm steam) / (curies iodine/gm water) is used.

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The retention of particulates in the SGs is limited by the moisture carryover. The maximum moisture carryover is 0.1%. Therefore, an alkali metal partition factor in the SGs of 0.001 (curies iodine/gm steam) / (curies iodine/gm water) is used.

All noble gas activity carried over to the secondary side through SG tube leakage is assumed to be immediately released to the outside atmosphere.

For Indian Point 3, plant cooldown to RHR operating conditions can be accomplished within 29 hours3.356481e-4 days 0.00806 hours 4.794974e-5 weeks 1.10345e-5 months after initiation of the locked rotor event. At 29 hours3.356481e-4 days 0.00806 hours 4.794974e-5 weeks 1.10345e-5 months after the accident, the RHR System is assumed to be placed into service for heat removal and there is no further steam release to the atmosphere from the secondary system. A primary and secondary side mass and energy balance was used to calculate the steam released from the steam generators from 0 to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and from 2 to 29 hours3.356481e-4 days 0.00806 hours 4.794974e-5 weeks 1.10345e-5 months .

4.1.1 Control Room Isolation It is assumed that the control room HVAC system begins in normal operation mode.

Once a high radiation alarm signal is reached there is a 20-minute manual (operator) action to place the control room in post accident operation mode. There are two radiation monitors which will produce a high radiation alarm. RI is inside the control room and R33 is inside the HVAC duct. The monitor that reaches its setpoint last was credited in the analysis. The dose rate in the control room exceeds the RI setpoint within 3 minutes.

The activity level in the air supply exceeds the R33 setpoint within 12 minutes. It is conservatively assumed that the control room HVAC does not fully enter the accident mode of operation until 32 minutes after event initiation.

4.2 Acceptance Criteria The offsite dose limit for a locked rotor is 2.5 rem TEDE per RG 1.183. This is 10% of the guideline value of 10CFR50.67. The limit for the control room dose is 5.0 rem TEDE per 10CFR50.67.

4.3 Results and Conclusions The locked rotor doses are:

Site Boundary 1.I rem TEDE Low Population Zone 1.4 rem TEDE Control Room 2.5 rem TEDE The acceptance criteria are met.

The site boundary doses reported are for the worst two-hour period, determined to be from 27 to 29 hours3.356481e-4 days 0.00806 hours 4.794974e-5 weeks 1.10345e-5 months .

The control room dose from activity outside the control room envelope was determined to be negligible.

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5.0 Rod Ejection Radiological Analysis It is assumed that a mechanical failure of a control rod mechanism pressure housing has occurred, resulting in the ejection of a rod cluster control assembly and drive shaft. As a result of the accident, some fuel clad damage and a small amount of fuel melt are assumed to occur. Due to the pressure differential between the primary and secondary systems, radioactive reactor coolant is discharged from the primary into the secondary system. A portion of this radioactivity is released to the outside atmosphere through either the main condenser, the atmospheric relief valves or the main steam safety valves.

Iodine and alkali metals group activity is contained in the secondary coolant prior to the accident, and some of this activity is released to the atmosphere as a result of steaming the steam generators following the accident. Finally, radioactive reactor coolant is discharged to the containment via the spill from the opening in the reactor vessel head. A portion of this radioactivity is released through containment leakage to the environment.

5.1 Input Parameters and Assumptions Separate calculations are performed to calculate the dose resulting from the release of activity to containment and subsequent leakage to the environment and the dose resulting from the leakage of activity to the secondary system and subsequent release to the environment. The total offsite and control room doses are the sum of the doses resulting from each of the postulated release paths and nuclides considered. A summary of input parameters and assumptions is provided in Table 18.

5.1.1 Source Term Less than 10% of the fuel rods in the core undergo DNB as a result of the rod ejection accident. In determining the offsite doses following a rod ejection accident, it is conservatively assumed that 10% of the fuel rods in the core suffer sufficient damage that all of their gap activity is released. Consistent with RG 1.183, ten percent of the total core activity of iodine and noble gases and 12 percent of the total core activity of alkali metals are assumed to be in the fuel-cladding gap. In the calculation of activity releases from the failed/melted fuel the maximum radial peaking factor of 1.7 was applied.

A small fraction of the fuel in the failed fuel rods is assumed to melt as a result of the rod ejection accident. It is assumed that 0.25% of the core melts. This is based on the assumption that 50% of the rods in DNB undergo centerline melting, with the melting limited to the inner 10% and occurring over 50% of the axial length of the affected rods.

For both the containment leakage release path and the primary to secondary leakage release path, all noble gas and alkali metal activity released from the failed fuel (both gap activity and activity from melted fuel) is available for release.

For the containment leakage release path, all of the iodine released from the gap of failed fuel and 25 percent of the activity released from melted fuel is available for release from containment.

For the primary to secondary leakage release path, all of the iodine released from the gap of failed fuel and 50 percent of the activity released from melted fuel is available for release from the reactor coolant system.

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Prior to the accident, the iodine activity concentration of the primary coolant is 1.0 piCi/gm of DE I-13 1 (ITS 3.4.16.2) and is given in Table 15. The noble gas and alkali metal activity concentration in the RCS at the time the accident occurs is based on a one percent fuel defect level and is given in Table 6. The iodine activity concentration of the secondary coolant at the time the rod ejection occurs is assumed to be equivalent to the technical specification limit of 0.1 gCi/gm of DE 1-131 and is given in Table 15. The alkali metal activity concentration of the secondary coolant at the time the rod ejection occurs is assumed to be 10% of the primary side concentration.

Iodine in containment is assumed to be 4.85% elemental, 0.15% organic and 95%

particulate.

Iodine released from the secondary system is assumed to be 97% elemental and 3%

organic.

5.1.2 Containment Release Pathway The containment is assumed to leak at the design leak rate of 0.1% per day for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of the accident and then to leak at half that rate (0.05% per day) for the remainder of the 30-day period following the accident considered in the analysis.

For the containment leakage pathway, no credit is taken for plateout onto containment surfaces or for containment spray operation which would remove airborne particulates and elemental iodine. Removal of iodine and alkali metal particulates by the fan cooler unit HEPA filters in containment is credited. No credit is taken for the fan cooler unit charcoal filters.

The analysis also considered the potential for complete elimination of the fan cooler unit filters combined with elimination of the spray additive (NaOH solution). Associated with the elimination of the spray additive would be the addition of trisodium phosphate stored in baskets in the containment sump to provide a passive approach to adjusting sump solution pH. If these changes were to be made to the plant, there would be no aerosol removal by the fan cooler filters but sedimentation removal of iodine and alkali metal particulates in containment would be credited. Retention of iodine in the sump solution is assured by the adjustment of the sump solution to a pH greater than or equal to 7.0.

This pH adjustment would be provided by the trisodium phosphate stored in the containment.

5.1.3 Primary to Secondary Leakage Release Pathway When determining doses due to the primary to secondary steam generator tube leakage, all the iodine, alkali metals group and noble gas activity (from prior to the accident and resulting from the accident) is assumed to be in the primary coolant (and not in the containment). The primary to secondary tube leakage and steaming from the steam generators continues until the reactor coolant system pressure drops below the secondary pressure. A bounding time of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months was selected for this analysis, although analyses of the small break LOCA pressure transient have shown that this would occur well before that time. A rod ejection pressure transient is similar to that of a small break LOCA.

Steam releases from the steam generators are conservatively assumed to continue for 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . The locked rotor steam releases are conservatively applied for this analysis. The locked rotor releases are conservative since they do not include ECCS injection to absorb decay heat.

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The amount of primary to secondary SG tube leakage is assumed to be equal to the technical specification limit of I gpm total. Although the primary to secondary pressure differential drops throughout the event, the constant flow rate is maintained.

An iodine partition factor in the SGs of 0.01 (curies iodine/gm steam) / (curies iodine/gm water) is used.

The retention of particulates in the SGs is limited by the moisture carryover. The moisture carryover is 0.1%. Therefore, an alkali metal partition factor in the SGs of 0.001 (curies iodine/gin steam) / (curies iodine/gm water) is used.

All noble gas activity carried over to the secondary side through SG tube leakage is assumed to be immediately released to the outside atmosphere.

5.1.4 Control Room Isolation In the rod ejection, as in the small break LOCA analysis, the low pressurizer pressure SI setpoint will be reached within 80 seconds from event initiation. The SI signal causes the control room HVAC to switch from the normal operation mode to the accident mode of operation. It is conservatively assumed that the control room HVAC does not fully enter the accident mode of operation until 140 seconds after event initiation.

5.2 Acceptance Criteria The offsite dose limit for a rod ejection is 6.3 rem TEDE per RG 1.183. This is -25% of the guideline value of 10CFR50.67. The limit for the control room dose is 5.0 rem TEDE per 10CFR50.67.

5.3 Results and Conclusions The rod ejection doses are:

Site Boundary 4.4 rem TEDE Low Population Zone 2.2 rem TEDE Control Room 0.9 rem TEDE The acceptance criteria are met.

The site boundary doses reported are for the worst two-hour period, determined to be from 0.0 to 2.0 hours0 days 0 hours 0 weeks 0 months .

The control room dose from activity outside the control room envelope was determined to be negligible.

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For the case in which it is assumed that the HEPA filters have been removed from the fan cooler filters and trisodium phosphate is provided to increase sump solution pH to

>7.0, the rod ejection doses are:

Site Boundary 5.2 rem TEDE Low Population Zone 3.9 rem TEDE Control Room 1.6 rem TEDE The acceptance criteria are met.

The site boundary doses reported are for the worst two-hour period, determined to be from 0.0 to 2.0 hours0 days 0 hours 0 weeks 0 months .

The control room dose from activity outside the control room envelope was determined to be negligible.

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6.0 Steam Line Break Radiological Analysis The complete severance of a main steam line outside containment is assumed to occur.

The affected SG will rapidly depressurize and release radioiodines initially contained in the secondary coolant and primary coolant activity, transferred via SG tube leaks, directly to the outside atmosphere. A portion of the iodine activity initially contained in the intact SGs and noble gas activity due to tube leakage is released to atmosphere through either the atmospheric dump valves (ADVs) or the safety valves (MSSVs). The steam line break outside containment will bound any break inside containment since the outside break provides a means for direct release into the environment. This section describes the assumptions and analyses performed to determine the amount of radioactivity released and the offsite and control room doses resulting from this release.

6.1 Input Parameters and Assumptions The analysis of the steam line break (SLB) radiological consequences uses the analytical methods and assumptions outlined in the RG 1.183. A summary of input parameters and assumptions is provided in Table 19.

For the pre-accident iodine spike case, it is assumed that a reactor transient has occurred prior to the SLB and has raised the RCS iodine concentration to the technical specification limit for a transient of 60 jiCi/gm of dose equivalent (DE) I-131. For the accident-initiated iodine spike case, the reactor trip associated with the SLB creates an iodine spike in the RCS which increases the iodine release rate from the fuel to the RCS to a value 500 times greater than the release rate corresponding to a maximum equilibrium RCS concentration of 1.0 ACi/gm of DE I-131. The duration of the accident-initiated iodine spike is limited by the amount of activity available in the fuel-clad gap of fuel pins containing cladding defects. Based on having 8 percent of the iodine in the fuel-clad gap, the gap inventory would be depleted within 3.0 hours0 days 0 hours 0 weeks 0 months and the spike is terminated at that time.

The noble gas activity concentration in the RCS at the time the accident occurs is based on a one percent fuel defect level. This is approximately equal to the technical specification value of 1OO/E bar ptCi/gm for gross radioactivity. The iodine activity concentration of the secondary coolant at the time the SLB occurs is assumed to be equivalent to the technical specification limit of 0.1 4Ci/gm of DE 1-131.

The technical specification limit for primary to secondary leakage is 1 gpm maximum plant total with 432 gpd (0.3 gpm) maximum per single steam generator. Consistent with RG 1.183, the primary to secondary leakge is apportioned between the faulted and intact steam generators to maximize the calculated doses. The primary to secondary leakage is modeled with 0.3 gpm going to the faulted steam generator and 0.7 gpm going to the intact steam generators. The density for this leakage is 62.4 lbvft3 The SG connected to the broken steam line is assumed to boil dry within the initial five minutes following the SLB. The entire liquid inventory of this SG is assumed to be steamed off and all of the iodine initially in this SG is released to the environment. In addition, iodine carried over to the faulted SG by tube leaks is assumed to be released directly to the environment with no credit taken for iodine retention in the SG.

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An iodine partition factor in the intact SGs of 0.01 (curies iodine/gm steam) / (curies iodine/gm water) is used.

All noble gas activity carried over to the secondary side through SG tube leakage is assumed to be immediately released to the outside atmosphere.

For Indian Point 3, plant cooldown to RHR operating conditions can be accomplished within 29 hours3.356481e-4 days 0.00806 hours 4.794974e-5 weeks 1.10345e-5 months after initiation of the main steam line break event. At 29 hours3.356481e-4 days 0.00806 hours 4.794974e-5 weeks 1.10345e-5 months after the accident, the RHR system is assumed to be placed into service for heat removal and there is no further steam release to the atmosphere from the intact steam generators. A primary and secondary side mass and energy balance was used to calculate the steam released from the intact steam generators from 0 to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and from 2 to 29 hours3.356481e-4 days 0.00806 hours 4.794974e-5 weeks 1.10345e-5 months .

Within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months after the accident, the reactor coolant system has been cooled to below 2121F, and there are no further steam releases to atmosphere from the faulted steam generator.

6.1.1 Control Room Isolation In the event of a SLB, the low steamline pressure SI setpoint will be reached shortly after event initiation. The SI signal causes the control room HVAC to switch from the normal operation mode to the accident mode of operation. It is conservatively assumed that the control room HVAC does not fully enter the accident mode of operation until I minute after event initiation.

6.2 Acceptance Criteria The offsite dose limit for a SLB with a pre-accident iodine spike is 25 rem TEDE per RG 1.183. This is the guideline value of I OCFR50.67. For a SLB with an accident-initiated iodine spike, the offsite dose limit is 2.5 rem TEDE per RG 1.183.

This is 10% of the guideline value of IOCFR50.67. The limit for the control room dose is 5.0 rem TEDE per 10CFR50.67.

6.3 Results and Conclusions The SLB accident doses are listed below.

For the pre-accident iodine spike:

Site Boundary 0.2 rem TEDE Low Population Zone 0.3 rem TEDE Control Room 0.6 rem TEDE For the accident-initiated iodine spike:

Site Boundary 0.5 rem TEDE Low Population Zone 0.8 rem TEDE Control Room 2.1 rem TEDE The acceptance criteria are met.

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The site boundary doses reported are for the worst two-hour period, determined to be from 0.0 to 2.0 hours0 days 0 hours 0 weeks 0 months for the pre-accident iodine spike and from 3.0 to 5.0 hours0 days 0 hours 0 weeks 0 months for the accident initiated iodine spike.

The control room doses from activity outside the control room envelope was determined to be negligible.

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7.0 Small Break LOCA Radiological Analysis An abrupt failure of the primary coolant system is assumed to occur and it is assumed that the break is small enough that the containment spray system is not immediately actuated by high containment pressure but that the core experiences substantial cladding damage such that the fission product gap activity of all fuel rods is released. Activity that is released to the containment is assumed to be released to the environment due to the containment leaking at its design rate. There is also a release path through the steam generators (primary to secondary) until the primary system becomes depressurized to below the secondary system pressure.

7.1 Input Parameters and Assumptions The analysis of the SBLOCA radiological consequences uses the analytical methods and assumptions outlined in RG 1.183, Appendix H (Rod Ejection) for release path modeling and RG 1.183, Table 2 for gap fractions.

Separate calculations are performed to calculate the dose resulting from the release of activity to containment and subsequent leakage to the environment and the dose resulting from the leakage of activity to the secondary system and subsequent release to the environment. The total offsite and control room doses are the sum of the doses resulting from each of the postulated release paths and nuclides considered. A summary of input parameters and assumptions is provided in Table 20.

7.1.1 Source Term In determining the offsite doses following a SBLOCA, it is assumed that all of the fuel rods in the core suffer sufficient damage that all of their gap activity is released and that no fuel in the core melts. Five percent of iodines, noble gases and alkali metals of the total core activity is in the fuel-cladding gap.

For both the containment leakage release path and the primary to secondary leakage release path all iodine, noble gas and alkali metal activity in the failed fuel gap is available for release.

It is assumed that a reactor transient has occurred prior to the accident and has raised the primary coolant iodine concentration to 60 j+/-Ci/gm of DE 1-131. The noble gas and alkali metal activity concentration in the RCS at the time the accident occurs is based on a one percent fuel defect level.

Iodine in containment is assumed to be 4.85% elemental, 0. 15% organic and 95%

particulate.

Iodine released from the secondary system is assumed to be 97% elemental and 3%

organic.

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7.1.2 Containment Release Pathway The containment is assumed to leak at the design leak rate of 0.1% per day for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of the accident and then to leak at half that rate (0.05% per day) for the remainder of the 30 day period following the accident considered in the analysis.

For the containment leakage pathway, no credit is taken for plateout onto containment surfaces or for containment spray operation which would remove airborne particulates and elemental iodine. Removal of iodine and alkali metal particulates by the fan cooler unit HEPA filters in containment is credited. No credit is taken for the fan cooler unit charcoal filters.

The analysis also considered the potential for complete elimination of the fan cooler unit filters combined with elimination of the spray additive (NaOH solution). Associated with the elimination of the spray additive would be the addition of trisodium phosphate stored in baskets in the containment sump to provide a passive approach to adjusting sump solution pH. If these changes were to be made to the plant, there would be no aerosol removal by the fan cooler filters but sedimentation removal of iodine and alkali metal particulates in containment would be credited. Retention of iodine in the sump solution is assured by the adjustment of the sump solution to a pH greater than or equal to 7.0.

This pH adjustment would be provided by the trisodium phosphate stored in the containment.

7.1.3 Primary to Secondary Leakage Release Pathway When determining doses due to the primary to secondary steam generator tube leakage, all the iodine, alkali metals group and noble gas activity (from prior to the accident and resulting from the accident) is assumed to be in the primary coolant (and not in the containment). The primary to secondary tube leakage and the steaming from the steam generators continue until the reactor coolant system pressure drops below the secondary pressure. A conservative time of I hour was used for this analysis, although analyses of the small break LOCA pressure transient have shown that this would occur well before that time. Steam releases from the steam generators are conservatively assumed to continue for 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . The locked rotor steam releases are conservatively applied for this analysis. The locked rotor releases are conservative since they do not include ECCS injection to absorb decay heat.

The amount of primary to secondary SG tube leakage is assumed to be equal to the technical specification limit of I gpm total. Although the primary to secondary pressure differential gradually drops, the constant flow rate is conservatively maintained.

An iodine partition factor in the SGs of 0.01 (curies iodine/gm steam) / (curies iodine/gm water) is used.

The retention of particulates in the SGs is limited by the moisture carryover. The maximum moisture carryover is 0.1%. Therefore, an alkali metal partition factor in the SGs of 0.001 (curies iodine/gm steam) / (curies iodine/gm water) is used.

All noble gas activity carried over to the secondary side through SG tube leakage is assumed to be immediately released to the outside atmosphere.

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7.1.4 Control Room Isolation In the small break LOCA analysis, the low pressurizer pressure SI setpoint will be reached within 80 seconds from event initiation. The SI signal causes the control room HVAC to switch from the normal operation mode to the accident mode of operation. It is conservatively assumed that the control room HVAC does not fully enter the accident mode of operation until 140 seconds after event initiation.

7.2 Acceptance Criteria The offsite dose limit for a LOCA is 25 rem TEDE per RG 1.183. This is the guideline value of 10CFR50.67. The limit for the control room dose is 5.0 rem TEDE per 10CFR50.67.

7.3 Results and Conclusions The SBLOCA doses are:

Site Boundary 11.0 rem TEDE Low Population Zone 5.5 rem TEDE Control Room 2.2 rem TEDE The acceptance criteria are met.

The site boundary doses reported are for the worst two-hour period, determined to be from 0.0 to 2.0 hours0 days 0 hours 0 weeks 0 months .

The control room dose from activity outside the control room envelope was determined to be negligible.

For the case in which it is assumed that the HEPA filters have been removed from the fan cooler filters and trisodium phosphate is provided to increase sump solution pH to

>7.0, the doses for the small-break LOCA are:

Site Boundary 13.1 rem TEDE Low Population Zone 9.8 rem TEDE Control Room 4.1 rem TEDE The acceptance criteria are met.

The site boundary doses reported are for the worst two-hour period, determined to be from 0.0 to 2.0 hours0 days 0 hours 0 weeks 0 months .

The control room dose from activity outside the control room envelope was determined to be negligible.

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8.0 Gas Decay Tank Rupture Radiological Analyses For the gas decay tank rupture, a failure is assumed that results in the release of the contents of one gas decay tank.

8.1 Input Parameters and Assumptions The major assumptions and parameters used to determine the doses due to the gas decay tank rupture are given in Table 21.

Consistent with the analysis of record the tank contents are assumed to be at the administratively controlled limit of 50,000 Curies of dose equivalent Xe-133. Dose equivalent Xe-133 is the amount of Xe-133 that results in the same gamma radiation dose as a given mixture of noble gases. A failure in the gaseous waste processing system is assumed to result in release of the tank inventory with a release duration of 5 minutes.

8.1.1 Control Room Isolation It is assumed that the control room HVAC system begins in normal operation mode. A high radiation monitor signal will notify the operators to place the control room in the accident mode of operation. While 20 minutes is the longest period of time anticipated for the operator to take action, it is conservative to switch the control room to the accident mode of operation at the end of the 5-minute release period. This minimizes the amount of purging of activity from the control room. Thus, the control room is analyzed with an isolation time of 5 minutes.

Although the unfiltered inleakage rate of 700 cfm listed in Table 4 is limiting for other events, a lower rate was determined to be conservative for the tank ruptures. After the activity release period, the air outside the control room has no activity, while the control room has a high concentration of noble gases. The inleakage would thus dilute the concentration of activity in the control room. Minimizing this inleakage maximizes the dose. The analysis conservatively neglected all inleakage in the calculation of the control room doses.

8.2 Acceptance Criteria The offsite dose limit for a gas decay tank rupture is 0.5 rem TEDE consistent with the guidance of Regulatory Guide 1.26 (Reference 13) which specifies 0.5 rem whole body or equivalent to any part of the body and of RG 1.183 (Reference 2) which specifies that doses will be determined as TEDE. The limit for the control room dose is 5.0 rem TEDE per IOCFR50.67.

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8.3 Results and Conclusions The gas decay tank rupture doses are:

Site Boundary 0.4 rem TEDE Low Population Zone 0.2 rem TEDE Control Room 0.1 rem TEDE The acceptance criteria are met.

The site boundary doses reported are for the worst two-hour period, which is from 0.0 to 2.0 hours0 days 0 hours 0 weeks 0 months .

The control room dose from activity outside the control room envelope was determined to be negligible.

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9.0 Volume Control Tank Rupture For the volume control tank rupture, a failure is assumed that results in the release of activity from the tank plus the noble gases and a fraction of the iodines from the letdown flow until the letdown path is isolated.

9.1 Input Parameters and Assumptions The major assumptions and parameters used to determine the doses due to the volume control tank rupture are given in Table 22.

The inventory of gases in the tank is based on continuous operation with one percent fuel defects and without any purge of the gas space. The inventory of iodine in the tank is based on operation of the plant with a primary coolant iodine concentration of 1.0 pCi/gm dose equivalent I-131 and with 90 percent of the iodine removed by the letdown demineralizer.

As a result of the accident, all of the noble gas in the tank and 1.0 percent of the iodine in the tank liquid is assumed to be released to the atmosphere over a period of 5 minutes.

After event initiation, letdown flow to the volume control tank continues at the maximum flow rate of 132 gpm (maximum letdown flow plus 10-percent uncertainty) until 30 minutes when the letdown line is assumed to be isolated. The primary coolant noble gas activities are based on operation with one percent fuel defects. The primary coolant iodine activity is assumed to be at the technical specification limit of 1.0 ACi/gm dose equivalent 1-131, which is reduced by 90 percent by the letdown demineralizer. All of the noble gas and 10 percent of the iodine in the letdown flow are assumed to become airborne and are released to the environment.

9.1.1 Control Room Isolation It is assumed that the control room HVAC system begins in normal operation mode. A high radiation monitor signal will notify the operators to place the control room in the accident mode of operation. While 20 minutes is the longest period of time anticipated for the operator to take action, it is conservative to switch the control room to the accident mode of operation at the end of the 5-minute tank activity release period. This minimizes the amount of purging of activity from the control room. Thus, the control room is analyzed with an isolation time of 5 minutes.

As discussed in Section 8.1.1, the tank rupture analyses conservatively neglected inleakage to the control room.

9.2 Acceptance Criteria The offsite dose limit for a volume control tank rupture is 0.5 rem TEDE consistent with the guidance of Regulatory Guide 1.26 (Reference 13) which specifies 0.5 rem whole body or equivalent to any part of the body. Since the dose limits for alternative source term application are in terms of TEDE (per RG 1.183 - Reference 2), the dose limit is 0.5 rem TEDE. The limit for the control room dose is 5.0 rem TEDE per I OCFR50.67.

Page 30 of 69

9.3 Results and Conclusions The volume control tank rupture doses are:

Site Boundary 0.42 rem TEDE Low Population Zone 0.16 rem TEDE Control Room 0.08 rem TEDE The acceptance criteria are met.

The site boundary doses reported are for the worst two-hour period, which is from 0.0 to 2.0 hours0 days 0 hours 0 weeks 0 months .

The control room dose from activity outside the control room envelope was determined to be negligible.

Page 31 of 69

10.0 Holdup Tank Failure During normal plant operation, water is added to the holdup tanks periodically as the primary coolant is diluted during the fuel cycle to provide reduction in the primary coolant boron concentration. As water enters the holdup tank, gases (the nitrogen cover gas and the noble gas and hydrogen that evolve out of solution from the water entering the tank) are displaced to the gaseous waste system. For the holdup tank failure, a failure is assumed that results in the release of the contents of the tank.

10.1 Input Parameters and Assumptions There are 3 holdup tanks and operation alternates between two tanks with the third tank retained empty for reserve holdup capability.

The major assumptions and parameters used to determine the doses due to the holdup tank failure are given in Table 23.

This analysis assumes that a holdup tank is filled with water over a 24-hour period and none of the gaseous activity entering the tank is discharged to the gaseous waste system.

The tank failure is assumed to occur immediately after the tank is filled.

The inventory of gases in the tank is based on plant operation with one percent fuel defects. As a result of the accident, all of the noble gas and iodine activity in the tank is assumed to be released to the atmosphere over a period of ten minutes.

10.1.1 Control Room Isolation It is assumed that the control room HVAC system begins in normal operation mode. A high radiation monitor signal will notify the operators to place the control room in the accident mode of operation. While 20 minutes is the longest period of time anticipated for the operator to take action, it is conservative to switch the control room to the accident mode of operation at the end of the 5-minute release period. This minimizes the amount of purging of activity from the control room. Thus, the control room is analyzed with an isolation time of 5 minutes.

As discussed in Section 8.1.1, the tank rupture analyses conservatively neglected inleakage to the control room.

10.2 Acceptance Criteria The offsite dose limits for a holdup tank rupture is 0.5 rem TEDE consistent with the guidance of Regulatory Guide 1.26 (Reference 13) which specifies 0.5 rem whole body or equivalent to any part of the body and of RG 1.183 (Reference 2) which specifies that doses will be determined as TEDE. The limit for the control room dose is 5.0 rem TEDE per IOCFR50.67.

Page 32 of 69

10.3 Results and Conclusions The holdup tank failure doses are:

Site Boundary 0.38 rem TEDE Low Population Zone 0.14 rem TEDE Control Room 0.10 rem TEDE The acceptance criteria are met.

The site boundary doses reported are for the worst two-hour period, which is from 0.0 to 2.0 hours0 days 0 hours 0 weeks 0 months .

The control room dose from activity outside the control room envelope was determined to be negligible.

Page 33 of 69

11.0 Fuel Handling Accident This accident assumes that a fuel assembly is dropped and damaged during refueling.

Analysis of the accident was performed with assumptions selected so that the results would be bounding for the accident occurring either inside containment or in the Fuel Storage Building. Activity released from the damaged assembly was assumed to be released to the outside atmosphere through either the Containment Purge System or the Fuel Storage Building Ventilation System.

The FHA was analyzed previously using the AST methodology and an SER was issued in March 2003 (Reference 14). That analysis took credit for 75% of the fuel rods in the damaged assembly having the lower fission product gap fractions identified in Table 3 of RG 1.183 (Reference 2). The analysis has been revised to no longer take credit for the RG 1.183 gap fractions. Additionally, the analysis has been revised to reflect changes in the control room HVAC flow rates and allowable inleakage.

11.1 Input Parameters and Assumptions The input parameters and assumptions are listed in Table 24.

The analysis of the FHA radiological consequences was performed using the analytical methods and assumptions outlined in RG 1.183 (Reference 2). This analysis allowed fuel movement 84 hours9.722222e-4 days 0.0233 hours 1.388889e-4 weeks 3.1962e-5 months after shutdown.

All activity released from the water pool was assumed to be released to the atmosphere in 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , using a linear release model (this is the release model used in the existing licensing basis for this event). No credit was taken for operating the Fuel Storage Building Ventilation System. No credit was taken for isolating containment for the FHA in containment. Since the assumptions and parameters for an FHA inside containment are identical to those for a FHA in the Fuel Storage Building, the radiological consequences were the same regardless of the location of the accident.

11.1.1 Source Term The calculation of the radiological consequences following an FHA used gap fractions of 12 percent for I-131, 30 percent for Kr-85, and 10 percent for all other nuclides. The value for 1-131 was taken from NUREG/CR-5009 (Reference 15). The values for Kr-85 and the other iodines and noble gases were taken from RG 1.25 (Reference 16). There are lower values identified in Table 3 of RG 1.183 (Reference 2), but these were not used because the conditions for their use (specified in footnote 11 in RG 1.183) have not been ensured.

As in the existing licensing basis, it was assumed that all of the fuel rods in the equivalent of one fuel assembly would be damaged to the extent that all of their gap activity would be released. The assembly inventory was based on the assumption that the subject fuel assembly had been operated at 1.7 times the core average power. The activity was conservatively increased by 4 percent to bound variations in core average enrichment, core mass, and cycle length (Table 25).

The decay time used in the analysis was 84 hours9.722222e-4 days 0.0233 hours 1.388889e-4 weeks 3.1962e-5 months .

Page 34 of 69

11.1.2 Iodine Chemical Form The iodine released from the fuel was assumed to be 95-percent cesium iodide (CsI),

4.85-percent elemental iodine, and 0.15-percent organic iodine. It was assumed that all of the CsI was dissociated in the water and that the iodine re-evolved as elemental iodine.

This was assumed to occur instantaneously. Thus, the FHA dose analysis was based on an initial iodine characterization of 99.85-percent elemental iodine and 0.15-percent organic iodine.

11.1.3 Water Scrubbing Removal of Activity The activity released from the damaged fuel rods was assumed to be contained within gas bubbles that rise up through the water and are released into the atmosphere above the pit.

As the bubbles pass through the water column, there is a significant removal of activity.

RG 1.183 (Reference 2) identifies a DF of 500 for elemental iodine and no removal for organic iodine and noble gases. The DF of 500 for elemental iodine is based on having a water height of 23 feet or more. (Per the Technical Specifications, there are requirements for 223 feet of water above the stored spent fuel and above the reactor vessel flange during fuel-handling operations.)

The DF of 500 for elemental iodine is also based on fuel rod pressure of s1200 psig.

There is the potential for fuel rod pressures to exceed 1200 psig (but remain less than 1500 psig). With this increase in fuel rod pressure, the DF is determined to remain above 400. Using a DF of 400 for elemental iodine and the defined iodine species split of 99.85-percent elemental and 0.15-percent organic, the overall DF would be 250.

However, RG 1.183 (Reference 2) also specifies the overall DF for iodine to be 200. The overall DF of 200 has an associated elemental iodine DF of 285, and this value was used in the analysis together with a DF of 1.0 for organic iodine and noble gases.

The cesium released from the damaged fuel rods was assumed to remain in a nonvolatile form and not be released from the water.

11.1.4 Filtration of Release Paths No credit was taken for removing iodine by filters, nor was credit taken for isolating release paths.

Although the containment purge will be automatically isolated on a purge line high-radiation alarm, isolation was not modeled in the analysis. The activity released from the damaged assembly was assumed to be released to the outside atmosphere over a 2-hour period. Since no filtration or containment isolation was modeled, this analysis supports refueling operation with the equipment hatch and the personnel air lock remaining open.

11.1.5 Control Room Isolation It was assumed that the control room HVAC System is manually switched over from the normal operation mode to the emergency mode of operation after a high radiation alarm is actuated.

It is assumed that the control room HVAC system begins in normal operation mode. A high radiation monitor signal will notify the operators to place the control room in the accident mode of operation. The longest period of time anticipated for the operator to Page 35 of 69

take action is 20 minutes. The control room is analyzed with an isolation time of 24 minutes.

11.2 Acceptance Criteria The offsite dose limit for an FHA is 6.3 rem TEDE per RG 1.183 (Reference 2). This is

-25 percent of the guideline value of IOCFR50.67. The limit for the control room dose is 5 rem TEDE, per IOCFR50.67.

11.3 Results and Conclusions The calculated doses due to the FHA are:

Site Boundary 5.7 rem TEDE Low Population Zone 2.1 rem TEDE Control Room 1.4 rem TEDE The acceptance criteria were met.

The SB dose reported was for the worst 2-hour period, determined to be from 0 to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

The control room dose from activity outside the control room envelope was determined to be negligible.

Page 36 of 69

12.0 Conclusions RG 1.183 (Reference 2) defines an alternative source term model for use in evaluating the radiological consequences of a postulated large break loss-of-coolant accident with core melt. This alternative source term model also forms the basis for determining the radiological consequences for other design basis accidents as provided in RG 1.183.

The alternative source term methodology, as defined in RG 1.183 and its appendices, has been incorporated into the Indian Point 3 Nuclear Power Plant's design basis accident analyses to support control room habitability. Analyses of the radiological consequences of the large break LOCA, steam generator tube rupture, locked rotor, rod ejection, steamline break, small break LOCA, gas decay tank rupture, volume control tank rupture and holdup tank failure have been made using the RG 1.183 methodology. The calculated doses do not exceed the defined acceptance criteria. The fuel handling accident, which had previously been analyzed using the alternative source term methodology (with an SER issued by Reference 14), was reanalyzed using revised source term modeling and revised control room HVAC flow rates and allowable inleakage.

This report supports the following changes to Indian Point 3 Nuclear Power Plant's design and operation:

The elimination of a requirement for filter efficiency operability requirement for the charcoal filters on the containment fan cooler units.

An allowable unfiltered inleakage into the control room of up to 700 cfm.

This report also supports the complete removal of all filters from the fan cooler units (both the HEPA and the charcoal filters) if plant changes are also made to implement a passive means for pH adjustment of the post-accident sump solution (i.e., the installation of baskets filled with trisodium phosphate). With the addition of trisodium phosphate for sump solution pH control, the spray additive can be deleted.

Page 37 of 69

13.0 References

1. TID-14844, "Calculation of Distance Factors for Power and Test Reactor Sites," U.S. AEC, Division of Licensing and Regulation, J. J. DiNunno, et. al, March 23, 1962.

2. Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors", July 2000.

3. NRC Final Rule 10CFR50.67, issued in Federal Register, Vol. 64, No. 246, pages 71990-72002, 12/23/99.

4. U.S. Nuclear Regulatory Commission NUREG-1465, "Accident Source Terms for Light-Water Nuclear Power Plants," February 1995.

5. EPA Federal Guidance Report No. 11, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion,"

EPA-520/1-88-020, September 1988.

6. CCC-371,"ORIGEN2.1: Isotope Generation and Depletion Code - Matrix Exponential Method," RSIC Computer Code Collection, Oak Ridge National Laboratory, February 1996.

7. EPA Federal Guidance Report No. 12, "External Exposure to Radionuclides in Air, Water and Soil," EPA 402-R-93-08 1, September 1993.

8. International Commission on Radiological Protection, "Radionuclide Transformations, Energy and Intensity of Emissions," ICRP Publication 38, 1983.

9. NUREG-0800, Standard Review Plan Section 6.5.2, "Containment Spray as a Fission Product Cleanup System," Revision 2, December 1988.

10. Industry Degraded Core Rulemaking (IDCOR) Program Technical Report 11.3, "Fission Product Transport in Degraded Core Accidents," Atomic Industrial Forum, December 1983

11. Murphy, K. G., Campe, K. M., "Nuclear Power Plant Control Room Ventilation System Design for Meeting General Criterion 19," Proceedings of the Thirteenth AEC Air Cleaning Conference held August 1974, published March 1975.

12. Regulatory Guide 1.4, "Assumptions Used For Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident," (Rev. 2, June 1974).

13. Regulatory Guide 1.26, "Quality Group Classifications and Standards for Water-, Steam-, and Radioactive-Waste-Containing Components of Nuclear Power Plants," Revision 2, June 1975.

14. Letter from P. Milano (NRC) to M. Kansler (Entergy Nuclear Operations. Inc.), "Indian Point Nuclear Generating Unit No. 3 - RE: Issuance of Amendment Affecting Adoption of Alternate Source Term for the Fuel Handling Accident (TAC No. MB5382)," Amendment No. 215, Docket No. 50-286, March 17, 2003.

15. NUREG/CR-5009, "Assessment of the Use of Extended Burnup Fuel in Light Water Reactors," February 1988.

16. NRC Regulatory Guide 1.25, "Assumptions Used for Evaluating the Potential Radiological Consequences of a Fuel Handling Accident in the Fuel Handling and Storage Facility for Boiling and Pressurized Water Reactors," March 1972.

Page 38 of 69

Table 1: Committed Effective Dose Equivalent Dose Conversion Factors Isotope DCF (rem/curie) Isotope DCF (rem/curie)

I-130 2.64E3 Cs-134 4.63E4 1-131 3.29E4 Cs-136 7.33E33 1-132 3.81E2 Cs-137 3.19E4 1-133 5.85E3 Cs-138 1.OIE2 1-134 1.3 1E2 Rb-86 6.62E3 1-135 1.23E3 Ru-103 8.95E3 Kr-85m N/A Ru-105 4.55E2 Kr-85 N/A Ru-106 4.77E5 Kr-87 N/A Rh-l105 9.55E2 Kr-88 N/A Mo-99 3.96E3 Xe-131 m N/ATc-99m 3.26EI Xe-133m N/A Y-90 8.44E3 Xe-133 N/A Y-91 4.88E4 Xe-135m N/A Y-92 7.81EI Xe-135 N/A Y-93 2.15E3 Xe-138 N/A Nb-95 5.81E3 Te-127m 2.15E4 Zr-95 2.36E4 Te- 127 3.18E2 Zr-97 4.33E3 Te-129m 2.39E4 La-140 4.85E3 Te-129 8.95EI La-141 5.81E2 Te-131m 6.4E3 La-142 2.53E2 Te-132 9.44E3 Nd-147 6.84E3 Sb-127 6.03E3 Pr-143 8.10E3 Sb-129 6.44E2 Am-241 4.44E8 Ce-141 8.95E3 Cm-242 1.73E7 Ce-143 3.39E3 Cm-244 2.48E8 Ce-144 3.74E5 Sr-89 4.14E4 Pu-238 3.92E8 Sr-90 1.3E6 Pu-239 4.29E8 Sr-91 1.66E3 Pu-240 4.29E8 Sr-92 8.1E2 Pu-241 8.25E6 Ba-139 1.7E2 Np-239 2.51 E3 Ba-140 3.74E3 Page 39 of 69

Table 2: Effective Dose Equivalent Dose Conversion Factors Isotope DCF (rem m3 /Ci sec) Isotope DCF (rem-m31Ci-sec)

I-130 0.3848 Cs-134 0.2801 1-131 6.734E-2 Cs-136 0.3922 1-132 0.4144 Cs-137 0.1066*

1-133 0.1088 Cs-138 0.4477 I-134 0.4810 Rb-86 1.780E-2 1-135 0.2953 Ru-103 8.325E-2 Kr-85m 2.768E-2 Ru-105 0.1410 Kr-85 4.403E-4 Ru-106 0.0 Kr-87 0.1524 Rh-105 1.376E-2 Kr-88 0.3774 Mo-99 2.694E-2 Xe- 131 m 1.439E-3 Tc-99m 2.179E-2 Xe-133m 5.069E-3 Y-90 7.030E-4 Xe-133 5.772E-3 Y-91 9.620E-4 Xe-135m 7.548E-2 Y-92 4.810E-2 Xe-135 4.403E-2 Y-93 1.776E-2 Xe-138 0.2135 Nb-95 0.1384 Te-127m 5.439E4 Zr-95 0.1332 Te-127 8.954E-4 Zr-97 3.337E-2 Te-129m 5.735E-3 La-140 0.4329 Te-129 1.018E-2 La-141 8.843E-3 Te-131m 0.2594 La-142 0.5328 Te-132 3.811E-2 Nd-147 2.290E-2 Sb-127 0.1232 Pr-143 7.770E-5 Sb-129 0.2642 Am-241 3.027E-3 Ce-141 1.269E-2 Cm-242 2.105E-5 Ce-143 4.773E-2 Cm-244 1.817E-5 Ce-144 3.156E-3 Sr-89 2.860E-4 Pu-238 1.806E-5 Sr-90 2.786E-5 Pu-239 1.569E-5 Sr-91 0.1277 Pu-240 1.758E-5 Sr-92 0.2512 Pu-241 2.683E-7 Ba-139 8.029E-3 Np-239 2.845E-2 Ba-140 3.175E-2

This is the DCF for Ba- 137m. The DCF for Cs-137 is low; however, a significant amount of Ba-137m is produced through decay. This is conservatively addressed by applying the DCF from Ba-137m to Cs-137.

Page 40 of 69

Table 3: Offsite Breathing Rates and Atmospheric Dispersion Factors Time Offsite Breathing Rates (m3 /sec) 0 - 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 3.5E-4 8 - 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 1.8E-4

>24 hours 2.3E-4 Offsite Atmospheric Dispersion Factors (sec/M3 )

Exclusion Area Boundary~') 1.03E-3 Low Population Zone 0 - 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 3.8E-4 2 - 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 1.9E-4

> I days 1.7E-5 Note:

1. This exclusion area boundary atmospheric dispersion factor is conservatively applied during all time intervals in the determination of the limiting 2-hour period.

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Table 4: Control Room Parameters Volume 47,200 ft3 Control Room Unfiltered In-Leakage <700 cfm(')

Normal Ventilation Flow Rates Filtered Makeup Flow Rate 0.0 cfm Filtered Recirculation Flow Rate 0.0 cfm Unfiltered Makeup Flow Rate <1500 cfml Unfiltered Recirculation Flow Rate (Not modeled - no impact on analyses)

Post Accident Recirculation Flow Rates Option I Option 2 Filtered Makeup Flow Rate >400 cfm >1500 cfm Filtered Recirculation Flow Rate >1000 cfm 0.0 cfm Unfiltered Makeup Flow Rate 0.0 cfm 0.0 cfm Unfiltered Recirculation Flow Rate (Not modeled - no (Not modeled - no impact on analyses) impact on analyses)

Filter Efficiencies Elemental Iodine 90%

Organic (Methyl) Iodine 90%

Particulate 99%

R33 CR Radiation Monitor Setpoint 3.33E-4 pCi/mL (2)

R33 CR Radiation Monitor Location Ventilation Line drawing from CR bulk air RI CR Gamma Dose Area Monitor Setpoint I mrad/hr RI CR Gamma Dose Monitor Location Wall in the control room outside of duct Delay to Switch CR HVAC from Normal Operation to 20 minutes Post Accident Operation after receiving a High Alarm Signal (radiation monitor) Based on Manual Action (min)

Delay to Switch CR HVAC from Normal Operation to 60 seconds Post Accident Operation after receiving an SI signal (sec)

Breathing Rate - Duration of the Event 3.5E-4 m3 /sec Occupancy Factors 0 - 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 1.0 1 - 4 days 0.6 4 - 30 days 0.4 Notes:

1. All of the reported doses modeled a maximum 700 cfm unfiltered inleakage modeled in both CR modes of operation, with the exception of the tank ruptures (Sections 8, 9 and 10) which conservatively ignored inleakage.

2. The monitor setpoint is based on a 0.2 Mev/disintegration source, which is similar to Xe-133.

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Table 5: Core Total Fission Product Activities (Based on 102% of 3216 MWt)

Isotope Activity (Ci) Isotope Activity (Ci) 1-130 3.78E+06 Cs-134 2.05E+07 1-131 9.1OE+07 Cs-136 5.96E+06 1-132 1.33E+08 Cs-137 1.19E+07 1-133 1.88E+08 Cs-138 1.72E+08 1-134 2.06E+08 Rb-86 2.36E+05 1-135 1.76E+08 Ru- 103 1.39E+08 Kr-85m 2.44E+07 Ru-105 9.58E+07 Kr-85 1.I IE+06 Ru-106 4.84E+07 Kr-87 4.69E+07 Rh-105 8.83E+07 Kr-88 6.60E+07 Mo-99 1.75E+08 Xe-131m 9.92E+05 Tc-99m 1.53E+08 Xe-133m 5.45E+06 Xe-133 1.79E+08 Y-90 9.16E+06 Xe-135m 3.68E+07 Y-91 1.14E+08 Xe-135 4.77E+07 Y-92 1.21E+08 Xe-138 1.55E+08 Y-93 1.39E+08 Nb-95 1.56E+08 Te-127m 1.28E+06 Zr-95 1.54E+08 Te-127 9.83E+06 Zr-97 1.55E+08 Te-129m 4.28E+06 La-140 1.65E+08 Te-129 2.92E+07 La-141 1.53E+08 Te-131m 1.33E+07 La-142 1.48E+08 Te-132 1.30E+08 Nd-147 6.07E+07 Sb-127 9.89E+06 Pr- 143 1.37E+08 Sb-129 2.97E+07 Am-241 1.44E+04 Cm-242 3.47E+06 Ce-141 1.52E+08 Cm-244 3.70E+05 Ce-143 1.43E+08 Ce-144 1.20E+08 Sr-89 8.84E+07 Pu-238 4.1 IE+05 Sr-90 8.79E+06 Pu-239 3.50E+04 Sr-91 1.1 IE+08 Pu-240 5.21 E+04 Sr-92 1.20E+08 Pu-241 1.17E+07 Ba-139 1.68E+08 Np-239 1.87E+09 Ba-140 1.60E+08 Page 43 of 69

Table 6: RCS Coolant Concentrations (Based on 1% Fuel Defects)

Isotope Activity (lxCi/gm)

I-130 0.096 1-131 4.67 I-132 3.18 I-133 6.28 I-134 0.682 I-135 3.05 Kr-85m 2.03 Kr-85 13.7 Kr-87 1.30 Kr-88 3.81 Xe-131m 3.23 Xe-133m 3.52 Xe-133 246 Xe-135m 0.625 Xe-135 9.56 Xe-138 0.714 Cs-134 8.82 Cs-136 5.46 Cs-137 4.43 Cs-138 1.08 Rb-86 0.0692 Plant Technical Specification limits primary coolant iodine concentration to 1.0 puCi/gram dose equivalent 1-131. These coolant concentrations are provided in Table 15.

Page 44 of 69

Table 7: Nuclide Decay Constants Decay Constant Decay Constant Isotope (hrb) Isotope (hr')

1-130 0.0561 Cs-134 3.84E-5 1-131 0.00359 Cs-136 2.2E-3 I-132 0.301 Cs-137 2.64E-6 1-133 0.0333 Cs-138 1.29 I-134 0.791 Rb-86 1.55E-3 1-135 0.105 Ru-103 7.35E-4 Kr-85m 0.155 Ru-105 0.156 Kr-85 7.38E-6 Ru-106 7.84E-5 Kr-87 0.545 Rh-105 1.96E-2 Kr-88 0.244 Mo-99 1.OSE-2 Xe-131m 0.00243 Tc-99m 0.115 Xe-133m 0.0132 Y-90 1.08E-2 Xe-133 0.00551 Y-91 4.94E-4 Xe-135m 2.72 Y-92 0.196 Xe-135 0.0763 Y-93 0.0686 Xe-138 2.93 Nb-95 8.22E-4 Te-127m 2.65E-4 Zr-95 4.5 1E-4 Te-127 7.41E-2 Zr-97 4.1E-2 Te-129m 8.6E-4 La-140 1.72E-2 Te-129 0.598 La-141 0.176 Te-131m 2.31E-2 La-142 0.45 Te-132 8.86E-3 Nd-147 2.63E-3 Sb-127 7.5E-3 Pr-143 2.13E-3 Sb-129 0.16 Am-241 1.83E-7 Ce-141 8.89E-4 Cm-242 1.77E-4 Ce-143 0.021 Cm-244 4.37E-6 Ce-144 1.02E-4 Sr-89 5.72E-4 Pu-238 9.02E-7 Sr-90 2.72E-6 Pu-239 3.29E-9 Sr-91 0.073 Pu-240 1.21E-8 Sr-92 0.256 Pu-241 5.5E-6 Ba-139 0.503 Np-239 0.0123 Ba-140 2.27E-3 Page 45 of 69

Table 8: Iodine Chemical Species Iodine Form RG 1.4 RG 1.183 Elemental 91% 4.85%

Organic 4% 0.15%

Particulate 5% 95%

Table 9: Fission Product Release Timing Release Phase Duration (TID-14844) Duration (RG 1.183)(1)

Coolant Activity instantaneous release 10 to 30 seconds Gap Activity instantaneous release 0.5 hour5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months Early In-vessel instantaneous release 1.3 hour3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months Note:

1. Releases are sequential.

Page 46 of 69

Table 10: Core Fission Product Release Fractions Gap Release (l) Early In-Vessel TID RG TED RG Noble gases n/a(2) 0.05 1.0 0.95 Halogens n/a(2) 0.05 0.5(3) 0.35 Alkali Metals n/a 0.05 0.01(4) 0.25 Tellurium group n/a 0 0.0(4) 0.05 Barium, Strontium n/a 0 0.01(4) 0.02 Noble Metals (Ruthenium group) n/a 0 0.0i(4) 0.0025 Cerium group n/a 0 0.01(4) 0.0005 Lanthanides n/a 0 0.01(4) 0.0002 Notes:

1. The TID-14844 methodology does not specifically address the gap release. The RG 1.183 methodology assumes that gap and early in-vessel (core melt) releases are sequential. The TID-14844 source term model assumes the instantaneous release of 50% of core iodine and 100% of noble gases, with no distinction made between gap activity release and early in-vessel release. The RG 1.183 source term assumes a release of gap activity (5% of core) followed by the in-vessel release as defined.

2. Gap fraction is not defined by TID-14844.

3. Per TID- 14844, half of this is assumed to plate out instantaneously.

4. Referred to in TID-14844 as "other fission products" but not typically included in dose analyses.

Page 47 of 69

Table 11: Nuclide Groups Group Title Elements in Group 1 Noble Gases Xe, Kr 2 Halogens I 3 Alkali Metals Cs, Rb 4 Tellurium Group Te, Sb 5 Barium, Strontium Ba, Sr 6 Noble Metals Ru, Rh, Mo, Tc 7 Cerium Group Ce, Pu, Np 8 Lanthanides La, Zr, Nd, Nb, Pr, Y,Cm, Am Page 48 of 69

Table 12: Assumptions Used for Large Break LOCA Dose Analysis Core activity See Table 5 Activity release fractions and timing See Tables 9 & 10 Iodine chemical form in containment Elemental 4.85%

Organic (methyl) 0.15%

Particulate (cesium iodide) 95%

Containment net free volume 2.61 E6 ft3 Containment sprayed volume 2.088E6 ft3 Fan cooler units Number in operation 3 Flow rate (per unit) 34,000 cfm Delay for fan cooler initiation 60 seconds Fan cooler filters Not credited Containment leak rates 0 - 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 0.1 weight %/day

> 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 0.05 weight %/day Spray operation Time to initiate injection sprays 67 seconds Injection sprays duration 43.9 minutes Delay time to switchover to recirculation sprays 3 minutes Time that recirculation sprays are terminated 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Spray flow rates Injection 2200 gpm Recirculation 1050 gpm Spray fall height 118.5 ft Containment spray removal coefficients Spray elemental iodine removal Injection 20 hf' Recirculation 5.0 hf' Spray particulate removal Injection 4.6 hrf Recirculation 2.2 hr-'

Recirculation after DF of 50 is reached 0.22 hrf Sedimentation particulate removal 0.1 hf' (Unsprayed region: from start of event, Sprayed region:

when sprays are not assumed to be operating)

Page 49 of 69

Table 12 (Cont.): Assumptions Used for Large Break LOCA Dose Analysis Containment spray DF Elemental 200 Particulate 1000 Credited containment sump volume 374,400 gal Leakage of sump solution outside of containment 0 - 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months 1.0 gph 4 - 6.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months 0.0 gph

> 6.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months 4.0 gph Iodine airborne fraction for leakage of sump solution outside of containment 0 - 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months 10.0%

4 - 6.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months NA 2.7%

> 6.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months Control Room atmospheric dispersion (X/Q) factors Releases from containment surface('):

0 - 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 3.57E-4 sec/M 3 2 - 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 3.12E-4 sec/M3 8 - 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 1.24E-4 sec/M3 24 - 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months 1.06E-4 sec/M3 96 - 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months 7.99E-5 sec/M3 Control Room atmospheric dispersion (X/Q) factors Releases from containment vent(2):

0 - 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 5.93E-4 sec/M3 2 - 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 4.92E-4 sec/M3 8 - 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 2.06E-4 sec/M3 24 - 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months 1.69E-4 sec/M3 96 - 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months 1.26E-4 sec/M3 Notes:

1. Used for activity released via containment leakage

2. Used for activity released via leakage of sump solution outside of containment (RCP seal leak-off and ECCS)

Page 50 of 69

Table 13: Steam Generator Tube Rupture Thermal Hydraulic Results Reactor Trip Time (sec) 392 Loss of offsite power At reactor trip Tube rupture break flow (Ibm)

Pre-Trip 38,500 Post-Trip (until 30 minutes) 99,500 Tube rupture break flow flashing fraction (%)

Pre-Trip 21 Post-Trip (until 30 minutes) 15 Steam released to environment Ruptured SG Pre-Trip (Ibm/sec) 1070.2 Post-Trip (until 30 minutes) (Ibm) 72,000 Intact SGs Pre-Trip (Ibm/sec) 3210.6 Trip - 2 Hour (Ibm) 526,000 2 - 8 Hour (Ibm) 1,160,000 8 - 29 Hour (Ibm) 1,580,000 Page 51 of 69

Table 14: Assumptions Used for SGTR Dose Analysis Source Data Reactor coolant iodine activity prior to accident Pre-accident iodine spike 60 gCi/gin of DE I-131 - See Table 15 Accident-initiated iodine spike 1.0 gCi/gm of DE 1-131 - See Table 15 Reactor coolant iodine appearance rate 335 times equilibrium rate. Spike appearance rates increase due to the accident-initiated spike are presented in Table 16.

Duration of accident-initiated iodine spike 4.0 hrs Reactor coolant noble gas and alkali metal 1.0% fuel defect level - See Table 6 activity prior to accident Secondary coolant iodine activity prior to 0. IgCi/gm of DE I-131 - See Table 15 accident Reactor coolant initial mass 1.960E8 grams Steam generator initial mass 2.88E7 grams/SG Offsite power Lost at time of reactor trip Species of iodine released to atmosphere Elemental 97%

Organic 3%

Particulate (cesium iodide) 0%

Page 52 of 69

Table 14 (Cont.): Assumptions Used for SGTR Dose Analysis Activity Release Data Ruptured steam generator Rupture flow See Table 13 Flashed rupture flow See Table 13 Steam releases See Table 13 Intact steam generators Primary to secondary leakage 0.3 gpm/SG Steam releases See Table 13 Iodine partition factor Condenser 0.01 SG 0.01 except for flashed break flow Control Room atmospheric dispersion (XIQ) factors Secondary releases:

0 - 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 1.19E-3 sec/M3 2 - 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 1.12E-3 sec/M3 8 - 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 5.59E-4 sec/M3 24 - 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months 4.27E-4 sec/M3 96 - 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months 3.35E14 sec/M3 Page 53 of 69

Table 15: Iodine Specific Activities (piCigm)

Primary Coolant Secondary Coolant Nuclide 1 pCi/gm 60 pCi/gm 0.1 PCi/gm I-130 0.0161 0.97 0.0016 1-131 0.7849 47.09 0.0785 1-132 0.5345 32.07 0.0535 1-133 1.0555 63.33 0.1056 1-134 0.1146 6.88 0.0115 1-135 0.5126 30.76 0.0513 Table 16: Iodine Spike Appearance Rates (Curies/Minute)(')

1-130 1-131 1-132 1-133 I-134 1-135 335 times the 4.2 146.1 314.6 239.0 143.6 165.8 equilibrium rate (SGTR) 500 times the 6.2 218.0 469.6 356.7 214.3 247.5 equilibrium rate (MSLB)

Notes:

1. Calculated based on the RCS concentration of 1.0 pCi/gm DE I-131, letdown flow of 120 gpm +10%

with perfect cleanup and RCS leakage of 11 gpm.

Page 54 of 69

Table 17: Assumptions Used for Locked Rotor Dose Analysis Source Term Core activity See Table 5 Fraction of fuel rods in core assumed to fail for dose considerations 5% of core Gap fractions 1-131 8% of core activity Kr-85 10% of core activity Other Iodine and Noble Gas nuclides 5% of core activity Alkali Metals 12% of core activity Radial peaking factor 1.7 Iodine chemical form after release to atmosphere Elemental 97%

Organic 3%

Particulate (cesium iodide) 0%

Reactor coolant iodine activity prior to accident 1.0 jiCi/gm of DE 1-131 - See Table 15 Reactor coolant noble gas activity prior to 1.0% fuel defect level - See Table 6 accident Reactor coolant alkali metal activity prior to 1.0% fuel defect level - See Table 6 accident Secondary coolant iodine activity prior to accident 0.1 glCi/gm of DE 1-131 - See Table 15 Release Modeling Primary to secondary leakage 1.0 gpm total Steam release to environment 0 - 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 405,000 Ibm 2 - 29 hours3.356481e-4 days 0.00806 hours 4.794974e-5 weeks 1.10345e-5 months 2,303,000 Ibm SG iodine water/steam partition coefficient 0.01 SG alkali metal water/steam partition coefficient 0.001 RCS mass 1.960E8 gm Page 55 of 69

Table 17 (Cont.): Assumptions Used for Locked Rotor Dose Analysis Release Modeling Cont.

Secondary side mass 3.193E7 gm/SG Control Room atmospheric Dispersion (x/Q)

Factors Secondary releases:

0 - 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 1.1 9E-3 sec/M3 2 - 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 1.12E-3 sec/M3 8 - 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 5.59E-4 sec/M3 24 - 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months 4.27E-4 sec/M 3 96 - 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months 3.35E-4 sec/M 3 Page 56 of 69

Table 18: Assumptions Used for Rod Ejection Dose Analysis Source Term Core activity See Table 5 Fraction of fuel rods in core that fail 10 (% of core)

Gap fractions Iodine 10% of core activity Noble Gas 10% of core activity Alkali Metals 12% of core activity Fraction of fuel melting 0.25% of core Radial peaking factor 1.7 Fraction of activity released from failed fuel (gap activity)

Containment leakage 100%

Primary to secondary leakage 100%

Fraction of activity released from melted fuel Containment leakage Iodine 25%

Noble Gas 100%

Alkali Metals 100%

Primary to secondary leakage Iodine 50%

Noble Gas 100%

Alkali Metals 100%

Reactor coolant noble gas and alkali metal activity prior 1.0% fuel defect level - See Table 6 to accident Reactor coolant iodine activity prior to accident 1.0 jiCi/gm of DE I-131 - See Table 15 Secondary coolant iodine activity prior to accident 0.1 jlCi/gm of DE 1-131 - See Table 15 Secondary coolant alkali metal activity prior to accident 10% of primary concentration Page 57 of 69

Table 18 (Cont.): Assumptions Used for Rod Ejection Dose Analysis Containment Leakage Release Path Containment net free volume 2.61 E6 ft3 Containment leak rates 0 - 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 0.1 weight %/day

> 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 0.05 weight %/day Iodine chemical form in containment Elemental 4.85%

Organic 0.15%

Particulate (cesium iodide) 95%

Spray removal in containment Not Credited Aerosol removal by fan cooler unit filters Number of FCUs operating 3 FCU filtered flow rate 8000 cfm/FCU Filter efficiency 0.9 Time delay before filtration begins 60 seconds Sedimentation removal in containment (applicable only for the potential plant modification to delete spray additive and to add baskets of trisodium phosphate in the containment to adjust sump pH post-accident) 0.1 hr-'

Particulate Iodine 0.1 hr-'

Alkali metals Primary to Secondary Leakage Release Path Primary to secondary leakage 1.0 gpm total Steam release to environment 0 - 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 405,000 Ibm

>2 hours 0 Ibm SG iodine water/steam partition coefficient 0.01 SG alkali metal water/steam partition coefficient 0.001 Iodine chemical form after release to atmosphere Elemental 97%

Organic 3%

Particulate (cesium iodide) 0%

RCS mass 1.960E8 gm Page 58 of 69

Table 18 (Cont.): Assumptions Used for Rod Ejection Dose Analysis Secondary side mass I 3.193E7 gm/SG Control Room Atmospheric Dispersion Factors Control Room atmospheric dispersion (X/Q) factors Releases from containment surface('):

0 - 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 3.57E-4 sec/M3 2 - 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 3.12E-4 sec/M3 8 - 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 1.24E-4 sec/M3 24 - 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months 1.06E-4 sec/M3 96 - 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months 7.99E-5 sec/M3 Control Room atmospheric dispersion (X/Q) factors Secondary releases(2):

0-2 hours 1.19E-3 sec/M 3 2 - 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 1.12E-3 sec/M3 8 - 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 5.59E4 sec/M3 24 - 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months 4.27E-4 sec/M3 96 - 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months 3.35E-4 sec/M3 Notes:

1. Used for activity released via containment leakage.

2. Used for activity released via secondary releases.

Page 59 of 69

Table 19: Assumptions Used for Steam Line Break Dose Analysis Source Term Reactor coolant iodine activity prior to accident Pre-accident iodine spike 60 g+/-Ci/gm of DE 1-131 - See Table 15 Accident-initiated iodine spike 1.0 J4Ci/gm of DE 1-131 - See Table 15 Reactor coolant iodine appearance rate increase due 500 times equilibrium rate. Spike appearance to the accident-initiated spike rates are presented in Table 16.

Duration of accident-initiated iodine spike 3.0 hrs Reactor coolant noble gas activity prior to accident 1.0% fuel defect level - See Table 6 Secondary coolant iodine activity prior to accident 0.1 IC i/gm of DE 1-131 - See Table 15 Release Modeling Primary to secondary leakage to the faulted SG 0.3 gpm Primary to secondary leakage to the intact SGs 0.7 gpm Initial mass in faulted SG 6.459E7 gm Time to release initial mass in faulted SG 5 min Time to cool RCS below 212'F and stop releases 72 hrs from faulted SG Iodine chemical form after release to atmosphere Elemental 97%

Organic 3%

Particulate (cesium iodide) 0%

Steam releases to environment from intact SGs 0 - 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 402,000 Ibm 2 - 29 hours3.356481e-4 days 0.00806 hours 4.794974e-5 weeks 1.10345e-5 months 2,273,500 Ibm Page 60 of 69

Table 19 (Cont.): Assumptions Used for Steam Line Break Dose Analysis SG iodine water/steam partition coefficient Faulted SG 1.0 Intact SGs 0.01 RCS mass 1.960E8 gin Initial intact SG mass 3.193E7 gm/SG Control Room atmospheric dispersion (X/Q) factors Intact SG releases:

0 - 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 1.19E-3 sec/M 3 2 - 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 1.12E-3 sec/M3 8 - 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 5.59E-4 sec/M3 24 - 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months 4.27E-4 sec/M3 96 - 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months 3.35E-4 sec/M3 Control Room atmospheric dispersion (X/Q) factors Faulted SG releases:

0-2 hours 1.18E-3 sec/M3 2 - 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 1.06E-3 sec/M3 8-24 hours 5A2E-4 sec/m3 24 - 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months 4.09E-4 sec/M3 96 - 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months 3.27E-4 sec/M3 Page 61 of 69

Table 20: Assumptions Used for Small Break LOCA Dose Analysis Source Term Core activity See Table 5 Fraction of fuel rods in core that fail 100 (% of core)

Gap fractions Iodine 5% of core activity Noble Gas 5% of core activity Alkali Metals 5% of core activity Fraction of fuel melting 0% of core Radial peaking factor Not applied since the 100% of the core is involved.

Fraction of activity released from failed fuel (gap activity)

Containment leakage 100%

Primary to secondary leakage 100%

Reactor coolant noble gas and alkali metal activity prior 1.0% fuel defect level - See Table 6 to accident Reactor coolant iodine activity prior to accident 60.0 4C i/gm of DE I-131 - See Table 15 Secondary coolant iodine activity prior to accident 0.1 +/-Ci/gm of DE I-131 - See Table 15 Secondary coolant alkali metal activity prior to accident 10% of primary concentration Page 62 of 69

Table 20 (Cont.): Assumptions Used for Small Break LOCA Dose Analysis Containment Leakage Release Path Containment net free volume 2.61 E6 ft3 Containment leak rates 0 - 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 0.1 weight %/day

> 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 0.05 weight %/day Iodine chemical form in containment Elemental 4.85%

Organic 0.15%

Particulate (cesium iodide) 95%

Spray removal in containment Not Credited Aerosol removal by fan cooler unit filters Number of FCUs operating 3 FCU filtered flow rate 8000 cfm/FCU Filter efficiency 0.9 Time delay before filtration begins 60 seconds Sedimentation removal in containment (applicable only for the potential plant modification to delete spray additive and to add baskets of trisodium phosphate in the containment to adjust sump pH post-accident) 0.1 hr1' Particulate Iodine 0.1 hrf' Alkali metals Primary to Secondary Leakage Release Path Primary to secondary leakage 1.0 gpm total Steam release to environment 0 - 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 405,000 Ibm

>2 0.0 Ibm SG iodine water/steam partition coefficient 0.01 SG alkali metal water/steam partition coefficient 0.001 Iodine chemical form in after release to atmosphere Elemental 97%

Organic 3%

Particulate (cesium iodide) 0%

RCS mass 1.960E8 gm Page 63 of 69

Table 20 (Cont.): Assumptions Used for Small Break LOCA Dose Analysis Secondary Side mass h 3.193er7 gF/SG Control Room Atmospheric Dispersion Factors Control Room atmospheric dispersion (X/Q) factors Releases from containment surface('):

0 - 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 3.57E-4 sec/M 3 2 - 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 3.12E-4 sec/M3 8 - 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 1.24E4 sec/M3 24 - 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months 1.06E-4 sec/M 3 96 - 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months 7.99E-5 sec/M3 Control Room atmospheric dispersion (X/Q) factors Secondary releases(2):

0 - 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 1.19E-3 sec/M3 2 - 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 1.12E-3 sec/M3 8 - 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 5.59E-4 sec/M3 24 - 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months 4.27E-4 sec/M3 96 - 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months 3.35E-4 sec/M3 Notes:

1. Used for activity released via containment leakage.

2. Used for activity released via secondary releases.

Page 64 of 69

Table 21: Assumptions Used for Gas Decay Tank Rupture Dose Analysis Gas decay tank inventory 50,000 Ci of dose equivalent Xe-133 Duration of release 5 minutes Control Room atmospheric dispersion (X/Q) factors Releases from containment vent:

0 - 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 5.93E-4 sec/M 3 Page 65 of 69

Table 22: Assumptions Used for Volume Control Tank Rupture Dose Analysis Volume control tank inventory (Ci)

I-130 1.97E-2 1-131 6.29E-1 I-132 9.61E-1 1-133 1.14E0 I-134 2.33E-1 I-135 7.38E-1 Kr-85m 1.61E2 Kr-85 2.24E2 Kr-87 4.96EI Kr-88 2.40E2 Xe-131 m 3.95E2 Xe-133m 4.18E2 Xe-133 3.04E4 Xe-135m 7.54EI Xe-135 9.57E2 Xe-138 6.68E0 Reactor coolant iodine activity prior to accident 1.0 g+/-Ci/gm of DE I-131 - See Table 15 Reactor coolant noble gas activity prior to accident 1.0% fuel defect level - See Table 6 Iodine partition factor in VCT 0.01 Duration of release of VCT contents 5 minutes Letdown flow rate 132 gpm Operator action time to isolate the letdown line 30 minutes Letdown line demineralizer iodine DF 10 Airborne fraction for iodine in water released from 0.1 letdown line Control Room atmospheric dispersion (X/Q) factors Releases from containment vent:

0 - 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 5.93E-4 sec/m 3 Page 66 of 69

Table 23: Assumptions Used for Holdup Tank Failure Dose Analysis Holdup tank volume 8500 ft3 Holdup tank full level 80%

Reactor coolant iodine activity prior to accident 1.0 pCi/gm of DE I-131 - See Table 15 Reactor coolant noble gas activity prior to accident 1.0% fuel defect level - See Table 6 Letdown line demineralizer iodine DF 10 Time assumed to fill holdup tank 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Iodine partition coefficient for holdup tank liquid 0.01 Duration of release 5 minutes Control Room atmospheric dispersion (X/Q) factors Releases from containment vent:

0 - 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 5.93E-4 sec/M3 Page 67 of 69

Table 24: Assumptions Used for FHA Analysis Source Term Core Total Fission Product Activity (84 hrs decay) See Table 25 Number of Fuel Assemblies 193 Radial Peaking Factor 1.70 Fuel Rod Gap Fraction 1-131 12%

Kr-85 30%

Other Iodines and Noble Gases 10%

Fuel Damaged One assembly Time after Shutdown 84 hours9.722222e-4 days 0.0233 hours 1.388889e-4 weeks 3.1962e-5 months Water Depth 23 feet Overall Iodine Scrubbing Factor 200 Noble Gases Scrubbing Factor 1 Filter Efficiency No filtration of releases assumed Isolation of Release No isolation of releases assumed Time to Release All Activity 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Time to Start Crediting Emergency Control Room HVAC 24 minutes Control Room Atmospheric Dispersion (X/Q) Factor Containment Vent:

0 - 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 5.93E-4 sec/M3 Page 68 of 69

Table 25: Core Source Term 84 Hours after Shutdown Isotopic Inventory, curies Iodine 1-130 3.41E4 1-131 6.90E7 1-132 6.38E7 1-133 1.17E7 1-134 O.OOEO 1-135 2.63E4 Noble Gases Kr-85m 5.62EI Kr-85 1.11E6 Kr-87 O.OOEO Kr-88 O.OOEO Xe-131m 9.71E5 Xe-133m 2.78E6 Xe-133 1.36E8 Xe-135m 4.21E3 Xe-135 7.86E5 Xe-138 O.OOEO Page 69 of 69

v t e Letter Sequence Request
TAC:MC3351, (Open)
Initiation Request, Request Acceptance Supplement, Supplement, Supplement, Supplement Administration Questions 30 September 2004 6 December 2004 13 January 2005 Answers Results Approval... Other: ML050810189, ML050820327
MONTHYEAR2004-06-02 [Table View]

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