Text

Enclosure 8: Evaluation of Changes for AST RE-Analysis for an Increase in Fuel Burnup
	ML26029A407
Person / Time
Site:	Callaway
Issue date:	01/28/2026
From:	Devoe M; Ameren Missouri, Numerical Advisory Solutions, Union Electric Co
To:	; Office of Nuclear Reactor Regulation
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ML26029A397	List: ML26029A398; ML26029A402; ML26029A404; ML26029A405; ML26029A406; ULNRC-06922, Enclosure 8: Evaluation of Changes for AST RE-Analysis for an Increase in Fuel Burnup;
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	ULNRC-06922 NAS-2357-007
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ULNRC-06922 Page l of 91 ENCLOSURE 8 EVALUATION OF CHANGES FOR AST RE-ANALYSIS FOR AN INCREASE IN FUEL BURNUP The following pages provide the technical report provided by Numerical Advisory Solutions supporting this license amendment request.

NAS-2357-007, "Evaluation of Changes for AST Re-analysis for an Increase in Fuel Burnup,"

Revision 0, dated September 22, 2025 90 pages follow this cover sheet

NUMERICAL ADVISORY SOLUTIONS a Zachry Group company Numerical Advisory Solutions, LLC Report Release amr l TARR EAENLARTCERRERORESAC ESERIES RT TEV TEE PERESEAT Memon NAS Report Release Report Number: NAS-2357-007 Revision Number: 0

Title:

Evaluation of Changes for AST Re-analysis for an Increase in Fuel Burnup Client: Ameren Missouri

==

Description:==

This report provides the Evaluation of Changes (EOC) document for the Alternative Source Term (AST) re-analysis performed in support of implementing an upgraded fuel design and an incremental increase in the maximum allowable burnup for use by Ameren Missouri in preparing a License Amendment Request (LAR). The EOC document reflects the Numerical Advisory Solutions, LLC analysis methodology and results documented in the referenced calculations.

Nhichadl \, DeVoe 09/19/2025 Author Date Mike DeVoe Reser erman, 9-19-2025 Reviewer Date Roger Gorman

[Michal \. Dale 9/22/2025 Project Mana Date Mike DeVoe 9/22/25 NAS Manageme Date Scott Ingalls 200 Regency Forest Drive, Suite 330 Cary, NC 27518 919-465-7230 www.numerical.com

NUMERICAL Evaluation of Changes for AST NAS-2357-007 ADVISORY SOLUTIONS Re-analysis for an Increase in Revision 0 2 Zachry Group company Fuel Burnup Page 2 of 6 TABLE OF CONTENTS 1.

Introduction and Background Impact Assessment References Number of Pages in Main Body

- 6 Attachments No. of Pages Attachment A Evaluation of Changes Document LIST OF TABLES Table 1-1: AST Calculations and Reports Supporting an Increase in Fuel Burnup

NUMERICAL Evaluation of Changes for AST NAS-2357-007 ADVISORY SOLUTIONS Re-analysis for an Increase in Revision 0 a Zachry Group company Fuel Burnup Page 3 of 6

1. Introduction and Background Callaway is transitioning to an upgraded fuel design, adopting new analysis methods, and incrementally increasing the allowable fuel burnup. These changes implement a more robust fuel design incorporating accident tolerant features, improve nuclear fuel utilization, and support initiatives such as extended fuel cycles and increased operating flexibility.

This evaluation focuses on the impact of the upgraded fuel design and incremental increase in the allowable burnup on the current licensing basis (CLB) AST dose analysis.

Other changes associated with the overarching LAR (upgraded fuel design and new analysis methods) are addressed elsewhere. The following calculations and reports support the AST dose re-evaluation.

Table 1-1: AST Calculations and Reports Supporting an Increase in Fuel Burnup Subject Document Reference l

Number Data Input Request NAS-2357-001

[1]

Source Term NAS-2357-002 ss 2]

l LOCA/DBA Dose Calculation NAS-2357-003 Fuel Handling Accident in Fuel Handling Building NAS-2357-004 (FHB) Dose Calculation i

Fuel Handling Accident in Reactor Containment NAS-2357-005 l

. Building (RCB) Dose Calculation Transient Steam Release Determination

=

AMEREN-CP-001 l Steam Release Disposition

__. NAS-2357-010

2. Purpose This report provides an EOC document reflecting the AST dose re-analysis for the proposed changes affecting the CLB AST dose analysis and results. Revision 1 of RG 1.183 [9] was used in performing the re-analysis of those events affected by the proposed changes. Revision l of RG 1.183 provides, in part, guidance for addressing fuel burnups in excess of 62 GWD/MTU.

This report provides Enclosure 8 of the LAR which is to be developed by Ameren.

3. Methodology The methodology employed follows the methodology developed and documented in Section 3 of the EOC calculation [8] associated with the full implementation of RG 1.183, Rev. 0 which is the CLB [12] as described in the Callaway Final Safety Analysis Report (FSAR) Chapter 15 dose analysis. For the EOC developed here, only the items impacted by the fuel design change, increase in burnup, and the use of RG 1.183, Rev.

1 [9] are included.

NUMERICAL Evaluation of Changes for AST NAS-2357-007 ADVISORY SOLUTIONS Re-analysis for an Increase in Revision 0 a 2achry Group company Fuel Burnup Page 4 of 6 Below is an outline of the EOC report for the selective adoption of RG.183, Rev. 1. evaluated herein.

It is derived from the EOC outline for full implementation of AST [8] and reflects that the AST re-analysis EOC report will be Enclosure 8 of the planned LAR submittal. Ameren is responsible for the overarching LAR Cover Letter and other required Enclosures.

- Evaluation of Changes (AST)

Summary Description Detailed Description Technical Analysis 3.1 Atmospheric Dispersion Factors 3.2 Accident Source Terms 3.3 Dose Analysis 4.

References Attachment A RG 1.183 Conformance Section 3.3 is expanded as:

3.3 Dose Analysis 3.3.1 Introduction 3.3.2 Common Analysis Inputs and Assumptions 3.3.2.1 CR Model 3.3.2.2 TSC Model 3.3.2.3 Steam Release Calculations 3.3.2.4 Dose due to External Sources 3.3.3 LOCA 3.3.4 FHAs (FHB and RCB)

Sections 3.3.3 and 3.3.4 are expanded, using Section 3.3.3 as an example, as:

3.3.3 LOCA (event title) 3.3.3.1 Introduction 3.3.3.2 Input Parameters and Assumptions 3.3.3.2.1 Source Term 3.3.3.2.2 Release Models 3.3.3.2.3 Control Room 3.3.3.3 Acceptance Criteria 3.3.3.4 Results and Conclusions

NUMERICAL Evaluation of Changes for AST NAS-2357-007 o

ADVISORY SOLUTIONS Re-analysis for an Increase in Revision 0 a Zachry Group company Fuel Burnup Page 5 of 6

4. Inputs The information contained in References [1] through [7], RG 1.183, Rev.

1 (Reference [9]), the Callaway FSAR (Reference [10]), and the Callaway FSAR Site Addendum (Reference [11 ] was used as input to the EOC document to describe and document the analyses performed by NAS.

Section 4, References, of the EOC document lists additional sources of information cited in the EOC.

5. Assumptions None.

6. Analysis Attachment A provides the EOC document.

7. Results and Conclusions The EOC document has been developed and is included in Attachment A.

7.1 Precautions and Limitations None.

8. Impact Assessment The proposed changes to the FSAR associated with the EOC are to be developed later. This is acceptable as Ameren does not intend to include proposed FSAR markups with the LAR. The final impact assessment cannot be made until the overarching LAR is submitted, reviewed and approved by the NRC, and the SER is issued.

g; NUMERICAL Evaluation of Changes for AST NAS-2357-007 it ADVISORY SOLUTIONS

_Re-analysis for an Increase in Revision 0 a

Zachry Group company Fuel Burnup Page 6 of 6

9. References 1.

Calculation NAS-2357-001, Rev. 0, Analysis Inputs for Changes to AST Dose to Accommodate Increase in Fuel Burnup Calculation NAS-2357-002, Rev. 0, Callaway Radiological Analysis Core Source Term for Increase in Fuel Burnup Calculation NAS-2357-003, Rev. 0, Loss of Coolant Accident

- AST Dose Re-Analysis for Increase in Fuel Burnup Calculation NAS-2357-004, Rev. 0, AST Dose Re-analysis of a Fuel Handling Accident in the Fuel Handling Building for an Increase in Fuel Burnup Calculation NAS-2357-005, Rev. 0, AST Dose Re-analysis of a Fuel Handling Accident in the Reactor Containment Building for an Increase in Fuel Burnup Calculation AMEREN-CP-001, Rev. 2, AST Dose Calculation at the Callaway Energy Center, Option 3, Transient Steam Release Determination Calculation NAS-2357-010, Rev. 0, Steam Release Disposition Report NAI-1990-021, Rev 3, Callaway AST Evaluation of Changes Regulatory Guide 1.183, Revision 1, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, October 2023.

. Callaway Final Safety Analysis Report, Rev. OL-27, April 2024.

. Callaway Final Safety Analysis Report Site Addendum, Rev. OL-27, April 2024.

. Letter, USNRC to Ameren Missouri, Callaway Plant, Unit No.1 Issuance of Amendment No. 233 for Adoption of Alternative Source Term and Revision of Technical Specifications (EPID L-2021-LLA-0177), September 20, 2023, and enclosed Safety Evaluation (ML23166B088).

NUMERICAL Evaluation of Changes for NAS-2357-007 ADVISORY SOLUTIONS AST Re-analysis for an Revision 0 a Zachry Group company Increase in Fuel Burnup Att. A, 84 pages Attachment A Evaluation of Changes Document The Evaluation of Changes for AST Re-analysis for an Increase in Fuel Burnup follows this cover sheet.

Attachment A to the NAS-2357-007 Report Release consists of this cover sheet and the 83-page EOC document for a total of 84 pages.

NUMERICAL Attachment A to NAS-2357-007 NAS-2357-007 e"

ADVISORY SOLUTIONS

_Evaluation of Changes for AST Re-analysis Revision 0 aZachry Group company for an Increase in Fuel Burnup Page l of 48 Ameren Missouri (Union Electric Company)

Callaway Plant AST Re-analysis for an Increase in Fuel Burnup Evaluation of Changes

NUMERICAL Attachment A to NAS-2357-007 NAS-2357-007 ADVISORY SOLUTIONS Evaluation of Changes for AST Re-analysis Revision 0 a Zachry Group company for an Increase in Fuel Burnup Page 2 of 48 TABLE OF CONTENTS 1.

Summary Description 2.

Detailed Description 2.1 Proposed Changes 2.2.

Background and Introduction 3.

Technical Analysis 3.1 Atmospheric Dispersion Factors 3.1.1 Meteorological Data 3.1.2 Offsite Atmospheric Dispersion Factors 3.1.3.

Onsite Atmospheric Dispersion Factors 3.2.

Accident Source Terms 3.2.1 Fuel 3.2.2 3.2.3 3.3 Dose Analysis 3.3.1.

Introduction 3.3.2.

Common Analysis Inputs and Assumptions 3.3.3.

Loss-of-Coolant Accident (FSAR Section 15.6.5.4) 3.3.4 Fuel Handling Accidents (FSAR Section 15.7.4.5) 4.

References Number of Pages in Main Body -48 Attachments Number of Pages Attachment A

- Regulatory Guide 1.183 Rev. 1, Conformance Table Total Number of Pages in Report -83

NUMERICAL Attachment A to NAS-2357-007 NAS-2357-007 aApvisoRY SOLUTIONS Evaluation of Changes for AST Re-analysis Revision 0 a Zachry Group company for an Increase in Fuel Burnup Page 3 of 48 LIST OF FIGURES Figure 3.1. Onsite Release/Receptor Location Sketch LIST OF TABLES Table 3-1.

FHA_FHB Release/receptor pair Information Table 3-2. y/Q Factors for FHA_FHB Release/Receptor Pairs Table 3-3. FHA_RCB Release/receptor pair Information Table 3-4. y/Q Factors for FHA_RCB Release/Receptor Pairs Table 3-5. EOC Batch Burnups for Source Term Table 3-6. Fuel Source Term Table 3-7. Offsite Breathing Rates Table 3-8. Total Control Room Shine Dose Table 3-9.

Table 3-10. Core Inventory Fraction Released Into Containment Table 3-11. Core Inventory Released Into Containment in RADTRAD-NAI Model....

Table 3-12. Containment Spray Parameters for LOCA dose Table 3-13. Summary of Inputs for LOCA Table 3-14. Pool Removal Coefficient Inputs and Results FHB and RCB Table 3-15. Summary of Inputs for FHA_FHB Table 3-16. Summary of Inputs for FHA_RCB Table A-1. Conformance with Regulatory Guide 1.183 Rev.

1 Main Sections Table A-2. Conformance with Regulatory Guide 1.183 Rev.

1 Appendix A (Loss-of-Coolant Accident)

Table A-3. Conformance with Regulatory Guide 1.183 Rev.

1 Appendix B (Fuel Handling Accident)

NUMERICAL Attachment A to NAS-2357-007 NAS-2357-007 ADVISORY SOLUTIONS Evaluation of Changes for AST Re-analysis Revision 0 a Zachry Group company for an Increase in Fuel Burnup Page 4 of 48

1. Summary Description This evaluation supports a request to revise Renewed Facility Operating License NPF-30 for the Callaway Plant. Ameren Missouri (Union Electric Company) requests Nuclear Regulatory Commission (NRC) review and approval of proposed revisions to the licensing basis of the Callaway Plant that support transition to an upgraded fuel design, adoption of improved analysis methods, and an incremental increase in the allowable fuel burnup. These changes implement a more robust fuel design incorporating accident tolerant features, improve nuclear fuel utilization, and support initiatives such as extended fuel cycles and increased operating flexibility while maintaining the maximum fuel enrichment less than 5 w/o U-235.

This evaluation focuses on the impact of the upgraded fuel design and incremental increase in the allowable burnup on the current licensing basis (CLB) dose analysis. The CLB dose analysis, Reference [2], was approved September 20, 2023 and is a full implementation of the Alternative Source Term (AST) as described in Nuclear Regulatory Commission (NRC)

Regulatory Guide (RG) 1.183 Revision 0, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors. The CLB justifies rod-average burnups up to 62 GWD/MTU.

Revision l to RG 1.183 was issued October 2023 and provides AST analysis guidance for accident tolerant fuel and rod-average burnups up to 68 GWD/MTU for LOCA and non-LOCA events. This revision is used to analyze the Chapter 15 dose analyses affected by the proposed changes.

Ameren Missouri is not requesting any exceptions to RG 1.183 Rev.1, guidance for the events re-analyzed.

NUMERICAL Attachment A to NAS-2357-007 NAS-2357-007 js ADVISORY SOLUTIONS Evaluation of Changes for AST Re-analysis Revision 0 a Zachry Group company for an Increase in Fuel Burnup Page 5 of 48

2. Detailed Description 2.1 Proposed Changes Fuel Design Change Transition to a fuel product incorporating the ADOPT'pellet, AXIOM cladding, and PRIME'fuel assembly.

Justification for the Change:

The proposed change to the fuel design provides for the use of a more robust fuel design incorporating accident tolerant features.

Acceptability of the revised fuel design is demonstrated by the acceptable AST dose results.

Incremental Increase in Allowable Fuel Burnup The allowable burnup for a limited number of new fuel design assemblies loaded on the periphery of the core is increased from 62 GWD/MTU to the maximum value provided in Regulatory Guide 1.183 Rev. 1.

Justification for the Change:

Improves nuclear fuel utilization and supports initiatives such as extended fuel cycles and increased operating flexibility while maintaining the maximum fuel enrichment less than 5 w/o U-235.

Acceptability of the incremental burnup increase is demonstrated by the acceptable AST dose results.

AST Analyses A license amendment request to adopt AST analyses performed in accordance with the guidance in Regulatory Guide 1.183 Rev. 0, Reference [1], was submitted for review and approval September 28, 2021 and supplemented by letters dated December 1, 2021; July 5, 2022; September 1, 2022; December 8, 2022; and May 9, 2023. The license amendment was issued September 20, 2023, Reference [2] and justifies rod-average burnups up to 62 GWD/MTU.

To support the proposed changes considered herein, the affected AST analyses are re-performed in accordance with the guidance in Regulatory Guide 1.183 Rev. 1, Reference [3]. This revision provides guidance for accident tolerant fuel and burnups up to a maximum rod-average of 68 GWD/MTU for LOCA and non-LOCA events.

The FSAR Chapter 15 dose consequence analyses events listed below were re-analyzed for the proposed changes:

e Loss-of-Coolant Accident (FSAR Section 15.6.5.4) e Fuel Handling Accident (FSAR Section 15.7.4.5)

These are described in detail in Section 3.3 of this Enclosure.

NUMERICAL Attachment A to NAS-2357-007 NAS-2357-007 s

ADVISORY SOLUTIONS Evaluation of Changes for AST Re-analysis Revision 0 a Zachry Group company for an Increase in Fuel Burnup Page 6 of 48 The events listed below were not re-analyzed for the proposed changes:

Main Steam Line Break (FSAR Section 15.1.5.3)

Loss of AC Power (FSAR Section 15.2.6.3)

Locked Rotor (FSAR Section 15.3.3.3)

Rod Ejection (FSAR Section 15.4.8.3)

Letdown Line Break (FSAR Section 15.6.2.1)

Steam Generator Tube Rupture (FSAR Sections 15.6.3.1.3 and 15.6.3.2.3)

Justification for the Change:

Acceptability of the AST analysis is demonstrated by acceptable AST dose results.

Justification for not re-analyzing the events listed above for the changes being evaluated is:

e No fuel failures are predicted for the extended burnup core design for these events. The change in core radionuclide inventory due to the fuel design change and increased burnup will not be released.

The current licensing basis primary and secondary coolant radionuclide inventory remains applicable as it was established using the Technical Specification limits on primary coolant activity, and these have not changed. (See Section 3.2.2)

The change in steam release for non-LOCA events, due to thermal property changes associated with the fuel design change, is relatively small with negligible impact on dose.

(See Section 3.3.2.3)

There have been no Technical Specification, plant parameter, or operating procedure changes affecting the dose analysis of these events.

Technical Specification (TS) Changes There are no TS changes required to support the AST re-analysis documented herein. This analysis does support the TS change adding WCAP-18446-P-A, Incremental Extension of Burnup Limit for Westinghouse and Combustion Engineering Fuel Designs, to the list of approved COLR methodologies in TS 5.6.5b allowing fuel burnup beyond 62 GWd/MTU.

NUMERICAL Attachment A to NAS-2357-007 NAS-2357-007 ADVISORY SOLUTIONS Evaluation of Changes for AST Re-analysis Revision 0 a Zachry Group company for an Increase in Fuel Burnup Page 7 of 48 2.2 Background and Introduction The current Callaway Plant licensing basis for DBA analysis source terms is Regulatory Guideline (RG) 1.183 Rev. 0, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors.

Callaway is requesting Nuclear Regulatory Commission (NRC) review and approval of proposed revisions to the licensing basis that support transition to an upgraded fuel design, adoption of improved analysis methods, and an incremental increase in the allowable fuel burnup. These changes implement a more robust fuel design incorporating accident tolerant features, improve nuclear fuel utilization, and support initiatives such as extended fuel cycles and increased operating flexibility while maintaining the maximum fuel enrichment less than 5 w/o U-235.

This evaluation focuses on the impact of the upgraded fuel design and incremental increase in the allowable burnup on the current licensing basis (CLB) dose analysis. The CLB dose analysis justifies rod-average burnups up to 62 GWD/MTU.

Revision l to RG 1.183 was issued October 2023 and provides AST analysis guidance for accident tolerant fuel and rod-average burnups up to 68 GWD/MTU for LOCA and non-LOCA events. This revision is used to analyze the Chapter 15 dose analyses affected by the proposed changes.

Many aspects of the CLB are unchanged and are not re-evaluated as they have been previously reviewed and approved. These include:

e Site meteorological data remains applicable.

e Calculations of Dispersion factors (y/Q) for the original (release-receptor) pairs remain applicable. The FHA in the RCB changes the release location from a diffuse containment wall to a very large open hatch. With no Control Room isolation, the initial normal air intake remains applicable as the receptor throughout the duration of the accident. The same change in Control Room receptor location is applicable to the FHA in the FHB.

Initial primary and secondary reactor coolant system radionuclide inventories remain applicable.

Plant parameters (air flow rates, set points), operating procedures, and Technical Specifications modeled in the analysis remain applicable.

The containment Sump pH analysis remains applicable (no volume or concentration changes for sources of borated water).

Only analyses and events impacted by proposed changes are re-evaluated. This includes:

e The Core radionuclide inventory e

The Loss-of-Coolant Accident e

The Fuel Handling Accidents Regulatory Guide (RG) 1.183 Rev. 1 and Standard Review Plan Section (SRP) 15.0.1 were used by Numerical Advisory Solutions, LLC (NAS) in preparing the AST analyses. These documents

NUMERICAL Attachment A to NAS-2357-007 NAS-2357-007

ADVISORY SOLUTIONS Evaluation of Changes for AST Re-analysis Revision 0 a Zachry Group company for an Increase in Fuel Burnup Page 8 of 48 were prepared by the NRC staff to address the use of ASTs at current operating power reactors.

The RG establishes the parameters of an acceptable AST and identifies the significant attributes of an AST acceptable to the NRC staff. In this regard, the RG provides guidance to licensees for operating power reactors on acceptable applications for an AST; the scope, nature, and documentation of associated analyses and evaluations; consideration of impacts on risk; and acceptable radiological analysis assumptions. The SRP provides guidance to the staff on the review of AST submittals.

The AST is characterized by the composition and magnitude of the radioactive material, the chemical and physical form of the radionuclides, and the timing of the releases of these radionuclides.

The revised AST analyses performed for Callaway Plant are discussed in detail in Section 3 of this Enclosure. A comparison of the analysis to the regulatory guidance contained in RG 1.183 Rev. l is provided in Attachment A to this Enclosure.

Detailed listings of the parameters and assumptions used for the AST, as compared to the Current Licensing Basis (CLB), are presented in the event discussions contained in Section 3 of this Enclosure.

i NUMERICAL Attachment A to NAS-2357-007 NAS-2357-007 ADVISORY SOLUTIONS Evaluation of Changes for AST Re-analysis Revision 0 a Zachry Group company for an Increase in Fuel Burnup Page 9 of 48

3. Technical Analysis 3.1.

Atmospheric Dispersion Factors 3.1.1 Meteorological Data There is no change in meteorological data from that submitted, reviewed and approved for the CLB, Reference [2] and documented in Chapter 2.3 (Tables 2.3-87 through 2.3-95) of the Callaway FSAR Site Addendum.

No new METD cases were required for this submittal.

3.1.2 Offsite Atmospheric Dispersion Factors There is no change in the offsite atmospheric dispersion factors from those submitted, reviewed and approved for the CLB, Reference [2] and documented in Chapter 2.3 (Table 2.3-96) of the Callaway FSAR Site Addendum. The limiting offsite dispersion factors used in the CLB Chapter 15 dose analysis are presented in FSAR Table 15A-2.

No new PAVAN cases were required for this submittal.

To ensure a conservative dose analysis, Section 5.3 of RG 1.183 Rev. l provides an acceptable methodology to adjust the 7/Q versus time assumptions, when the maximum release does not coincide with the maximum ¥/Q, such that the most adverse release of radioactive materials to the environment occurs coincident with the period of most unfavorable atmospheric dispersion.

Such adjustments were made for the LOCA event. The Low Population Zone (LPZ) x/Q versus time assumptions are adjusted to ensure a conservative dose analysis. The Exclusion Area Boundary (EAB) y/Q values do not require adjustment as the dose analysis assumes a bounding time independent value.

3.1.3 Onsite Atmospheric Dispersion Factors 3.1.3.1 Introduction There is no change in the onsite atmospheric dispersion factors from those submitted, reviewed and approved for the CLB, Reference [2] and documented in Chapter 2.3 (Table 2.3-96) of the Callaway FSAR Site Addendum. The limiting offsite dispersion factors used in the CLB Chapter 15 dose analysis are presented in FSAR Table 15A-2.

No new ARCON96 cases were required for this submittal.

Although this analysis does not require the development of additional y /Q values or release-receptor pairs, the analysis does apply the existing CLB values in a different manner for several events:

e FHA_FHB: Changes in release location and control room isolation assumptions require application of existing y/Q values in a revised manner. See Section 3.1.3.2.

FHA_RCB: Changes in release location and control room isolation assumptions require application of existing y/Q values in a revised manner. See Section 3.1.3.3.

a NUMERICAL Attachment A to NAS-2357-007 NAS-2357-007 Pe s"

ADVISORY SOLUTIONS Evaluation of Changes for AST Re-analysis Revision 0 a Zachry Group company for an Increase in Fuel Burnup Page 10 of 48 LOCA: To ensure a conservative dose analysis, Section 5.3 of RG 1.183 Rev. l provides an acceptable methodology to adjust the y/Q versus time assumptions, when the maximum release does not coincide with the maximum ¥/Q, such that the most adverse release of radioactive materials to the environment occurs coincident with the period of most unfavorable atmospheric dispersion. Such adjustments were made for the MCR and TSC for the containment leakage, ECCS leakage, and RWST back-leakage LOCA dose Cases.

3.1.3.2 FHA_FHB Release/Receptor andx/Q Changes For reference, a sketch of the Callaway site layout depicting the onsite release/receptor locations is shown as Figure 3.1.

The FHA_FHB performed for this analysis does not credit isolation of the normal Control Room outside air intake. The release/receptor pair information is shown in Table 3-1. The corresponding 7/Q values, developed for the CLB, are shown in Table 3-2. The ¥/Q table is included here as the CLB assumed CR isolation and used only the 0-2 hr y/Q value.

3.1.3.3 FHA_RCB Release/Receptor andX/Q Changes The FHA_RCB performed for this analysis does not credit isolation of the normal Control Room outside air intake and revises the release point from the Containment as a diffuse release to the Emergency Personnel Access Hatch as a point release. The release/receptor pair information is shown in Table 3-3. The corresponding y/Q values, developed for the CLB, are shown in Table 3-4. The y/Q table is included here as the CLB assumed CR isolation and used y/Q values for a Containment (Diffuse) release.

NAS-2357-007 Revision 0 Page 11 of 48 NUMERICAL Attachment A to NAS-2357-007

, apvisory soLutions Evaluation of Changes for AST Re-analysis a Zachry Group company for an Increase in Fuel Burnup Figure 3.1. Onsite Release/Receptor Location Sketch.

(Plant North is 133° 33 28 counterclockwise from True North)

Control Buildinl Generat ee") Auxiliary Building Handling Building Waste Building SPNAYV SYNE Emergency Control Room Intake Normal Control Room Intake Unit Vent Stack RWST Vent FHB (Nearest Point to Receptor)

Closest ADV Vent Closet MSSV Vent Closest Main Steam Line (Nearest Point to Receptor)

Closest Feedwater Line (Nearest Point to Receptor)

Containment Maintenance Hatch Steam Jet Air Ejector (Nearest Point to Receptor)

Condenser (Nearest Point to Receptor)

Turbine Driven AFW Exhaust Vents Reactor Building Wall (Nearest Point to Receptor)

Technical Support Center (TSC)

Emergency Personnel Access Hatch

=

NUMERICAL Attachment A to NAS-2357-007 NAS-2357-007 ADVISORY SOLUTIONS Evaluation of Changes for AST Re-analysis Revision 0 "a

Zachry Group company for an Increase in Fuel Burnup Page 12 of 48 I

Table 3-1. FHA_FHB Release/receptor pair Information

. Release Point Receptor Horizontal Release Height Intake Height Direction Point Distance (m)

Above Plant

' Above Plant Lookingat Grade (m)

Grade (m)

- Source From l

' Receptor (° from

]

[

True North)

FHB Closest. l CB intake 2.

22.5 l

l Point (Normal)

Table 3-2. y/Q Factors for FHA_FHB Release/Receptor Pairs_

Release Point Receptor 0-2hry/Q 2-8hry/Q 8-24hry/Q 1-4dayx/Q 4-30 day 7/O Point (s/m?)

(s/m3)

_(s/m*)

(s/m?*)

(s/m?)

FHB Closest CB intake Point

_ (Normal) 2.23E-03 1.85E-03

+

7.49E-04 4.30E-04 3.17E-04 Note: The 72-hour decay time is added to the time scale in RADTRAD-NAI executions.

Table 3-3. FHA_RCB Release/receptorparr Information.

Release Point Receptor Horizontal Release Height

Grade (m)

Above Plant Source From Grade(m)

_ Receptor (° from.

True North)

Emergency

- CB intake 43 22.5 338 Personnel (Normal)

Access Hatch l

Table 3-4. y/Q Factors for FHA_RCB Release/Receptor Pairs

_l Receptor l! 0-2hry/Q 2-8hry/O l 8-24hrx/Q 1-4dayy7/O l 4-30 day y/O Point

=

(sim)

Sl (s/m*)

l (sim)

(s/*)

(s/m?)

Emergency

' CB intake

)

Personnel

(Normal) 2.66E-03 2.43E-03 1.01E-03 l

5.57E-04 4.48E-04 l Access Hatch l

Release Point Note: The 72-hour decay time is added to the time scale in RADTRAD-NAI executions.

NUMERICAL Attachment A to NAS-2357-007 NAS-2357-007 ADVISORY SOLUTIONS Evaluation of Changes for AST Re-analysis Revision 0 a Zachry Group company for an Increase in Fuel Burnup Page 13 of 48 3.2.

Accident Source Terms 3.2.1 Fuel The calculation of the core fission product employs ORIGEN-S of the Oak Ridge National Laboratory (ORNL) SCALE 6.1.3 code package (Reference [4]). ORIGEN-S is an isotopic depletion and decay code which allows the user to specify fuel type, enrichment and periods of irradiation/decay and uses the latest cross-section data from ORNL to determine the existing nuclide inventory at specified intervals.

The inventory of radionuclides in the reactor fuel is based on a core power level of 3636 MWt (3565 MWt plus 2% postulated calorimetric uncertainty) with an operating period of 568.6 EFPD, representative of anticipated operating cycles at Callaway Plant. Decay occurring during refueling outages is conservatively neglected. Fuel assembly average enrichments up to and including 4.95 w/o U-235 and burnups up to the maximum value provided in RG 1.183, Rev. 1 are enveloped. Six fuel batches are considered:

The Batch 1 assemblies operated at an average power of 58.92 MW/MTU for 511.1 EFPD, 45.49 MW/MTU for 511.0 EFPD, and 13.81 MW/MTU for 568.6 EFPD.

The Batch 2 assemblies operated at an average power of 53.14 MW/MTU for 511.1 EFPD, 46.49 MW/MTU for 511.0 EFPD, and 14.43 MW/MTU for 568.6 EFPD.

The Batch 3 assemblies operated at an average power of 54.29 for 511.0 EFPD and 41.86 MW/MTU for 568.6 EFPD.

The Batch 4 assemblies operated at an average power of 51.73 MW/MTU for 568.6 EFPD.

The Batch 5 assemblies operated at an average power of 47.51 MW/MTU for 568.6 EFPD.

The Batch 6 assemblies operated at an average power of 119.59 MW/MTU for 568.6 EFPD.

The total burnup (MWD/MTU) of Batches l through6 at the end of the enveloping cycle are shown in Table 3-5.

Table 3-5. EOC Batch Burnups for Source Term Barc l Number of l

Burnup assemblies (MWd/MTU 4

60,749 28

<<59, 115 76 151,547 17 29.414 60, 27,014 l 6

.- i8

,68,000 __

The ORIGEN-GS results are presented in Table 3-6 and are based on end of cycle activities.

Attachment A to NAS-2357-007 NAS-2357-007 NUMERICAL ADVISORYSOLUTIONS Evaluation of Changes for AST Re-analysis Revision 0 a Zachry Group company for an Increase in Fuel Burnup Page 14 of 48 For comparison, the Fuel source term for the CLB dose analysis (FSAR Table 15A-3) is also presented.

Table 3-6. Fuel Source Term l

Extended CLB EB/CLB Isotope Burnup (EB)

Core Activity Activity Core Activity.

(Ci)

Ratio oo (Ci) l Kr-85 1.119E+06 9.677E+05 1.16 Kr-85m 2.431E+07 2.469E+07 0.98 Kr-87 =ss<<4<<783E+07 4.866E+07

=>

0.98 Kr-88 l

6.390E+07 6.507E+07 s<<O0<<.98 Rb-86

=.2.133E#05 l =.834EH0S 1.16 Sr-89

[

9.062E+07 9.252E+07 ss<<0<<.98 Sr-90 8.410E+06

° 7.220E+06

<<1.16 Sr-91 1.133E+08 s<<dLISTE+08)=s(ié<<iSB Ct Sr-92 1.218E+08 1.235E+08 0.99l Y-90 9.095E+06 7.816E+06 1.16 Y-91 1.193E+08 1.214E+08 <<

0.98 Y-92

"-1.233E+08 1.250E+08 l

0.99 Y-93 l

1.400E+08 1.416E+08 0.99 Zr-95 1.636E+08 1.651E+08 l

Zr-97

~~

1.644E+08 1.651E+08.

Nb-95 >

<<1.644E+08 1.659E+08

~

Mo-99 i

1,809E+08 1.811E+08 Tc-99m 1.603E+08 1.603E+08 Ru-103

~

-1,557E+08 1.541E+08 Ru-105 1.

100E+08 1.080E+08 Ru-106 5.363E+07 4.835E+07 l Rh-105 9.860E+07 9.707E+07 Sb-127 "8.

958E+06 8.907E+06 Sb-129.

2.831 E+07 2.816E+07 Te-127 8781E+06

° 8.717E+06 Te-127m 1

,468E+06 1.443E+06 C

1.02 Te-129 2.605E+07 2.590E+07 l

Te-129m 5.005E+06

~

4.97 1 E+06

~

Te-131m

-1.889E+07

"" 1_886E+07

-1389E+08

1.390E+08 9.758E+07

=l = 9.758E+0O7 L.414E+08

<<d1.414EE+08

a NUMERICAL Attachment A to NAS-2357-007 NAS-2357-007 ZINE.

ADVISORY SOLUTIONS _ Evaluation of Changes for AST Re-analysis Revision 0 a Zachry Group company for an Increase in Fuel Burnu Page 15 of 48 p

8

T f

Table 3-6. Fuel Source Term Extended l

CLB

-EB/CLB Isotope Burnup (EB)

Core Activity Activity Core Activity (Ci) l Ratio (Ci) 1-133 1.991E+08 1.996E+08 1

1.00 1-134 2.234E+08 2.240E+08 1.00 1-135 1.891E+08 1.893E+08 1.09 Xe-133.

1.991E+08 1.995E+08 l

1.00 Xe-135 4.886E+07 4.704E+07 1.04 Cs-134 1.778E+07 1405E+07 97 Cs-136 l

5.212E+06 4.531E+06

<<1.45 Cs-137 1.189E+07 7

1.008E+07 l

1.18 L_

d ad Ba-139

~-1.801E+08 1.807E+08 1.00 Ba-140 l

1.708E+08 1.715E+08 1.00 La-140

~

1.776E+08 1.777E+08 1.00 Lal4) 1.587E+08 1.594E+08 1.00 La-142 1.507E+08 1.515E+08 0.99 Ce-141 1.613E+08 1.620E+08 1.00 Ce-143 l

1.483E+08 1.492E+08 0.99 Ce-144 j

(l1:264E+08

=

s1.213E+08 sg.

4 Pr-143 1.452E+08 1.460E+08 0.99 Nd-147 6.412E+07 6.421 E407.. 1.00 Np-239 L.9ISE+09

=

1.907 +09 1.00 Pu-238 3.613E+05 2.539E+05 4

Pu-239 3.001E+04 2.836E+04_

1.06 Pu-240

~

4.327E+04 3.890E+04 Ll Pu-24]

, ~ :1.299E+07 l

LI7IE+O7 oT.

Am-24]

1.414E+04 1.130E+04 ds Cm-242.

3.853E+06

=:

3.1288 +06 1.23. ° Cm-244 5.081E+05 2.900E+05 1.75 1-130

[__

2.085E+06 L809E406 l

145 5 Kr-83m 1.1S1E+07 1.164E+07 0.99 Xe-138 1.690E+08 1.697E+08 ~~

1.90 Xe-131m 1.309E+06 1.280E+06 1.02

~

Xe-133m 6.202E+06 6.204E+06 1.00 Xe-135m 4.183E+07

=

-1.785E+08 7.00 Cs-134m 4.567E+06

4.054E+06 11300:

Rb-88 i

6.512E+07 6.626E+07 0.98

Bi a

bic NUMERICAL

- GPQE) ="

ADVISORY SOLUTIONS 4

aZachry Group company l

Isotope Rb-89 Sb-124 Sb-125 Sb-126 Te-131 Te-133 Te-134 Te-125m Te-133m Ba-141 Ba-137m_

Pd-109 Rh-106 Rh-103m Tc-101 Eu-154 Eu-155 Eu-156 La-143.

Nb-97 l

Nb-95m Pm-147 Pm-148 Pm-149 Pm-151 Pm-148m Pr-144 Pr-144m Sm-153 Y-94 Y-95 Y-91m Br-82 Br-83 Br-84 Attachment A to NAS-2357-007 Evaluation of Changes for AST Re-analysis for an Increase in Fuel Burnup Table 3-6. Fuel Source Term 7

l L

Extended Burnup (EB)

Core Activity (Ci) 8.548E+07 8.408E+04 8.466E+05 4.920E+04 8.269E+07 1.069E+08 1.763E+08 1.856E+05 9.349E+07 1.580E+08 1.130E+07 3.416E+07 6.100E+07 t

+

1.557E+08 1.681E+08 7.296E+05

~

2.942E+05 2.592E+07 1.465E+08 1.656E+08 1.876E+06 1.663E+07 1.760E+07 6.058E+07 1.902E+07 3.836E+06 1.273E+08 1.770E+06 "4.889E+07 1.478E+08 1.578E+08 6.561E+07

~

3.510E+05 1.139E+07 2.062E+07 Lo.

ap CLB (Ci) 8.706E+07 7.037E+04 TAI8E+05 4.691E+04 8.272E+07 1.0710E+08 1.772E+08 1.590E+05 9.382E+07 1.587E+08 9.583E+06 3.214E+07 5.544E+07 1.540E+08 1.680E+08 5.726E+05 2.368E+05

" 2.182E+07 1.475E+08 1.662E+08 1.894E+06 1.575E+07

~

1,720E+07 5.969E+07 1.881E+07 3.619E+06 1.222E+08 1.699E+06 4.468E+07 1.493E+08 1.591 E+08 6.664E+07 3.049E-+05 1.154E+07 2.094E+07 l

EB/CLB Core Activity Activity Ratio 0.98 1.19 1.14

~

1.05 1.00 1.00

=l 1.00 1.17 1.00 l 1.00 1.18 1.06 1.10 1.01 1.00 127l 1.24 l

119 0.99 1.00 0.99 1.06 1.02 1.01 1.01 1.06 1.04 1.04 L

J 109 0.99 0.98 0.99 l

lL LIS, 0.99 0.98 NAS-2357-007 Revision 0 Page 16 of 48

NUMERICAL Attachment A to NAS-2357-007 NAS-2357-007

ADVISORY SOLUTIONS Evaluation of Changes for AST Re-analysis Revision 0 a

Zachry Group company for an Increase in Fuel Burnup Page 17 of 48 Table 3-6. Fuel Source Term "Extended CLB EB/CLB© Isotope Burnup (EB)

Core Activity Activity Core Activity (Ci)

Ratio (Ci)

Am-242 7.194E+06 l

6.242E+06 1.15 Np-238 4.739E+07 3.962E+07 1.20 Pu-243 4.940E+07 4.014E+07 1.23 3.2.2 RCS The initial RCS inventory represents steady state operation with 1% fuel defects and Dose Equivalent I-131 and Dose Equivalent Xe-133 at the existing TS 3.4.16 limits of 1.0 wCi/gm and 225 wCi/gm, respectively. The changes being evaluated do not affect this inventory and the CLB RCS source term, shown in FSAR Table 15A-5, remains applicable.

3.2.3 FHA Phase 2 Protracted Release Stable Iodine Concentration Section B.3 of RG 1.183 Rev.1 provides a calculation sequence for determining the evolution removal coefficient for a pool as a function of the stable iodine concentration, pH, surface area, and volume. See section 3.3.4.2.1.2.

NUMERICAL Attachment A to NAS-2357-007 NAS-2357-007 ADVISORY SOLUTIONS Evaluation of Changes for AST Re-analysis Revision 0 a Zachry Group company for an Increase in Fuel Burnup Page 18 of 48 3.3.

Dose Analysis 3.3.1 Introduction The licensing basis for the radiological consequences analyses for Chapter 15 of the Callaway Plant FSAR is currently based on methodologies and assumptions that are derived from RG 1.183 Rev. 0 (Reference [1]) To support the proposed changes addressed in this amendment, RG 1.183 Rev.

1 (Reference [3_) is being used to calculate the offsite, control room, and TSC radiological consequences for the dose analysis affected by the changes. RG 1.183 Rev.

1 provides AST analysis guidance for accident tolerant fuel and rod-average burnups up to 68 GWD/MTU for LOCA and non-LOCA events. The following accidents are re-analyzed:

e Loss-of-coolant accident (LOCA) e Fuel handling accident (FHA)

Each accident and the specific input assumptions are described in detail in the following sections.

The analyses are performed using Version 1.3 (QA) of the RADTRAD-NAI computer code.

RADTRAD-NAI is described in Reference [5] and is qualified and maintained under a 1lOCFR50 Appendix B Quality Assurance program by Numerical Advisory Solutions, LLC. RADTRAD-NAI implements the dose analysis models and methodology described in NUREG/CR-6604 (Reference [6]) and its supplements (References [7] and [8]). The calculation models in RADTRAD-NAI are consistent with those outlined in RG 1.183 Rev. 1.

Technical Specification Amendment 223, Refence [2], approved the use of the RADTRAD-NAI computer code and model for Callaway.

3.3.2 Common Analysis Inputs andAssumptions The inputs and assumptions in this section are common to all analyses discussed in Section 3.3 of this Enclosure. The accident-specific inputs and assumptions are discussed in Sections 3.3.3 through 3.3.4.

The total effective dose equivalent (TEDE) doses were determined at the EAB for the worst 2-hour interval. The TEDE doses at the LPZ, the technical support center, and main control room are determined for the duration of the event (30 days).

The TEDE dose is equivalent to the committed effective dose equivalent (CEDE) from inhalation plus 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 Rev.1 guidance. The dose conversion factors (DCFs) used in determining the CEDE dose are from Table 2.1 of Federal Guidance Report (FGR) No. 11 (Reference [9]) and the DCFs used in determining the EDE dose are from Table IHI.1 of FGR No. 12 (Reference [10]). DCFs used in determining the EDE dose specific to ground deposition are from Table III.3 of FGR No. 12 (Reference [10]).

This analysis uses the CLB DCFs which are listed in FSAR Table 15A-4.

NUMERICAL Attachment A to NAS-2357-007 NAS-2357-007

@ apvisory sotuTions Evaluation of Changes for AST Re-analysis Revision 0 a Zachry Group company for an Increase in Fuel Burnup Page 19 of 48 The breathing rates used in the offsite dose calculations are provided in Table 3-7 and the atmospheric dispersion factors are discussed in Section 3.1.2 of this Enclosure. These are identical to the CLB values listed in FSAR Table 15A-1. However, to incorporate guidance presented in Section 5.3 of RG 1.183 Rev1. and ensure a conservative dose analysis, the application of breathing rates and dispersion factors is modified, as necessary, for the LOCA dose evaluation. See Table 3-13.

Consistent with the CLB, no credit is taken for the radioactive decay during release and transport or for cloud depletion by ground deposition during transport to the control room, EAB, or outer boundary of the LPZ.

The reactor coolant activity is based on operation with 1% fuel defects. The core and coolant activities are based on a core power of 3636 MWt, which is the nominal core power of 3565 MWt plus a 2% calorimetric uncertainty. The core activities, reflecting the proposed changes, and coolant activities present in the primary and secondary systems during normal plant operation are listed in Table 3-6 of this Enclosure and discussed in Section 3.2.2. Table 3-6 also compares the revised core activities to the CLB analysis values listed in FSAR Table 15A-3. As discussed in Section 3.2.2, there is no change to the coolant activities present in the primary and secondary systems listed in FSAR Table 15A-5.

Table 3-7. Offsite Breathing Rates an l

Time a

l Offsite Breathing Rate(m?/sec) 0 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 3.5E-04 824 hours0.00954 days 0.229 hours 0.00136 weeks 3.13532e-4 months 1.8E-04

> 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months

_-2.3E-04

NUMERICAL Attachment A to NAS-2357-007 NAS-2357-007 s*

ADVISORY SOLUTIONS Evaluation of Changes for AST Re-analysis Revision 0 a Zachry Group company for an Increase in Fuel Burnup Page 20 of 48 3.3.2.1 Control Room Model The RADTRAD-NAI AST dose evaluation model used in this analysis is unchanged from that reviewed and approved for the CLB evaluation (Reference [2]). Input values are revised to reflect updated analysis assumptions and regulatory guidance. These revisions are discussed in Sections 3.3.3 and 3.3.4 for the LOCA and FHA events, respectively.

The unified RADTRAD-NAI AST dose evaluation model is included in the CLB as FSAR Figures 15A-1 and 15A-3.

It consists of 19 compartments and 43 flow paths. The unified model integrates the complex Callaway Control Room, Control Building, Control Building Equipment Room and their interconnected HVAC flow paths into one model. This model allows for time dependent HVAC alignment changes, as well as providing for explicit treatment of mixing volumes, filter placement within flow paths, and direct incorporation of the results of Callaway specific Control Room (CR)/Control Building (CB)/Control Building Equipment Room (CBER) inleakage measurements. The model also consolidates RADTRAD-NAI models for each of the Chapter 15 dose events into one model that can be used, with appropriate event specific inputs, for all events.

The control room portion of the Unified model is shown in detail for the CLB as FSAR Figure 15A-3. The AST analysis supports maximum values of 60 cfm for the Control Room inleakage (Qcr), 6000 cfm for the Control Building inleakage (Qcg), and 300 cfm for the Control Building Equipment Room inleakage (QcBer) with the restriction, imposed by the LOCA analysis, that Qcs < 6000 100

Qcr as shown in FSAR Figure 15A-2 for the CLB.

As this re-analysis for the FHA_FHB and FHA_RCB accidents do not credit Control Room isolation, FSAR Figure 15A-2 is not applicable to these accidents.

The Control Room and Control Building flow path filter removal efficiencies are unchanged from the CLB. The filter removal efficiencies for Flow Path 16 reflect not crediting the Control Building Pressurization System charcoal adsorbers in the AST analysis.

At the start of all the events considered, the control room ventilation system is in normal mode.

In this mode, unfiltered air from the environment enters the control building and the control room. Receipt of a safety injection (SI) actuation signal or a high radiation signal from the control room air intake monitors will isolate the control room and initiate the emergency mode of operation, including a delay.

After emergency mode is initiated, outside air is brought into the control building through safety grade filters. Makeup air is brought into the control room via both trains of the control room filtration system, which draws in air from the control building. Unfiltered air also leaks into the control building, control room equipment room, and the control room equipment room via assumed inleakage rates described above. In addition, a filtered recirculation flow is modeled for the control room during emergency mode operation. Air in the control room and control building is discharged at flow rates to match the total inflow to the compartments.

For events that rely solely on the control room air intake monitors for control room isolation, the unfiltered inleakage to the control room will continue to be associated with the normal mode air

=

NUMERICAL Attachment A to NAS-2357-007 NAS-2357-007 ADVISORY SOLUTIONS Evaluation of Changes for AST Re-analysis Revision 0 a Zachry Group company for an Increase in Fuel Burnup Page 21 of 48 intake. This results in the modeling of the control room unfiltered inleakage with the normal mode atmospheric dispersion factors until SI actuation signal completes the control room isolation.

The control room ventilation isolation signal starts both trains of the control room filtration system. However, a failure of one of the filtration fans is assumed at the start of emergency mode anda larger unfiltered inflow to the control room is assumed since only half of the makeup flow to the control room can pass througha filter. After a defined time of 30 minutes, operator action isolates the failed train and reduces the unfiltered inflow to the control room and consequently lowers the filtered inflow to the control building.

The unfiltered air inleakage to the control room, control building, and control room equipment rooms described above are conservatively analyzed for each of the events discussed in Sections 3.3.3 through 3.3.4 of this Enclosure. The results of these analyses demonstrate that the applicable regulatory control room dose limit is met for each event.

NUMERICAL Attachment A to NAS-2357-007 NAS-2357-007 ADVISORY SOLUTIONS Evaluation of Changes for AST Re-analysis Revision 0 a Zachry Group company for an Increase in Fuel Burnup Page 22 of 48 3.3.2.2 Technical Support Center Model FSAR Figure 15A-1 illustrates how the TSC is incorporated in the RADTRAD-NAI Unified model for the CLB. There are no changes to TSC input parameters for this re-analysis.

At the start of all the events considered, the TSC ventilation system is in normal mode. In this mode, unfiltered air from the environment enters the TSC. After emergency mode is manually initiated, outside makeup air is brought into the TSC through safety grade filters. Unfiltered air also leaks into the TSC via an assumed inleakage rate. In addition, a filtered recirculation flow is modeled during emergency mode operation. Air is discharged at a flow rate to match the total inflow to the compartment.

3.3.2.3 Steam Release Calculations In support of the CLB dose analyses, Ameren chose to develop new steam releases for each of the Callaway Plant FSAR Chapter 15 steam release transient analyses requiring an AST dose analysis. These secondary system steam releases were calculated using the RETRAN-3D thermal-hydraulic system code (Reference [11]). RETRAN-3D is classified by the NRC as safety-related software and is maintained and run under a 10CFR50 Appendix B program.

RETRAN-3D input decks for each of the various Chapter 15 analyses were provided by Ameren and then modified as appropriate to implement a steam release calculation in accordance with assumptions outlined in NRC Regulatory Guide 1.183 for the following accident scenarios:

Main Steamline Break, Loss of AC Power, Locked Rotor, Rod Ejection, Steam Generator Tube Rupture with a Failed Open ASD, and Steam Generator Tube Rupture, Overfill case.

The RETRAN-3D Steam releases served as input to the CLB AST RADTRAD-NAI dose analyses.

In support of proposed changes evaluated herein, the RETRAN-3D steam releases were updated to account for the fuel thermal properties of the new fuel design and RETRAN-3D error corrections. The updated results showed relatively minor changes in the steam release. Dose sensitivity studies have shown these minor changes have negligible impact on the dose results.

Therefore, continued use of the CLB steam releases for the above non-LOCA events isjustified.

NUMERICAL Attachment A to NAS-2357-007 NAS-2357-007 ADVISORY SOLUTIONS Evaluation of Changes for AST Re-analysis Revision 0 a Zachry Group company for an Increase in Fuel Burnup Page 23 of 48 3.3.2.4 Dose Due to External Sources From Section 4.2.2 of Reference [3], the radioactive material releases and radiation levels used in control room dose analysis should be determined using the same source term, transport, and release assumptions used for determining the EAB and LPZ TEDE values, unless these assumptions would result in non-conservative results for the control room.

The approach bases the source term on the LOCA Containment leakage cases and uses the MicroShield computer code, Reference [12], for the shielding calculation to determine shine doses. The software dedication within the applicable Numerical Advisory Solutions, LLC quality assurance plan is Reference [13]. The shine doses calculated for the LOCA event are conservatively applied to all Chapter 15 AST dose events.

Shine doses as a function of time are determined utilizing the output of RADTRAD-NAI. The RADTRAD-NAI models generated activity in curies for each isotope as a function of time and location. MicroShield does not perform time dependent analysis; therefore, discrete time edits at reasonable intervals were selected from the RADTRAD-NAI output files to run in each MicroShield model. The resultant exposure rates (with buildup) for each time interval were then used to calculate a cumulative dose for the accident by multiplying the rate by the length of the time interval. For conservative reasons, only the time periods between 0 and 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months are utilized. The shine dose rate at 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months is assumed constant for the remainder ofthe accident.

3.3.2.4.1 Control Room Dose due to External Sources From Section 4.2.1 and 4.2.3 of Reference [3], dose rates in the control room should also consider external sources as radiation shine from:

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

Three MicroShield models were developed for determination of the shine dose to the Control Room. These models analyzed shine from the containment, shine from the radioactive plume/cloud, and shine from radiological air filters. The cumulative shine dose to the Control Room, 1.91E-02, is the sum of these three components as shown in Table 3-8.

Table 3-8. Total Control Room Shine Dose CR Shine Component l

Dose (rem)

_ Containment 5.76E-03

_ Cloud 2,17E-03_

Filter 1.12E-02 Total 1.91E-02

NUMERICAL Attachment A to NAS-2357-007 NAS-2357-007 s"

ADVISORY SOLUTIONS Evaluation of Changes for AST Re-analysis Revision 0 a Zachry Group company for an Increase in Fuel Burnup Page 24 of 48 3.3.2.4.2 TSC Dose due to External Sources Dose rates in the TSC should also consider external sources as radiation shine from:

radioactive material in the reactor containment, and the external radioactive plume released from the facility.

A MicroShield model was developed for determination of the shine dose to the TSC. This model analyzed shine from the radioactive plume/cloud. The containment shine component of the TSC dose is ignored for the Callaway analysis. The TSC is heavily shielded from Containment by the Unit 1 Turbine Building making any containment shine contribution negligible compared to the cloud shine component. The cumulative shine dose to the TSC due to radioactive plume/cloud is 1.78E-01 rem.

3.3.2.5 Dose Due to Control Room Access The CLB, Reference [2] and FSAR Table 15.6-8, includes an evaluation of LOCA transit dose.

Although Callaway did not previously include transit dose in the licensing basis, the NRC staff identified this as a sufficiency item they expected to see during the pre-submittal meeting for implementation of AST.

Since that time, the industry has pushed back on the issue. During review of DG-1389, the predecessor to RG 1.183 Rev. 1, the PWROG commented/proposed in Reference [16] that the statement in Section 4.2 of the draft guidance discussing transit dose be removed as transit dose is not required to be explicitly considered in control room design or habitability, only a small subset of licensees include transit dose in their Chapter 15 control room doses, and there is no definitive guidance for calculating transit doses with a supporting basis for removal that control room doses should be limited to doses accrued while occupying the control room.

RG 1.183 Rev.

1 was issued without the transit dose discussion from the DG. Ameren interprets this as tacit NRC agreement that LOCA transit dose calculations are not required, and Ameren has chosen to not address control room transit dose for this submittal.

NUMERICAL Attachment A to NAS-2357-007 NAS-2357-007 ADVISORY SOLUTIONS Evaluation of Changes for AST Re-analysis Revision 0 a Zachry Group company for an Increase in Fuel Burnup Page 25 of 48 3.3.3 Loss-of-Coolant Accident (FSAR Section 15.6.5.4) 3.3.3.1 Introduction An abrupt failure of the main reactor coolant pipe is assumed to occur. Activity from the RCS is released to containment anda portion of this activity is released to the atmosphere via the mini-purge system prior to containment isolation. 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 is released to the environment by means of containment leakage. In addition, once recirculation of the emergency core cooling system (ECCS) is established, iodine activity in the sump solution may be released to the environment by means of leakage from ESF equipment outside containment in the auxiliary building, and by means of leakage from the ESF to the RWST with subsequent leaking or venting. The total offsite doses are the sum of the doses resulting from the four postulated release paths. The total onsite doses are the sum of the doses resulting from the four postulated release paths plus the dose due to external shine.

3.3.3.2 Input Parameters andAssumptions The analysis of the LOCA radiological consequences uses the analytical methods and assumptions outlined in RG 1.183 Rev. 1, Appendix A, Reference [3]. A summary of input parameters and assumptions is provided in Table 3-13. A listing of the CLB parameters is also included in Table 3-13 for comparison to the AST parameters. Conformance with the Main Body and Appendix A of RG 1.183 Rev. l is shown in Attachment A, Sections A.1 and A.2, respectively.

3.3.3.2.1 Source Term The iodine activity in the RCS at the time of the accident is assumed to be at the TS limit of 1.0 uCi/gm of DE I-131 for the maximum equilibrium RCS concentration. The noble gas activity concentration in the RCS at the time of the accident is assumed to be at the TS limit of 225 uCi/gm of DE Xe-133 for the maximum equilibrium RCS concentration. The activity for the remaining nuclide groups is at a 1% fuel defect level. The RCS activities are listed in FSAR Table 15A-5. These values are used in the containment purge release pathway.

For modeling the containment leakage, ECCS leakage, and RWST back-leakage release pathways, the analysis assumes fuel melt occurs in the entire core and the release of activity occurs over a 4.73 hour8.449074e-4 days 0.0203 hours 1.207011e-4 weeks 2.77765e-5 months interval as shown in Table 3-9. The gap release phase occurs at 30 seconds and ends in the first 0.23 hour2.662037e-4 days 0.00639 hours 3.80291e-5 weeks 8.7515e-6 months and the release from the melted fuel occurs over the subsequent 4.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months . This modeling is consistent with RG 1.183 Rev. 1, Table 5.

Section 3.2.1 of this Enclosure describes the development of the core fuel activities used in this analysis, which are listed in Table 3-6 The gap and fuel melt activity release fractions for the various nuclide groups are listed in Table 3-10 which was developed from Tables 2 and 6 of RG 1.183 Rev. 1.

=

NUMERICAL Attachment A to NAS-2357-007 NAS-2357-007 w

ADVISORY SOLUTIONS Evaluation of Changes for AST Re-analysis Revision 0 a

Zachry Group company for an Increase in Fuel Burnup Page 26 of 48 Table 3-9. LOCA Release Phases: RG 1.183 Rev. 1 Phase l

Onset

_ Duration Gap release 0.5 minutes=

0.23 hours2.662037e-4 days 0.00639 hours 3.80291e-5 weeks 8.7515e-6 months l

l. 0.00833 hours l

l Early In-vessel i

0.23 hours2.662037e-4 days 0.00639 hours 3.80291e-5 weeks 8.7515e-6 months .

4.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months Table 3-10. Core Inventory Fraction Released Into Containment Gap Release Early In-vessel Phase

_ Phase Noble gases 0.022 0.94

_ Xe, Kr Halogens l 0.007

_ 0.37 I, Br Alkali metals 0.005 0.23

Cs, Rb Tellurium metals l 0.007 (030 Te, Sb, Se Barium, Strontium ! 1.4x10° l 0.004 l Ba, Sr

Noble metals 0.00 0.006

_ Ru, Rh, Pd, Co Cerium group

- 0.00 1.5x107 La, Zr, Nd, Eu, Pm, Pr, 7

l l

a Sm, Y, Cm, Am

_ Lanthanides 0.00

" 1.5x107 Ce, Pu, Np, Zr Molybdenum ___l 0.00 0.10 Mo, Te, Nb Group Elements Because the current version of RADTRAD-NALI is limited to nine Chemical Groups (not including Non-Radioactive Aerosols) with Barium and Strontium as separate groups, and because the Lanthanides and Cerium groups have the same release fractions in Table 3-10, the Cerium elements are moved into Group 9 to be combined with the Lanthanides, and the Molybdenum elements are moved into Group 8.

These changes are incorporated in Table 3-11.

iz Table 3-11. Core Inventory Released Into Containment in RADTRAD-NAI Model Group Gap Release

_ Early In-vessel number Phase Phase 1

l Noble gases 0.022.

0.94 Xe, Kr Halogens 0,007 0.37

_ LE, Br 3

l Alkalimetas

=

0.005, (itéi~*<< NS RN 47 Tellurium metals 0.007 0.30 Te, Sb, Se Group Name Elements

NUMERICAL Attachment A to NAS-2357-007 NAS-2357-007 ADVISORYSOLUTIONS Evaluation of Changes for AST Re-analysis Revision 0 a Zachry Group company for an Increase in Fuel Burnup Page 27 of 48 ae i

Table 3-11. Core Inventory Released IntoContainment in RADTRAD-NAI Model l

Sand 6 Strontium, Barium 1.4x103 0.004

__Ba,Sr__

7 Noble metals 0.00 0.006

_.. Ru, Rh, Pd, Co 8

_ Molybdenum 0.00 0.10 l Mo, Te,Nb Lanthanides plus La, Zr, Nd, Eu, Pm, 9

P 0.00 1.5x107

Pr, Sm, Y, Cm, Am l

l Cerium Group l

lus Ce, Pu, Np Iodine available for release to the atmosphere is assumed to be 95% particulate, 4.85%

elemental, and 0.15% organic for the containment leakage pathway and 97% elemental and 3%

organic for the other pathways.

3.3.3.2.2 Release Models 3.3.3.2.2.1 l Containment Leakage For the containment leakage pathway, all activity released from the fuel is assumed to go into the unsprayed portion of containment before being mixed with the sprayed portion of the containment. The time-dependent removal of elemental iodine and particulates from the containment atmosphere is accomplished by containment sprays, radioactive decay, and leakage from containment. The noble gases and organic iodine are subject to removal only by radioactive decay and leakage from containment.

The maximum free volume of the containment modeled in the containment leakage pathway is 2.7E6 ft3. The portion of this volume covered by spray drops (85%) and its unsprayed portion (15%) are modeled separately. The mixing rate between the sprayed and unsprayed regions, as allowed by item A-2.3 of Appendix A of RG 1.183 Rev. 1, is assumed to be two turnovers of the unsprayed regions per hour which is 13,500 cfm. A conservative delay of 2 minutes is assumed before mixing is credited. Mixing continues for the remainder of the event.

The containment is assumed to leak at the design leak rate of 0.2% 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 (0.1% per day) for the remainder of the 30-day period considered in the analysis.

Containment sprays are actuated at 2 minutes following accident initiation and terminate 4.8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months following accident initiation.

Section III.4.c (1) of SRP 6.5.2 (Reference [15]) identifies a methodology for the determination of spray removal of elemental iodine. The removal rate constant is determined by:

As = 6KgTF/VD

where, As = spray removal rate coefficient, hr" Kg = Gas phase mass transfer coefficient, m/min T = Time of fall of the spray drops, min F = Volume flow rate of sprays, m3/hr

NUMERICAL Attachment A to NAS-2357-007 NAS-2357-007 s*

ADVISORY SOLUTIONS l Evaluation of Changes for AST Re-analysis Revision 0 a Zachry Group company for an Increase in Fuel Burnup Page 28 of 48 V = Containment sprayed volume, m?

D = Mass-mean diameter of the spray drops, m (units shown above are representative)

The resulting removal coefficient for elemental iodine is 37.3 hr !. SRP 6.5.2 allows for elemental iodine removal credit of up to 20 hr!; therefore, 20 hr! is used in the analysis. The elemental removal is modeled until a decontamination factor (DF) of 200 is reached at which time elemental iodine removal is terminated. For this analysis, a DF of 200 would be reached at 6.92 hours0.00106 days 0.0256 hours 1.521164e-4 weeks 3.5006e-5 months if containment spray were not terminated at 4.8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

Also, the spray removal coefficient is a function of the spray flow rate. The analysis used a spray removal coefficient based on a conservatively low spray flow rate to bound both modes (injection and recirculation) of spray operation.

Section III.4.c (4) of SRP 6.5.2 identifies a methodology for the determination of spray removal of particulates (aerosols).

The removal rate constant is determined by:

dp = 3hFE /2VD

where, dp =spray removal rate for particulates, hr!

h = Drop fall height, m F = Volume flow rate of sprays, m?/hr V = Containment sprayed volume, m?

E/D = Ratio of dimensionless collection efficiency (E) to the average spray drop diameter (D), m!.

(units shown above are representative)

The resulting removal coefficient for particulates is 6.46 hr! until a DF of 50 is reached at 5.06 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . After this time the particulate removal coefficient would be reduced bya factor of 10 to 0.646 hr ' if containment spray were not terminated at 4.8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

The sump pH is maintained at an equilibrium value greater than 7.0, FSAR Section 6.5.2.3 and Table 15.6-6 (Sheet 2). Therefore, as in the CLB, no re-evolution of iodine is assumed to occur.

A summary of input parameters used in the containment spray analysis is provided in Table 3-12.

Attachment A to NAS-2357-007 Evaluation of Changes for AST Re-analysis for an Increase in Fuel Burnup NUMERICAL ADVISORY SOLUTIONS a Zachry Group company Table 3-12. Containment Spray Parameters for LOCA dose Implementation of Burnup Increase and Rev.1 of AST Regulatory Guide 2.7 million (max.)

Containment volume (ft*)

. Fraction of containment volume that is sprayed (%)

85 Beginning of containment spray delivery (minutes) 2 Duration or time at which containment spray ends (hours) 4.8 3086 Spray flow rate (gpm)

Fall height of spray drops (feet) 131.4 (average)

Organic iodide spray removalcoefficient (hours)

Zero Elemental iodine spray removal coefficient (hours')

20

__ gas phase masstransfer coeff., K, (ft/min) l 9.5 terminal velocity (ft/min) a 790 mass mean drop diameter (microns) maximum Decontamination Factor Particulate/aerosol spray removal coeff. (hours"') for beginning of event l

collection efficiency + drop diameter, (meters)

Particulate/aerosol spray removal coeff. (hours!) for Decontamination Factor > 50 E/D (meters)

. Minimum equilibrium pH for sump liquid during recirc.

NAS-2357-007 Revision 0 Page 29 of 48

Current/previous AST Licensing Basis 2.7 million (max.)

85 2

4 3086 131.4 (average)

Zero

=

NUMERICAL Attachment A to NAS-2357-007 NAS-2357-007 ADVISORY SOLUTIONS Evaluation of Changes for AST Re-analysis Revision 0 l

a Zachry Group company for an Increase in Fuel Burnup Page 30 of 48 3.3.3.2.2.2.

Emergency Core Cooling System Leakage For the ECCS leakage pathway, all iodine activity released from the fuel is assumed to be in the sump solution immediately. The only removal of activity from the sump is by radioactive decay or leakage to the auxiliary building. The sump volume is 428,000 gallons. When ECCS recirculation is established following the LOCA, leakage is assumed to occur from ESF equipment in the auxiliary building. Recirculation is modeled to initiate at the time of control room isolation (62 seconds) and continues throughout the event.

The leakage to the auxiliary building is modeled at a rate of2gpm. The leakage value was doubled in accordance with RG 1.183 Rev. 1. The analysis assumes that 10% of the iodine activity in the leakage becomes airborne and is available for release to the environment. The activity of the airborne leakage is further reduced as it is released through the auxiliary building vent filters with 90% efficiency for all forms of iodine.

3.3.3.2.2.3 Refueling Water Storage Tank Back-Leakage For the RWST back-leakage pathway, a portion of the ECCS recirculation is assumed to leak into the RWST. All iodine activity released from the fuel is assumed to be in the sump solution immediately. The only removal of activity from the sump is by radioactive decay or leakage to the RWST. The sump volume is 428,000 gallons. Recirculation is modeled to initiate at the time of control room isolation (62 seconds) and continues throughout the event.

Leakage to the RWST is modeled at a rate of 4 gpm (3 gpm below the water line, and 1 gpm above the water line). The activity is modeled to be delivered directly to the gas filled portion of the RWST; however, only 10% of the activity in the 1gpm becomes airborne for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of the event and 8% thereafter is available for release to the environment. The vast majority of the radioiodine in the 3 gpm delivered below the water line is retained in the liquid remaining in the RWST. This retention in the liquid is supported by a calculation performed in accordance with NUREG/CR-5950, accounting for gradual changes in pH and iodine concentration in the RWST liquid. The release rate from the RWST to the environment is based on the volume displacement from the incoming leakage. An adjustment is made to account for a reduction in the RWST gas volume available for dilution as the leakage into the RWST increases the water level. The RWST gas volume is decreased by the rate of leakage into the RWST.

3.3.3.2.2.4.

Containment Purge For the containment purge system release pathway, all of the initial primary coolant activity is instantly released from the RCS and is evenly distributed throughout the containment volume.

The total initial mass in containment modeled in this pathway is 723,290 Ibm (551,068 Ibm from the RCS and 172,222 Ibm as air). The only removal of activity from containment is by radioactive decay or the purge flow. The maximum air flow rate from containment to the environment as a function of time is modeled until the purge line is isolated at 11 seconds.

Instead of simply using the normal mini-purge flow rate of 4000 cfm the air flow utilized reflects additional flow resulting from the increase in containment pressure resulting from the mass and energy release from the RCS. Note that the assumed iodine chemical fractions do not impact the analysis results since spray removal is not credited for removal of RCS activity in containment.

NUMERICAL Attachment A to NAS-2357-007 NAS-2357-007 ADVISORY SOLUTIONS Evaluation of Changes for AST Re-analysis Revision 0 a

Zachry Group company for an Increase in Fuel Burnup Page 31 of 48 3.3.3.2.3 Control Room In the event of a LOCA, the low pressurizer pressure SI setpoint will be reached almost immediately following the break. The SI signal causes the control room to switch from the normal operation mode to the emergency operation mode. The switch is conservatively modeled at 62 seconds following event initiation, which includes a 60-second delay from the initiating signal. As discussed in Section 3.3.2.1, operator action is taken after 30 minutes to remedy the half-train failure in the Control Building/Control Room Emergency HVAC to isolate the ventilation train with failed filtration.

As discussed in Section 3.3.2.4.1, the dose to control room personnel from external sources was calculated to be 0.019 rem TEDE. These external sources include the activity remaining in containment following the LOCA, the activity cloud outside of the control room in the environment, and the activity buildup on recirculation filters. This is added to the dose calculated from the four release paths discussed above.

As discussed in Section 3.3.2.5, the dose to control room personnel due to transit to and from the control room has not been calculated for this submittal. This is a change from the CLB which includes transit dose in the LOCA dose evaluation.

Note that the assumed iodine chemical fractions for the containment purge pathway do not impact the analysis results since neither spray nor filtration is credited for removal of the RCS activity in the release path. Also, since control room isolation does not occur until after mini-purge is isolated, no filtration of air flow to the control room is modeled.

3.3.3.3 Acceptance Criteria The EAB and LPZ dose acceptance criterion for a LOCA is 25 rem TEDE per RG 1.183 Rev.

1 which is also the 10 CFR 50.67 limit. The acceptance criterion for the control room dose is 5 rem TEDE per 10 CFR 50.67. The acceptance criterion for the TSC dose is 5 rem TEDE as allowed by GDC 19, in accordance with Reference [14].

The EAB doses are calculated for the worst 2-hour interval. The control room, TSC, and LPZ doses are calculated for 30 days.

NAS-2357-007 Revision 0 Page 32 of 48 NUMERICAL Attachment A to NAS-2357-007 ADVISORY SOLUTIONS Evaluation of Changes for AST Re-analysis a Zachry Group company for an Increase in Fuel Burnup Results and Conclusions The following summarizes the doses for each pathway analyzed for the LOCA dose consequence analysis:

Containment Leakage ECCS Leakage RWST Back-Leakage Containment Purge Containment Leakage ECCS Leakage RWST Back-Leakage Containment Purge Control Room Containment Leakage ECCS Leakage RWST Back-Leakage Containment Purge External Shine Containment Leakage ECCS Leakage RWST Back-Leakage Containment Purge External Shine Control room TSC 1.252 0.826 0.265 0.001 0.178 3.41 9.98 4.80 2.52 rem TEDE rem TEDE rem TEDE rem TEDE rem TEDE rem TEDE rem TEDE rem TEDE rem TEDE rem TEDE rem TEDE rem TEDE rem TEDE rem TEDE rem TEDE rem TEDE rem TEDE rem TEDE The total LOCA doses from all the pathways are listed below:

rem TEDE rem TEDE rem TEDE rem TEDE The EAB doses are reported for the worst 2-hour interval, determined to be from 2.74 to 4.74 hours8.564815e-4 days 0.0206 hours 1.223545e-4 weeks 2.8157e-5 months for the containment leakage case, 4.65 to 6.65 hours7.523148e-4 days 0.0181 hours 1.074735e-4 weeks 2.47325e-5 months for the ECCS leakage case, 320 to 322 hours0.00373 days 0.0894 hours 5.324074e-4 weeks 1.22521e-4 months for the RWST back leakage case, and 0 to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for the containment mini-purge case It is concluded that for a LOCA, the accident doses meet the applicable acceptance criteria.

NUMERICAL ADVISORY SOLUTIONS

=

a Zachry Group company Attachment A to NAS-2357-007 Evaluation of Changes for AST Re-analysis for an Increase in Fuel Burnup Table 3-13. Summary of Inputs for LOCA Core inventory Release fraction from fuel Release timing from fuel L.

Chemical Groups Chemical form of release

Dose conversion factors X/Q limiting short term atmospheric dispersion factors L.

Breathing rates (cubic meters/second)

Control Room (all the time)

EAB and LPZ 0

8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 8

24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months L

_Containment (air) leakage case Chemical form of Iodine released to the containment Containment net free air volume (ft?)

. Containment leak rate (percent per day) first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months PercentofContainment Leakage that is unfiltered Implementation of Burnup Increase and Rev.1 of AST

- Regulatory Guide 107 isotopes in Table 3-6 Table 2 in RG 1.183 Rev.

1 l (See Table 3-10)

Table 5 in RG 1.183 Rev.

1

_ (See Table 3-9)

_ Table 6 in RG 1.183 Rev.

1 l (SeeTable 3-11).

Section 3.5 of RG 1.183 Rev.

1, (No change from CLB)

No change

_ No change in values.

Application is changed per RG 1.183 Rev.

1 Section 5.3.to ensure maximum X/Q is used during period of maximum

_ release.

ee No change in values.

No change.

Application is changed per RG

_ 1.183 Rev.

1 Section 5.3.to

ensure maximum dose during period of maximum release for RWST back-leakage case.

No change No change 1 No change No change

_' No change

+;

NAS-2357-007 Revision 0 Page 33 of 48 Current/previous AST Licensing Basis 1 FSAR Table 15A-3 Table 2 in RG 1.183 Rev. 0.

l (FSAR Table 15.6-6)

Table 4 in RG 1.183 Rev. 0 Table 5 in RG 1.183 Rev. 0

" Section 3.5 of RG 1.183 Rev.

0.

FSARTable 15.6-6

_ FSAR Table 15A-4 (FGR 11 and 12).

FSAR Table 15A-2

FSAR Table 15A-1 (Sheet 2) l 3.5x 104 3.5 x 10+

1.8x 104 1 2.3 x 10° 95% aerosol as cesium iodide (CsI) 4.85% elemental

0.15% organic 2.7 million 0.20 0.10 100

NUMERICAL Attachment A to NAS-2357-007 ADVISORY SOLUTIONS Evaluation of Changes for AST Re-analysis "a Zachry Group company for an Increase in Fuel Burnup Table 3-13. Summary of Inputs for LOCA Equilibrium sump pH Fraction of containment volume that is sprayed (%)

Exchange rate between sprayed and unsprayed (cfm)

Spray removal coefficients (per hour)

- Elemental iodine

- Organic iodine

- Particulates/aerosol Decontamination Factor limits for Containment Spray

- Elemental iodine

- Organic iodine

- Particulates/aerosol l

Beginning of containment spray mixing between sprayed and unsprayed regions (minutes)

_ Duration of containment spray (hours)

Natural deposition coefficient for aerosol in unsprayed region of containment (hour)

ECCS recirculation leakage to the RAB Fraction of core activity released to sump

- Noble gas

- Iodine

- Particulates/aerosol Minimum containment sump liquid volume (gallons)

Chemical form of Iodine released by evaporation of leakage Leake rate (gpm) constant for 30 dayduration Fraction of iodine in leakage that is airborne on release to RAB floor RAB Emergency exhaust ESF filter efficiency (%)

No change Implementation of Burnup l

Increase and Rev.1 of AST Regulatory Guide No change l

No change No change 20.0 prior to 4.8 hrs 0.0 thereafter No change 6.46 prior to 4.8 hrs 0.0 thereafter No change No change 48 l

No change l

Zero 37.7%

l 100% (remains in liquid) 428,000 No change Ne change No change NAS-2357-007 Revision 0 Page 34 of 48 Current/previous AST Licensing Basis FSAR Section 6.5.2.3.

> 7.0 l

85 13,500 20.0 prior to DF limit 0.0 6.46 prior to DF limit

_ 0.646 thereafter 200.0 0.0 Zero 40%

"100% (remains in liquid) 428,000 97% elemental

_ 3% organic l

2 gpm l

l

al NUMERICAL Attachment A to NAS-2357-007 NAS-2357-007 3N:

ADVISORY SOLUTIONS Evaluation of Changes for AST Re-analysis Revision 0 a Zachry Group company for an Increase in Fuel Burnup Page 35 of 48 Table 3-13. Summary of Inputs for LOCA Implementation of Burnup Item Increase and Rev.] of AST. CicencinyBasie.

AST l Regulatory Guide J

, ECCS recirculation leakage to the RWST Leake rate (gpm) constant for 30 day duration No change FSAR Section 15.6.5.4.1.3 4 gpm total:

3 gpm delivered below water line in tank and 1 gpm above.

Fraction of iodine airborne

>> No change FSAR Section 15.6.5.4.1.3 0.1 (of1 gpm delivered above water line) for first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 0.08 (of1 gpm above) for remainder of event l Gallon initial air volume at beginning of recirculation No change 256,626 allons)

Amount of stable iodine in the core (grams)

, 20,273

_ 17,370

' Case forContainment mini-purge initially in operation i

oe RCS specific activity (wCi/gm)

No change FSAR Table 15A-5 lodine (D.E. equiv. 1.0 wCi/gm I-131)

OL-27 Noble gas (D.E. equiv. 225 wCi/gm Xe-133)

Alkali metal (based on 1% fuel defects in FSAR Table 11.1-5) 7 RCS liquid mass (pounds)

_ No change 551,068 Initial air mass inside the Reactor Containment Building No change

172,222

_(pounds)

Time delay to close two initially open 18 mini-purge lines _ No change 11 seconds

_Credit for filtration of exhaust flow

__l No change none l Direct release of activity from fuel

_No change none

_ Time of Control Room Isolation

_ No change Not applicable Control Room Control building No change Mixing Volume (ft)

- 148,000 Filtered air intake (cfm)

Prior to operator action (0 30 min) 900 After operator. action (30 min 720 hrs) 450 Unfiltered inleakage (cfm)

FSAR Figure 15A-2 Filter efficiency (particulates) (%)

95

NUMERICAL Attachment A to NAS-2357-007 NAS-2357-007 ADVISORY SOLUTIONS Evaluation of Changes for AST Re-analysis Revision 0 a Zachry Group company for an Increase in Fuel Burnup Page 36 of 48 Table 3-13. Summary of Inputs for LOCAL l

Implementation of Burnup

- Increase and Rev.1 of AST

rurentPrevious AST icensing Basis Regulatory Guide l

/

Control Room No change Volume (ft?)

48,500 Filtered flow from Control Building (cfm) 440 Unfiltered flow from Control Bldg (cfm)

Prior to operator action (0 30 min) ct)

After operator. action (30 min 720 hrs) 0 Filtered recirculation (cfm)

- 1030 Filter efficiency (all forms of Iodine) (%)

95

__ Unfiltered inleakge (cfm) l l FSAR Figure 15A-2 Control Room Breathing rate (m?/sec)_

, No change

, 3.5 x 104 Control Room occupancy fraction

)

No change (hours) 0-24 1.0 24-96 0.6 96

- 720 0.4 i

1.91E-02 rem TEDE 1.18E-02 rem TEDE l Shine dose to control room operators j See Section 3.3.2.4.1

NUMERICAL Attachment A to NAS-2357-007 NAS-2357-007 s*

ADVISORY SOLUTIONS l Evaluation of Changes for AST Re-analysis Revision 0 a Zachry Group company for an Increase in Fuel Burnup Page 37 of 48 3.3.4 Fuel Handling Accidents (FSAR Section 15.7.4.5) 3.3.4.1 Introduction The postulated fuel handling accident has been analyzed for two cases: a fuel handling accident outside the containment, i.e., the fuel handling building (FHA_FHB), and a fuel handling accident inside the reactor building (FHA_RCB).

The accident is defined as the dropping of a spent fuel assembly onto the fuel storage area floor, refueling pool floor, or cask loading pool. All the fuel rods contained in the dropped assembly are assumed to be damaged. In addition, for the analyses of the accident in the reactor building the dropped assembly is assumed to damage 20 percent of the rods of an additional assembly.

3.3.4.2. Input Parameters and Assumptions The analysis of the FHA radiological consequences uses the analytical methods and assumptions outlined in RG 1.183 Rev. 1, Appendix B (Reference [3]). A summary of input parameters and assumptions is provided in Table 3-15 for the FHB case and in Table 3-16 for the RCB case. A listing of the CLB parameters is also included in these Tables for comparison to the AST parameters. Conformance with the Main Body and Appendix B of RG 1.183 Rev.

1 is shown in Attachment A, Sections A.1 and A.3, respectively.

3.3.4.2.1 Source Terms The source term developed for the FHA analysis is represented by a combination of the Nuclide Inventory File (.nif), the Release Fraction and Timing (.rft) file, and the power level set in the Plant Scenario File (.psf). The core fission product activity is provided in Table 3-6. Appendix B of RG 1.183 Rev. 1 discusses a release with two phases.

Phase I Instantaneous releasefrom damaged rods, 0 -2 hrs It is assumed that one fuel assembly is damaged for the FHA_FHB case and that the equivalent of 1.2 fuel assemblies is damaged for the FHA_RCB case. The calculations use the PWR gap release fractions shown in Table 4 of RG 1.183 Rev. 1, Reference [3]. These gap fractions represent steady-state fission product release from the fuel rod plenum and gap for fuel operating at power levels below the burnup-dependent power envelopes depicted in Figure l of the RG.

The Callaway extended burnup core design operates within this envelope and therefore Table 4 is applicable. In the calculation of activity releases from the damaged fuel, the maximum radial peaking factor of 1.65 is applied.

The decay time used in determining the inventory of the damaged rods is 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . Thus, the analysis supports the TS bases B 3.9.7 assumption of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months decay time prior to movement of irradiated fuel.

Section B-2 of RG 1.183 Rev.

1 provides guidelines for determining a DF for elemental iodine as a function of rod internal pressure and water depth between 19 and 23 feet. For an assumed rod internal pressure of 1500 psig the DF is calculated as 414.6.

Attachment A to NAS-2357-007 NAS-2357-007 a

NUMERICAL re y=

ADVISORY SOLUTIONS

_Evaluation of Changes for AST Re-analysis Revision 0 ww a Zachry Group company for an Increase in Fuel Burnup Page 38 of 48 Per Section B-4, the retention of noble gases in the water in the fuel pool or reactor cavity is negligible (i.e., the DF is 1, consistent with the CLB). Particulate radionuclides are assumed to be retained by the water in the fuel pool or reactor cavity (i.e., the DF is infinite) for the first two hours.

3.3.4.2.1.2 Phase 2 Protracted release due to re-evolution as elemental iodine, 2hrs to 30 days Section B.3 of RG 1.183 Rev. l provides a calculation sequence for determining the evolution removal coefficient, Ae (sec!), for a pool as a function of the stable iodine concentration, pH (as influenced by boron concentration), surface area, and volume. The removal coefficient, Xe, is converted to leakage rate to the environment (% per day) for use in RADTRAD-NAI. Table 3-14 summarizes the input paraments and evolution removal coefficient results for the FHB and RCB pools.

l Table 3-14. Pool Removal Coefficient Inputs and Results

- FHB and RCB Pool Parameter i

FHB

[

RCB Total core inventory of stable iodine, g

__ 20,273.28 20,273.28 Pool pH /

4.88 /

4.88/

Boron concentration, ppm

, 2585 2585 l_Surface area, ft l

1668 ISI Volume, gal 399,533 l

326,409 l

[ Evolution removal coefficient, sec!

L.10E-10

1.49E-10

_RADTRAD-NAI leakage rate,% perday=

/s<<<9.S2E-4.

js 1.28E-3 3.3.4.2.1.3 lNon-LOCA Transient Fission Gas Release (other thanfor Rod Ejection)

Section 3.2 of RG 1.183 Rev.1 (Reference [3]) states that applicants should address the risk, using engineering judgment or experimental data, of additional fission product release as a possible result of fuel pellet fragmentation due to impact loads for the FHA.

The assessment performed for this application concludes that, because neither a reactivity insertion nor a prolonged heatup of the fuel is applicable to the FHA, current Fission Gas Release models for this event are expected to be sufficiently conservative.

This conclusion is consistent with the corresponding draft information presented in Reference

[18], DG-1424. More specifically, the corresponding fission gas release discussion on page 31 of DG-1424 has been changed to Qualify the subject non-LOCA DBAs as high temperature, and Replace fuel handling accident with main steamline break.

NUMERICAL Attachment A to NAS-2357-007 NAS-2357-007

ADVISORY SOLUTIONS Evaluation of Changes for AST Re-analysis Revision 0 a Zachry Group company for an Increase in Fuel Burnup Page 39 of 48 3.3.4.2.2 Release Model FHA_FHB All activity released from the fuel pool is 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 .

No credit is taken for mixing or holdup in the fuel building atmosphere. No credit is taken for filtration by the ESF emergency filtration system.

FHA_RCB All gap activity assumed available for release is assumed to be released over two hours. No credit is taken for mixing or holdup in the reactor building atmosphere. No credit is taken for isolation of containment. The accident analysis assumes that a direct pathway exists between containment and the atmosphere for the duration of the release. No credit is taken for filtration.

3.3.4.2.3 Control Room FHA_FHB This submittal does not credit control room isolation for the FHA_FHB. The CLB credited automatic isolation.

FHA_RCB This submittal does not credit control room isolation for the FHA_RCB. The CLB credited automatic isolation.

3.3.4.3 Acceptance Criteria The EAB and LPZ dose acceptance criterion for a FHA is 6.3 rem TEDE per RG 1.183 Rev.

1 which is approximately 25% of the 10 CFR 50.67 limit. The control room dose acceptance criterion is 5 rem TEDE per 10 CFR 50.67. The acceptance criterion for the TSC dose is 5 rem TEDE as allowed by GDC 19, in accordance with Reference [14].

The EAB doses are calculated for the worst 2-hour interval. The control room, TSC, and LPZ doses are calculated for 30 days.

NUMERICAL Attachment A to NAS-2357-007 NAS-2357-007 ADVISORY SOLUTIONS Evaluation of Changes for AST Re-analysis Revision 0 "a Zachry Group company for an Increase in Fuel Burnup Page 40 of 48 3.3.4.4 Results and Conclusions The FHA_FHB case results are listed below:

EAB 2.8E-01 rem TEDE LPZ 9.8E-02 rem TEDE Control room 2.02 rem TEDE TSC 3.3E-01 rem TEDE The FHA_RCB case results are listed below:

EAB 3.4E-01 rem TEDE LPZ 1.2E-01 rem TEDE Control room 2.88 rem TEDE TSC 3.7E-01 rem TEDE The EAB doses reported are for the worst 2-hour interval, determined to be from 0 to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months (post release, 72 to 74 hours8.564815e-4 days 0.0206 hours 1.223545e-4 weeks 2.8157e-5 months overall) for both the FHA_FHB and FHA_RCB cases.

It is concluded that for a FHA, the accident doses meet the applicable acceptance criteria.

Attachment A to NAS-2357-007 NAS-2357-007 Revision 0 Page 41 of 48 NUMERICAL ye" ADVISORYSOLUTIONS Evaluation of Changes for AST Re-analysis a Zachry Group company for an Increase in Fuel Burnup Table 3-15. Summary of Inputs for FHALFHB Implementation of l

Burnup Increase and Current/previous AST Rev.1 of AST Licensing Basis L

Regulatory Guide 1.65 1.65 72 72 1.0 1.0 Table 3 of Rev. 0 of RG Table 4 of Rev.

lof RG,_ !-183 supplemented by a

ws 1.183 NUREG/CR-5009 for Gap activity as a fraction of total activity in fuel l

high burnup fuel. (See FSAR Table 15.B.1 (Sheet 4))

Radial Peaking Factor L

Decay Time(hours)_

Number offuelassemblies affected L

Percent of gapactivity released 100 ae All rods with no exceptions/Vviolations to Number of High Burnup Fuel Pins Analyzed envelope in Fig.

l of RG 1.183 Rev. 1 ae 95.0% aerosol 0% aerosol cemical makeup of effective iodine release from fuel 4.85% elemental 99.85% elemental i.

0.15% organic 0.15% organic

_l Infinity for aerosol Iodine l Effective Pool Decontamination Factors for Phase l 414.6 for elemental 200 for Iodine as short-term release (2-hour duration)

Iodine per Equation B-1 of RG 1.183 Rev.

1 1 for Noble Gas 1.0 for organic Iodine l

1.0 for Noble Gas

_Minimum water depth (feet) 23 23 Le rod internal pressure FHB HVAC exhaust filter efficiency (%)_

0 0

i) 0 1500 psig at 120°F l

Not Applicable

_Building mixing volume (% oftotal volume)_

NUMERICAL Attachment A to NAS-2357-007 NAS-2357-007 ADVISORYSOLUTIONS Evaluation of Changes for AST Re-analysis Revision 0 "a Zachry Group company for an Increase in Fuel Burnup Page 42 of 48 Table 3-15. Summary of Inputs for FHA_FHB l

Implementation of Burnup Increase and Current/previous AST Rev.1 of AST Licensing Basis Regulatory Guide Very high flow rate to ensure that activity is completely released FHB exhaust rate from FHB over 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months l

for Phase 1 Very high flow rate to ensure that activity is completely released from FHB over 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months (no holdup ofactivity in FHB (no holdup of activity in L_

FHB) oo.

te isolation time Not applicable l 120 seconds Delay for Control Room Isolation Not applicable

=: 120 seconds 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for Phase 1 Activity release period 7202=718 hours for 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months l

l Phase 2 Long term (Phase 2) release/escape rate for radioactive l

1.10E-10 iodine gas re-evolution from solution in the Spent Fuel (per equations in App. B Not applicable l Pool, de (sec!)

of [3])

l Total Spent Fuel Pool free water volume 399,533 gallons Notapplicable 1668 fi? (including Cask Total Spent Fuel Pool surface area Loading Pit) l Potal core inventory of stable iodine (1-127 and I-129) 20,273.28 grams l.

Not applicable 1

4.88 (based on a high pH of water in the Spent Fuel Pool boron concentration of l

Not applicable 2585 ppm)

Not applicable l

X/Q limiting short term atmospheric dispersion factors Offsite for EAB and LPZ No change FSAR Table 15A-2 On-site for Control Room (hours)

( sec/meters*)

(hours)

( sec/meters?)

0-74 2.23E-03 0-Isolation 2.23E-03 74-80 1.85E-03 Isolation2 1.17E-03 80 -96 7.49E-04 2-8 1.04E-03 96

- 168 4,30E-04 8-24 4.27E-04 168 -792 3.17E-04 24-96 2.34E-04

__(no isolation) l 96-720 1948-04 l

_-d "Credit for decay or depletion during transit to EAB and

_LPZ?

None

-2357-

-9357-NUMERICAL Attachment A to NAS-2357-007 NAS 2357 007 a

ADVISORY SOLUTIONS Evaluation of Changes for AST Re-analysis Revision 0 a Zachry Group company for an Increase in Fuel Burnup Page 43 of 48 ae Table 3-15. Summary of Inputs for FHA_FHB Implementation of Item Burnup Increase and Current/previous AST Rev.1 of AST Licensing Basis Regulatory Guide Breathing rates (cubic meters/second) control room (entire event) 3.5x 104 offsite 0

8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 3.5x 104 8

24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Control Room Occupancy factors?

First day between l to 4 days after 4 day Time delay for isolation of normal Control Room Not applicable Ventilation (minutes)_

a l

Unfiltered air flow through Control Room (CFM)

- 5000 as normal air flow 5000 for first 2 minutes l

forduration of" coe FSAR Figure 15.A-accident.

2 Allowable Inleakage Values for time after

' Control Room isolation and pressurization L

Filter Efficiency (%)

Control Building Pressurization: Elemental and Organic l Not applicable with no isolation of normal Control Room Control Room Recirculation: Elemental, Organic, and ventilation

, Aerosol

() For the AST FHA_FHB analysis, the timings of the X/Qs, breathing rates, and control room occupancy factors were adjusted by adding 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to account for the delay before fuel movement Control Building Pressurization: Aerosol_

NUMERICAL Attachment A to NAS-2357-007 NAS-2357-007 ADVISORY SOLUTIONS Evaluation of Changes for AST Re-analysis Revision 0 a Zachry Group company for an Increase in Fuel Burnup Page 44 of 48 Table 3-16. Summary of Inputs for FHA_RCB l

Implementation of Burnup Increase and Current/previous AST Rev.1 of AST Licensing Basis

_ Regulatory Guide Radial Peaking Factor 1.65 1.65 Decay Time (hours) 72 1.2 1.2 Table 3 of Rev. 0 of RG Table 4 of Rev.

1 of 1.183 supplemented by l

RG 1.183 NUREG/CR-5009 for high burnup fuel. (See Number of fuel assemblies affected FSAR Table 15.B.1 (Sheet 4))

l Gap activity as a fraction of total activity in fuel Percent of gap activity released 100 All rods with no l

Number of High Burnup Fuel Pins Analyzed exceptions/Vviolations to 32 envelope in Fig. l of RG 1.183 Rev.1 l

95.0% aerosol 0% aerosol 4.85% elemental l

99.85% elemental l

0.15% organic 0.15% organic Infinity for aerosol Iodine Chemical makeup of effective iodine release from fuel rods Effective Pool Decontamination Factors for Phase 1 l

414.6 for elemental l

fe i

as short-term release (2-hour duration)

Iodine per Equation B-200 for Iodine 1 of RG 1.183 Rev. 1 1 for Noble Gas l

1.0 for organic Iodine 1.0 for Noble Gas

. Minimum water depth (feet) 230 23

_Fuel rod internalpressure 1500psig at 120°F

_ Not Applicable FHB HVAC exhaustfilter efficiency (%)

i 1) ae 0

. Building mixing volume (%oftotal volume) 0 0.

Very high flow rate to ensure that activity is l

ee exhaust rate completely released Very high flow rate to ensure that activity is completely released from from containment over

° l

FHB over 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 2 hours for Phase 1

NUMERICAL Attachment A to NAS-2357-007 ADVISORY SOLUTIONS Evaluation of Changes for AST Re-analysis a Zachry Group company for an Increase in Fuel Burnup Table 3-16. Summary of Inputs for FHA_RCB Implementation of Burnup Increase and Rev.1 of AST l

Regulatory Guide 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for Phase 1 Activity release period 720 2 = 718 hours0.00831 days 0.199 hours 0.00119 weeks 2.73199e-4 months for Long term (Phase 2) release/escape rate for radioactive l

1.49E-10

_ iodine gas re-evolution from solution in the Spent Fuel

=

(per equations in App.

lPool, Ae (sec!)

B of [3])

oo Spent Fuel Pool free water volume 326,409 gallons Total Spent Fuel Pool surface area 1511 fv l Total core inventory of stable iodine (I-127 andI-129) 20,273.28 grams boron concentration of 2585 ppm)__

None l

4.88 (based on a high l pH of water in the Cavity Pool L

Credit for decay or depletion during transit to EAB and LPZ?

Breathing rates (cubic meters/second) control room (entire event) 3.5x 10" offsite 0-8 hours 3.5x 104 8

24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months

[

after 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Control Room Occupancy factors!

First day between l to 4 days

_ after 4 days

' Time delay for isolation of normal Control Room Not applicable

_ Ventilation (minutes) l Unfiltered air flow through Control Room (CFM) 5000 as normal air flow for duration of co. FSAR Figure 15.A-2 accident.

Phase 2 od.

NAS-2357-007 Revision 0 Page 45 of 48 Current/previous AST Licensing Basis Not applicable Not applicable Not applicable

_ Not applicable Not applicable None 3.5x 107 3.5 x 104 1.8 x 104 2.3.x 104 LL 5000 for first 2 minutes Allowable Inleakage Values for time after Control Room isolation and pressurization

NUMERICAL

=

ADVISORY SOLUTIONS a Zachry Group company L

Item X/Q limiting short term atmospheric dispersion factors Off-site for EAB and LPZ On-site for Control Room L..

con Efficiency (%)

Control Building Pressurization: Elemental and Organic

_Control Building Pressurization: Aerosol Control Room Recirculation: Elemental, Organic, and Aerosol Attachment A to NAS-2357-007 Evaluation of Changes for AST Re-analysis for an Increase in Fuel Burnup Table 3-16. Summary of Inputs forFHA_RCB Implementation of Burnup Increase and Rev.1 of AST Regulatory Guide No Change (hours)

( sec/meters?)

2.66E-03 2.43E-03 1.01E-03 5.57E-04 4.48E-04 (no isolation)

Not applicable with no isolation of normal Control Room ventilation l

8 24 96 720 NAS-2357-007 Revision 0 Page 46 of 48 Current/previous AST Licensing Basis FSAR Table 15A-2 From FSAR Table 15A-2 (hours)

( sec/meters*)

7.12E-03 8.61E-04 7.54E-04 3.22E-04 1.84E-04 1.50E-04 0

Isolation Isolation 2

2-8 24 96 0

95 95

) For the AST FHA_RCB analysis, the timings of the X/Qs, breathing rates, and control room occupancy factors were adjusted by adding 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to account for the delay before fuel movement:

l

NUMERICAL Attachment A to NAS-2357-007 NAS-2357-007 s*

ADVISORY SOLUTIONS Evaluation of Changes for AST Re-analysis Revision 0 a Zachry Group company for an Increase in Fuel Burnup Page 47 of 48

4. References 1.

NRC, Regulatory Guide (RG) 1.183 Rev. 0, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, July 2000.

Letter, USNRC to Ameren Missouri, Callaway Plant, Unit No.1 Issuance of Amendment No. 233 for Adoption of Alternative Source Term and Revision of Technical Specifications (EPID L-2021-LLA-0177), September 20, 2023, and enclosed Safety Evaluation (ML23166B088).

NRC, Regulatory Guide (RG) 1.183 Rev. 1, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, October 2023.

Numerical Advisory Solutions, NAS-2368-001, Revision 0, Software Dedication Assessment SCALE 6.1.3/ORIGEN-S, 01/29/2025.

Numerical Applications Inc., NAI-9912-04, Revision 7, RADTRAD-NAI Release Version 1.3(QA) Documentation, January 2017.

NUREG/CR-6604, RADTRAD: A Simplified Model for RADionuclide Transport and Removal and Dose Estimation, December 1997.

NUREG/CR-6604, Supplement 1, RADTRAD: A Simplified Model for RADionuclide Transport and Removal and Dose Estimation, June 1999.

NUREG/CR-6604, Supplement 2, RADTRAD: A Simplified Model for RADionuclide Transport and Removal and Dose Estimation, October 2002.

Environmental Protection Agency 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.

. Environmental Protection Agency Federal Guidance Report No. 12, External Exposure to Radionuclides in Air, Water, and Soil, EPA-402-R-93-081, September 1993.

. "RETRAN-3D A Program for Transient Thermal-Hydraulic Analysis of Complex Fluid Flow Systems," M. P. Paulsen, et al., NP-7450(A), Research Project 889-10, EPRI, Rev. 11, May 2017.

. Grove Engineering, MicroShield Users Manual, Version 9, Rev. 5, 2011

. Numerical Advisory Solutions, NAS-2372-001, Revision 0, Software Dedication Assessment MicroShield 9 Code,, 02/04/2025.

. Supplement l to NUREG-0737, Clarification of TMI Action Plan Requirements:

Requirements for Emergency Response Capability, January 1983.

. NUREG-0800, Standard Review Plan 6.5.2, Revision 4, Containment Spray as a Fission Product Cleanup System, March 2007.

Attachment A to NAS-2357-007 NAS-2357-007 NUMERICAL s"

ADVISORY SOLUTIONS l Evaluation of Changes for AST Re-analysis Revision 0 a Zachry Group company for an Increase in Fuel Burnup Page 48 of 48

16. PWROG Letter OG-22-117, PWROG Comments on Draft Regulator Guide (DG), DG-1389, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, 87 FR 23891: Docket ID NRC-2021-0179, June 20, 2022.

17. Document PNNL-18212, Rev. 1, Update of Gap Release Fractions for Non-LOCA Events Utilizing the Revised ANS 5.4 Standard, June 2011.

18. NRC Draft Regulatory Guide DG-1425 (Proposed Revision 2 to Regulatory Guide 1.183)

Alternative Radiological Source Terms for Evaluation Design Basis Accidents at Nuclear Power Reactors, November 2024. (ADAMS accession no. ML24304A864)

"=

NUMERICAL Attachment A to NAS-2357-007 NAS-2357-007

- ADVISORY SOLUTIONS Evaluation of Changes for AST Re-analysis Revision 0 2

Zachry Group company for an Increase in Fuel Burnup Page A-1 of A-35 Attachment A Regulatory Guide 1.183 Rev. 1, Conformance Tables A.1 RG 1.183 Rev. 1, Main Body Page A-2 A.2 RG 1.183 Rev. 1, Appendix A LOCA Page A-21 A.3 RG 1.183 Rev. 1, Appendix B FHA Page A-29 Note: In Tables A.1 through Table A-3 the text shown in the RG Position columns is taken from Regulatory Guide 1.183 Rev. 1.

Therefore, references to footnotes, tables, and numbered references may be found in the regulatory guide.

_ Section A.l 3.

3.1 ACCIDENT SOURCE TERM "Fission Product Inventory NUMERICAL

=

ADVISORY SOLUTIONS a Zachry Group company RG 1.183 Rev. 1, Main Body Table A-1. Conformance with Regulatory Guide 1.183 Rev. 1 Main Sections Analysis iz Regulatory Guide 1.183 Rev. 1 Position The inventory of fission products in the reactor core and available for release to the containment should be based on the maximum full power operation ofthe core with, as a minimum, current licensed values for fuel enrichment, fuel burnup, and an assumed core power equal to the current licensed rated thermal power times the ECCS evaluation uncertainty. The period of irradiation should be of sufficient duration to allow the activity of dose-significant radionuclides to reach equilibrium or to reach maximum values. The core inventory should be determined using an appropriate isotope generation and depletion computer code such as ORIGEN2 or ORIGEN-ARP. Core inventory factors (Ci/MWt) provided in TIDI 14844 and used in some analysis computer codes were derived for low burnup, low enrichment fuels and should not be used with higher burnup and higher

___enrichment fuels.

For the MHA LOCA, all fuel assemblies in the core are assumed to be affected and the core average inventory should be used. For DBA events that do not involve the entire core, the fission product inventory of each of

_ the damaged fuel rods is determined by dividing the total core inventory by the term based upon full power, core average number of fuel rods in the

. core. To account for differences in power level conditions. The FHA source term is across the core, radial peaking factors from the facility's core operating limits derived from the core source term, the report (COLR) or technical specifications should be applied in determining number of damaged fuel rods, and the inventory of the damaged rods.

Attachment A to NAS-2357-007 Evaluation of Changes for AST Re-analysis for an Increase in Fuel Burnup

~~ Conforms

Conforms For the MHA LOCA, all fuel assemblies were assumed NAS-2357-007 Revision 0 Page A-2 of A-35 Basis of Conformance l The inventory of fission products in the reactor core and l available for release to the containment was based on the maximum full power operation with a core thermal power of 3565 MWt.

A 2% calorimetric uncertainty is applied as a multiplier on the total core inventory

' resulting from the ORIGEN runs.

Core design parameters (enrichment, burnup, and MTU loading) are based on cycles 29-31 to model a bounding cycle.

Core inventory factors (Ci/MWt) provided in TID-14844 are not used here.

to be affected and the core average inventory was used.

A peaking factor of 1.65 was used for DBA events that do not involve the entire core (fuel handling accident),

with fission product inventories for damages fuel rods determined by multiplying the total core inventory by the fraction of damaged rods.

L.

RG Section 3.1 3.2 NUMERICAL ADVISORY SOLUTIONS a

Zachry Group company

[

Table A-1.

Conforman Attachment A

to NAS-2357-007 Evaluation of Changes for AST Re-analysis for an Increase in Fuel Burnup Regulatory Guide 1.183 Rev.

1 Position No adjustment to the fission product conservative assembly peaking factor inventory should be made for events postulated to occur during power which corresponds to the maximum fuel operations at less than full rated power or those postulated to occur at the rod peaking factor permitted at the beginning of core life.

For events postulated to occur while the facility is

shutdown, e.g.,

a fuel handling

accident, radioactive decay from the time of shutdown may be modeled.

l Release Fractions The core inventory release fractions, by radionuclide

groups, for the gap l

release and early in-vessel damage phases for MHA LOCAs are listed in

Table 2

for PWRs.

These fractions are applied to the equilibrium core inventory described in Regulatory Position 3.1.

Table 2

PWR Core Inventory Fraction Released Into Containment Noble Gases Halogens Alkali Metals Tellurium Metals Ba, Sr Noble Metals Cerium Group Lanthanides Molybdenum OF)

Release Phase 0.022 0.007 0.005 0.007 1.4x 103 0.00 0.00 0.00 0.00 Early In-Vessel Phase 0.94 0.37 0.23 0.30 0.004 0.006 1.5x 107 1.5x 107 0.10 Analysis Conforms ce with Regulatory Guide 1.183 Rev.

1

Main Sections

No adjustments for less than full power were made in any analysis.

For the fuel handling

accident, 72-hours of radioacti NAS-2357-007 Revision 0

Page A-3 of A-35 Basis of Conformance decay after shutdown was modeled.

[

For the LOCA

event, the core inventory release fractions, by radionuclide

groups, for the gap release and early in-vessel damage phases in Table 2

were utilized.

NUMERICAL Attachment A to NAS-2357-007 NAS-2357-007 l

ADVISORY SOLUTIONS Evaluation of Changes for AST Re-analysis Revision 0 2 Zachry Group company for an Increase in Fuel Burnup Page A-4 of A-35 Table A-1. Conformance with Regulatory Guide 1.183 Rev. 1 Main Sections Ssaon Regulatory Guide 1.183 Rev. 1 Position Analysis l

Basis of Conformance 3.2 Release Fractions For non-LOCA DBAs, table 4 for PWRs list the maximum steady-state Conforms For non-LOCA events, the fraction of the core inventory l fission product release fractions residing in the fuel rod void volume assumed to be in the gap by radionuclide group in Table (plenum and pellet-to-cladding gap), by radionuclide group, available for 4 was utilized in conjunction with the maximum core release upon cladding breach.

radial peaking factor of 1.65.

The applicability of the steady-state fission product release fractions in table 4 is limited to currently approved (as of the issuance of this RG) full-The Westinghouse fuel rod is a currently approved full length UO> fuel rod designs operating up to a maximum rod-average length UO) fuel rod design operating up to a maximum burnup of 68 GWd/MTU at power levels below the burnup-dependent rod-average burnup consistent with the limits of RG 1.183 power envelopes depicted in figure 1. In figure 1, the bounding rod-average

- Rev. l at power levels below the burnup-dependent power

power refers to the rod-average linear heat generation rate ofthe peak rod, envelopes depicted in figure 1.

and the peak power refers to the maximum local linear heat generation rate

in the core. Licensees should make adjustments to account for power uncertainties and plant maneuvering when comparing operating power histories to figure 1.

Table4 PWR Steady -State Fission Product Release Fractions Residing in the Fuel Rod Plenum and Gap Group Fraction 1-131 0.07 1-132 0.07 Kr-85 0.40 Other Noble Gases 0.06 Other Halogens 0.04 Alkali Metals 0.20

3.2 Release Fractions NUMERICAL Attachment A to NAS-2357-007

[ADVISORY SOLUTIONS Evaluation of Changes for AST Re-analysis l

a Zachery Group company for an Increase in Fuel Burnup Table A 1. Conformance with Regulatory Guide 1.183 Rev. 1 Main Sections Regulatory Guide 1.183 Rev. 1 Position Analysis For non-LOCA DBAs involving a rapid increase in fuel rod power, such as the... PWR control rod ejection accident, additional Conforms fission product releases may occur as a result of pellet fracturing and grain boundary separation. This transient fission gas release (TFGR) increases the amount of activity available for release into the reactor coolant system for fuel rods that experience cladding breach. The empirical database suggests that TFGR is sensitive to both local fuel burnup and peak radial average fuel enthalpy rise. As a result, separate low-burnup and high-burnup TFGR correlations for stable, long-lived radionuclides (e.g.,

krypton (Kr)-85 and cesium-137) are provided, as follows:

pellet burnup < 50 GWd/MTU, TFGR (long-lived isotopes) = maximum[ (0.26

AH) 13)/ 100, 0],

{Equation 1) pellet burnup > $0 GWd/MTU, TFGR (long-lived isotopes) = maximum[ (0.26

AH) 5)/ 100, 0],

(Equation 2) where TFGR nsient fission gas release fraction, and AH = increase in radial average fuel enthalpy, A calories per gram For stable, long-lived noble gases (e.g., Kr-85) and alkali metals (e.g., cesium-137), the transient fission product release is equivalent to the above burnup-dependent correlations.

For volatile, short-lived radioactive isotopes such as halogens (e.g.,

iodine ({)-131) and xenon (Xe) and Kr noble gases except Kr-85 (e.g., Xe-133, Kr-85m), the transient fission product release correlations should be multiplied by a factor of 0.333. The low-burnup and high-burnup TFGR correlations for volatile, short-lived radioisotopes are as follows pellet burmup < 50 GWd/MT!

TFGR (short-lived isotopes) = 0.333

maximum[ (0.26

AH)

- 13)/ 100, 0],

(Equation 3) pellet burnup > 50 GWd/MTU.

TFGR (short-lived isotopes) = 0.333

maximum [ (0.26

AH)

- 5)/ 100, 0),

(Equation 4) where TFGR=

transient fission gas release fraction, and AH=

increase in radial average fuel enthalpy, A calories per gram.

NAS-2357-007 Revision 0 Page A-5 of A-35 r

Basi:

F L.

asis of Conformance The fission product releases for the Control Rod Ejection Accident use the Trap correlations for stable, long-lived, and volatile, short-lived radioactive isotopes with the less than 50 GWd/MTU correlations assumed for all fuel that experience cladding breach.

The less than 50 GWd/MTU correlations are used as they produce higher release fractions for the assumed applicable burnup dependent enthalpy rise.

NUMERICAL Attachment A to NAS-2357-007 NAS-2357-007 ADVISORY SULUTIONS.

aluation ofChanges for AST Re-analysis Revision 0 for an Increase in Fuel Burnup Page A-6 ofA-35 8 Zachry Group company Table A 1. Conformance with Regulatory Guide 1.183 Rev.

1 Main Sections sein [

Regulatory Guide 1.183 Rev. 1 Position Analysis Basis of Conformance Release From page 23:

Conforms Engineering Judgement Fractions For the remaining non-LOCA DBAs that predict fuel rod cladding failure, such as the PWR reactor coolant pump locked rotor The Phenomena Identification and Ranking Tables (PIRTs) regarding the consequences of Fue!

and fuel handling accident, additional fission product releases may occur as a result of fuel pellet fragmentation (e.g, fracturing ha

© 7

Fragmentation, Relocation, and Dispersal (FFRD) presented in Reference [a] are based on a of high-burnup rim region) due to loss of pellet-to-cladding mechanical constraint or impact loads.

TFGR (Transient Fission Gas collect f

rt opi In particular, Section 3.3 of [a}

is devoted to subject of Fuel Release) has been experimentally observed under a variety of accident conditions. At the time ofissuance of Revision l of this uechon of expert opinions. In particular, Section 3.3 of

[a]

is devoted to subject of Fue RG, no consensus exists on the mechanism or the computation of TFGR for these events; therefore, future applicants should Handling Accidents, Overall resultsare presented in Sections 3.34 and 4 l 3

In summary, the address this using engineering judgment or experimental data effect of fuel dispersal and Transient Fission Gas Release on the FHA is expected to be small

° compared with the LOCA. For the most part, the impact was scored as low.

No serious implications are recognized for High Burnup Fuel in FHAs.

Experimental Data The expert opinion expressed in Section 3.3.2 of[a] (that the extent offuel fragmentation during an FHA would be near negligible compared to what had been observed under LOCA conditions) is based the experimental results presented in Reference [b]

Because neither a reactivity insertion nor a prolonged heatup of the fuel is applicable to the FHA, current Fission Gas Release models for this event are expected to be sufficiently conservative. This conclusion is consistent with the corresponding draft information presented in Reference [c].

More specifically, the corresponding fission gas release discussion on page 31 of DG-1424 [c] has been changed to

+

Qualify the subject non-LOCA DBAs as high temperature, and

~

Replace fuel handling accident with main steamline break References a.

NUREG/CR-7307, Phenomena Identification and Ranking Tables on High Burnup Fuel Fragmentation, Relocation, Dispersal, and Its Consequences for Design-Basis Accidents in Pressurized-and Boiling-Water Reactors, published May 2024.

(ADAMS accession no ML24155A058)

Viassopoulos, E., D. Papaioannou, R. Nasyrow, V. Rondinella, S. Caruso, and E.W Schweitzer, Experimental Study on the Mechanical Stability ofa59 GWd/MTU Nuclear Fuel Rod, In:

Top uel 2021 Conference, Santander, Spain, October 2428, 2021.

NRC Draft Regulatory Guide DG-1425 (Proposed Revision 2 to Regulatory Guide 1.183) Alternative Radiological Source Terms for Evaluation Design Basis Accidents at Nuclear Power Reactors, November 2024, (ADAMS accession no.ML24304A864)

RG Section 33 NUMERICAL Attachment A to NAS-2357-007 NAS-2357-007 ADVISORY SOLUTIONS Evaluation of Changes for AST Re-analysis Revision 0

a Zachry Group compan ry roup company for an Increase in Fuel Burnup Regulatory Guide 1.183 Rev. 1 Position Analysis rc

Timing of Release Phases Conforms Table 5 tabulates the onset and duration of each sequential release phase for LOCA DBAs at PWRs. The specified onset is the time following the initiation of the accident (i.e., time = 0). The early in-vessel phase immediately follows the gap release phase. The activity released from the core during each release phase should be modeled as increasing in a linear fashion over the duration of the phase. For non-LOCA DBAs in which fuel damage is projected, the release from the fuel gap and the fuel pellet should be assumed to occur instantaneously with the onset of the projected damage.

Table 5 MHA LOCA Release Phases PWRs Phase Onset End Time Gap Release 0.5 min.

0.23 hours2.662037e-4 days 0.00639 hours 3.80291e-5 weeks 8.7515e-6 months Early In-Vessel 0.23 hours2.662037e-4 days 0.00639 hours 3.80291e-5 weeks 8.7515e-6 months 4.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months For facilities licensed with leak-before-break methodology, the onset of Not the gap release phase may be assumed to be 10 minutes. A licensee may propose an alternative time for the onset of the gap release phase, based on facility-specific calculations using suitable analysis codes or on art accepted topical report shown to be applicable to the specific facility. In the absence of approved alternatives, the gap release phase onsets in Applicable

_ Table 4 should be used.

l Page A-7 of A-35 Table A 1. Conformance with Regulatory Guide 1.183 Rev. 1 Main Sections Basis of Conformance The Table 5 PWR onset and durations for the DBA LOCA releases were utilized in the analysis.

Note that the gap release was modeled beginning at 30 seconds and ending at 0.23 hours2.662037e-4 days 0.00639 hours 3.80291e-5 weeks 8.7515e-6 months in order to model the early in-vessel release beginning at 0.23 hours2.662037e-4 days 0.00639 hours 3.80291e-5 weeks 8.7515e-6 months .

No additional delays in gap release were assumed for the DBA analyses.

Attachment A

to NAS-2357-007 Evaluation of Changes for AST Re-analysis for an Increase in Fuel Burnup NAS-2357-007 Revision 0

Page A-8 of A-35 NUMERICAL a"

ADVISORY SOLUTIONS

adZachry Group company Table A

1.

Conformance with Regulatory Guide 1.183 Rev.

1

Main Sections RG Section 3.

l Radionuclide Composition l

Analysis l

Conforms Regulatory Guide 1.183 Rev.

1 Position Table 6

lists the elements in each radionuclide group that should be considered in design basis analyses.

Table 6

Radionuclide Groups Group Elements Noble Gases Xe, Kr Halogens I,

Br Alkali Metals Cs, Rb Tellurium Group Te, Sb, Se, Ba, Sr Noble Metals Ru, Rh, Pd, Co Lanthanides La, Zr, Nd, Eu, Pm, Pr, Sm, Y,

Cm, Am Ce, Pu, Np Mo, Te, Nb Cerium Molybdenum Chemical Form Conforms Of the radioiodine released from the reactor coolant system (RCS) to the containment in a

postulated

accident, 95%

of the iodine released should be assumed to be cesium iodide (CsI),

4.85%

elemental

iodine, and 0.15%

organic iodide.

This includes releases from the gap and the fuel pellets.

With the exception of elemental and organic iodine and noble

gases, fission products should be assumed to be in particulate form.

The same chemical form is assumed in releases from fuel pins in FHAs and from releases from the fuel pins through the RCS in DBAs other than FHAs or LOCAs.

However, the transport of these iodine species following release from the fuel may affect these assumed fractions.

The accident-specific appendices to this regulatory guide provide additional details.

Basis of Conformance

_I The Table 6

elements in each radionuclide group were considered in design basis analyses.

As a

slight reorganization of existing

elements, the Molybdenum elements are assigned as the eighth

group, and because they have the same release fraction in Table 2,

the Cerium elements are combined with Lanthanides in the ninth chemical group.

This change to chemical group designations has also been incorporated in the associated Nuclide Inventory Files.

For releases from the reactor coolant system (RCS) to the containment, 95%

of the iodine released was assumed to be cesium iodide (CsI),

4.85%

elemental

iodine, and 0.15%

organic iodide.

Fission products were assumed to be in particulate form with the exception of elemental and organic iodine and noble

gases,

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NUMERICAL Attachment A

to NAS-2357-007 NAS-2357-007 x"

ADVISORY SOLUTIONS Evaluation of Changes for AST Re-analysis Revision 0

a Zachry Group company for an Increase in Fuel Burnup Page A-11 of A-35 Table A

1.

Conformance with Regulatory Guide 1.183 Rev.

1 Main Sections Regulatory Guide 1.183 Rev.

1 Position Analysis Basis of Conformance TEDE should be determined for the most limiting receptor at the outer

[

Conforms ee EOE a

ete ne low sos amin PZ) r boundary of the low population zone (LPZ) and should be used in ry Popl

° ae nearer with the specified breathing rates..

cetermming compliance with the dose criteria in 10 CFR 50.67 and 10 CFR Specifically for the LOCA case for ECCS recirculation leakage to the

RWST, the limiting 2-hour dose occurs well after the first 8

hours.

For the LPZ dose

location, the initial breathing rate of 3.5x10 cubic meters per second (m?/s) was therefore applied for the full 30 day duration of the accident.

For the first 8

hours, the breathing rate of persons off site should be assumed to be 3.5x10%

m?/s.

From 8

to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following the

accident, the breathing rate should be assumed to be 1.8x10 m?/s.

After that and until the end of the

accident, the rate should be assumed to be 2.3x10" l

3/s.

[

Control Room Dose Consequences

[

The TEDE analysis considered all significant sources ofl radiation that would cause exposure to Control Room l

The TEDE analysis should consider all sources of radiation that will cause personnel.

For

Callaway, the limiting Control Room exposure to control room personnel.

The applicable sources will vary from dose included:

facility to

facility, but typically will include:

e Contamination of the control room atmosphere by e

Contamination of the control room atmosphere by the intake or the intake or infiltration of the radioactive material infiltration of the radioactive material contained in the radioactive contained in the radioactive plume released from the plume released from the

facility, facility, Contamination of the control room atmosphere by the intake or Contamination of the control room atmosphere by infiltration of airborne radioactive material from areas and the intake or infiltration of airborne radioactive structures adjacent to the control room

envelope, material from the Control

Building, Radiation shine from the external radioactive plume released from:

Radiation shine from the external radioactive plume the

facility, released from the

facility, Radiation shine from radioactive material in the reactor Radiation shine from radioactive material in the containment, l

reactor containment, Radiation shine from radioactive material in systems and Radiation shine from radioactive material in Control components inside or external to the control room

envelope,

¢.g.,

Room recirculation filters and radioactive material radioactive material buildup in recirculation filters.

in the Control Building.

ources of Radiation

RG Section 4.2.2 NE NUMERICAL i

ADVISORY SOLUTIONS a Zachry Group company Table A 1. Conformance Material Releases and Radiation Levels Regulatory Guide and radiation levels determined using the sal The radioactive material releases room dose analysis should be transport, and 1.183 Rev. 1 Position release assumptions used for determining the Attachment A to NAS-2357-007 Evaluation of Changes for an Increase i with Regulatory Guide used in the control me source term, EAB and the LPZ TEDE values, unless these assumptions would result in nonconservative results for the control room.

Transport Models

_ The models used to transport

control room, and the shielding models used to Conforms radioactive material into and through the determine radiation dose

rates from external sources, should be structured to provide suitably

conservative estimates of the exposureto control Engineered Safety Features room personnel.

T Conforms Credit for engineered safety features that mitigate airborne radioactive material within the control room may be assumed. Such features may include control room isolation or pressurization, or intake or recirculation filtration. Refer to ESF l the SRP (Referenc Maintenance Criteria

_ Atmosphere Cleanup Water-Cooled Nuclear 35), for guidance. The control room design i LOCA and the pro d for other

as advantageous.

engineered safeguar

" In some cases, placing reliance on RMs can delay the ¢ to build up to concentrations equivalent to the alarm s of different radionuclide Section 6.5.1, Atmospheric emb Cleanup System, of ry Guide 1.52, Design, Testing, and t Engineered-Safety-Feature on and Adsorption Units of Light-er 2012 (Reference ized for the MHA trol room isolati diation monitors r selected accidents, dents. Several aspects of the delay for activity etpoint and the effects accident isotopic mixes ON monitor response.

for AST Re-analysis n Fuel Burnup Analysis T Conforms ~~

The radioactive material releases Revis

1.183 Rev. 1 Main Sections

ee NAS-2357-007 ion Page A-12 of A-35 Basis of Conformance and radiation levels used in the Control Room dose analyses were determined using the same source assumptions used for TEDE values.

The models used to transport term, transport, and release determining the EAB and the LPZ

T 7:

radioactive material into l and through the Control Room, and the shielding models used to determine radiation dose rates

- sources, were developed to p

- estimates of the exposure to airborne radioactive material Credit for engineered safety features from external rovide suitably conservative Control Room personnel.

that mitigate within the Control Room and Control Building were assumed as appropriate.

FSAR Table 9.4-2 presents the basis for compliance with Reg. Guide 1.52.

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NUMERICAL Attachment A

to NAS-2357-007 NAS-2357-007 ADVISORY SOLUTIONS Evaluation of Changes for AST Re-analysis Revision 0

a Zachry Group company for an Increase in Fuel Burnup Page A-15 of A-35 Table A

1.

Conformance with Regulatory Guide 1.183 Rev.

1

Main SectionsSection I

5.

ANALYSIS ASSUMPTIONS AND METHODOLOGY Regulatory Guide 1.183 Rev.

1 Position Analysis Basis of Conformance 5.1 General Considerations 5.1.1

° Analysis Quality Conforms The DBA analyses were

prepared, reviewed, and maintained per 10 CFR 50 Appendix B

and the guidance consistent with Regulatory Guide 1.183.

r The analyses discussed in this guide are reanalyses of the design basis safety analyses required by 10 CFR 50.67 or evaluations required by 10 CFR 50.34, 10 CFR Part 52, and GDC 19.

These analyses are considered to be a

significant input to the evaluations required by 10 CFR 50.92 or 10 CER 50.59 and 10 CFR Part 52.

The licensee should

prepare, review, and maintain these analyses in accordance with quality assurance programs that comply with Appendix B,

Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing

Plants, to 10 CFR Part 50.

These design basis analyses were structured to provide a

conservative set of assumptions to test the performance of one or more aspects of the facility design.

Many physical processes and phenomena are represented by conservative bounding assumptions rather than being modeled directly.

The staff has selected assumptions and models that provide an appropriate and prudent safety margin against unpredicted events in the course of an accident and compensate for large uncertainties in facility parameters, accident progression, radioactive material transport, and atmospheric dispersion.

Licensees should exercise caution in proposing deviations based on data from specific accident sequences, since the DBAs were never 1

intended to represent any specific accident sequence; the proposed deviation may not be conservative for other accident sequences.

fe RG Section 5.1.2 NUMERICAL Attachment A to NAS-2357-007 ADVISORY SOLUTIONS Evaluation of Changes for AST Re-analysis for an Increase in Fuel Burnup adZachry Group company Table A 1. Conformance with Regulatory Guide 1.183 Rev. 1 Main Sections Regulatory Guide 1.183 Rev. 1 Position

~ Credit for Engineered Safeguard Features The licensee may take credit for accident mitigation features that are classified as safety related, are required to be operable by technical specifications, are powered by emergency power sources, and are either automatically actuated or, in limited cases, have actuation requirements explicitly addressed in emergency operating procedures. However, the

licensee should not take credit for ESFs that would affect the generation of the source term described in tables 1 and 2. Additionally, the licensee should assume the single active component failure that results in the most limiting radiological consequences. Assumptions about the occurrence and timing of a loss of offsite power should be selected with the objective of maximizing the postulated radiological consequences. The licensee should consider design basis delays in the actuation of these features, especially l forfeatures that rely on manual intervention.

Analysis ns Conforms NAS-2357-007 Revision 0 Page A-16 of A-35 Basis of Conformance Credit was taken for Engineered Safeguard Features with failure assumptions to maximize the calculated doses.

Assumptions regarding the occurrence and timing ofa loss of offsite power were also selected with the objective of maximizing the postulated radiological consequences.

NUMERICAL Attachment A

to NAS-2357-007 NAS-2357-007

="

ADVISORY SOLUTIONS Evaluation of Changes for AST Re-analysis Revision 0

for an Increase in Fuel Burnup Page A-17 of A-35 a

Zachry Group company Table A

1.

Conformance with Regulatory Guide 1.183 Rev.

1

Main Sections Section Regulatory Guide 1.183 Rev.

1 Position Analysis Basis of Conformance 5.13 Assignment of Numeric Input Values l

Con orms The numeric values that were chosen as inputs to the analyses required by 10 CFR 50.67 were selected with The licensee should select the numerical values to be used as inputs to the the objective of determining a

conservative postulated dose analyses with the objective of determining a

conservative postulated dose.

dose.

In some instances, a

particular parameter may be conservative in one portion of an analysis but nonconservative in another portion of the same For a

range of

values, the value that resulted in a

analysis.

For

example, an assumption of minimum containment system conservative postulated dose was used.

spray flow is usually conservative for estimating iodine scrubbing

but, in many

cases, may be nonconservative when determining sump pH.

Sensitivity analyses may be needed to determine the appropriate value to use.

As a

conservative alternative, the limiting value applicable to each portion of the analysis may be used in the evaluation of that portion.

A single value may not be applicable for a

given parameter for the duration of the

event, particularly for parameters affected by changes in density.

For a

parameter addressed by technical specifications, the value used in the analysis should be that identified in the technical specifications.

16 Ifa range of values or a

tolerance band is specified, the value that would result in a

conservative postulated dose should be used.

If the parameter is based on the results of less frequent surveillance testing (e.g.,

steam generator nondestructive testing),

the degradation that may occur between periodic tests should be considered in establishing the analysis value.

a Applicability of Prior Licensing Basis Conforms Licensee has ensured that analysis assumptions and methods are compatible with the AST and the TEDE Licensees should ensure that analysis assumptions and methods are criteria.

compatible with the ASTs and the TEDE criteria.

l

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=

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WARN TWOldaWNN Auedwos dnoip Asyouz

RG Section 5.3 NUMERICAL Attachment A

to NAS-2357-007 ADVISORY SOLUTIONS Evaluation of Changes for AST Re-analysis Revision 0

a Zachry Group company Regulatory Guide 1.183 Rev.

1 Position Atmospheric dispersion factors (x/Q values) for the

EAB, the

LPZ, the control

room, and, as applicable, the onsite emergency response facility (i.e.,

the TSC)17 that the staff approved during initial facility licensing or in subsequent licensing proceedings may be used in performing the radiological analyses identified in this

guide, provided that such values remain relevant to the particular

accident, release characteristics that affect

plume

rise, its release

points, and receptor locations.

Licensees should ensure that any previously approved values remain accurate and do not include any misapplication of a

methodology or calculational errors in the identified values.

RG 1.145, Revision 1,

Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power

Plants, issued November 1982 (Ref.

39),

and RG 1.194, Atmospheric Relative Concentrations for Control Room Radiological Habitability Assessments at Nuclear Power

Plants, issued June 2003 (Ref.

40),

document methods for determining y/Q values.

RG 1.145 and RG 1.194 should be used if the FSAR

¥/Q values are to be

revised, or if values are to be determined for new release

points, receptor distances, or release characteristics that affect plume rise.

In addition to calculating control room y/Q

values, the modeling methodology outlined in RG 1.194 may be modified to estimate offsite y/Q values at offsite boundaries out to distances of 1,200 m

if using the procedures consistent with RG 1.249, Use of ARCON Methodology For Calculation Of Accident-Related Offsite Atmospheric Dispersion Factors (Ref.

41).

EAB

/Q values are determined for the limiting 2-hour period within a

30-day period following the start of the radioactivity release.

Control room y/Q

~

values are generally determined for initial averaging periods of 0-2 hours and 28

hours, and LPZ y/Q values are generally determined for an initial averaging period of 0-8 hours.

Control room and LPZ x/Q values are also generally determined for averaging periods of 8-24

hours, 2496

hours, and 96-720 hours for an Increase in Fuel Burnup Analysis Conforms NAS-2357-007 Page A-19 of A-35 Table A

1.

Conformance with Regulatory Guide 1.183 Rev.

1

Main Sections Basis of Conformance l

The re-calculation of atmospheric dispersion factors was performed for the EAB and LPZ using the NRC computer code PAVAN according to the guidance of Regulatory Guide 1.145 and for the control room and TSC intakes with new release points using the NRC computer code ARCON96 according to the guidance of Regulatory Guide 1.194.

The meteorological data used in the calculation were collected in accordance with Callaway site-specific measurements program and Regulatory Guide 1.23.

See FSAR Table 15A-2 for summary values and also Callaway FSAR Site Addendum Section 2.3.3 On-Site Meteorological Measurement

Programs, and Section 2.3.4 Short-Term Diffusion Estimates.

NUMERICAL Attachment A to NAS-2357-007 NAS-2357-007

- ADVISORY SOLUTIONS Evaluation of Changes for AST Re-analysis Revision 0 a Zachry Group company for an Increase in Fuel Burnup Page A-20 of A-35 Table A 1. Conformance with Regulatory Guide 1.183 Rev. 1 Main Sections Regulatory Guide 1.183 Rev. 1 Position Analysis Basis of Conformance l Seen 5.3 The source term defined in TID-14844 assumes that the entire sourceterm [l continued _ is instantaneously released into the containment atmosphere. Therefore, the maximum release rate coincides with the most conservative 0-2 hour y/Q value. In contrast, the AST is assumed to develop over specified time intervals, with the maximum release rate occurring sometime after accident initiation.

To ensure a conservative dose analysis, the period of the most adverse release of radioactive materials to the environment, with respect to doses, should be assumed to occur coincident with the period of most unfavorable atmospheric dispersion. One acceptable methodology for calculating the control room and LPZ 7/Q values is as follows. If the 0-2 hour x/Q value is

calculated, this value should be used coincident with the maximum 2-hour release to the environment. If the maximum 2-hour release occurs at the beginning of the period of releases to the environment, the 28 hour3.240741e-4 days 0.00778 hours 4.62963e-5 weeks 1.0654e-5 months y/Q value should be used for the remaining 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months of the first 8-hour time period. If the maximum 2-hour release occurs sometime after the beginning of the releases, the 2-8 hour y/Q value should be used before and after the maximum 2-hour release for a combined total of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . The 8-24, 24-96, and 96-720 hour y/Q values should similarly be used for the remainder of the release duration. Figure 2 provides examples of aligning y/Q values with the maximum 2-hour release.

The maximum two-hour TEDE at the EAB was determined by using single constant inputs for atmospheric dispersion factor and breathing rate. If the worst 2-hour dose does not occur at the beginning of the

. accident, then the timing of the maximum X/Q value for each dose location is shifted to occur at the time of maximum dose. This was performed in accordance with the example presented in Figure 2 of Section 5.3.

NUMERICAL Attachment A to NAS-2357-007 NAS-2357-007

="

ADVISORY SOLUTIONS Evaluation of Changes for AST Re-analysis Revision 0 2 Zachry Group company for an Increase in Fuel Burnup Page A-21 of A-35 A.2 RG 1.183 Rev. 1, Appendix A LOCA

[

Table A-2. Conformance with Regulatory Guide 1.183 Rev. 1 Appendix A (Loss-of-Coolant Accident)

Santon Regulatory Guide 1.183 Rev. 1, Position Analysis Basis of Conformance Appendix Assumptions for Evaluating the Radiological Consequences ofaLWR Loss-of-Coolant Accident A

Section A-1 l Source Term The inventory of fission products in the reactor core and available for release to the containment was based on the maximum full power operation with a core thermal power of 3637 MWt (102%

of 3565 MWt nominal power).

Core design parameters (enrichment, burnup, and MTU loading) are based on the cycles 29 through 31 with conservative increases in enrichment and burnup. Margin is added to the EOC core inventory, calculated with ORIGEN-S, to account for potential core design differences in future cycles. For the DBA LOCA, all fuel assemblies were assumed to be affected, and a conservatively bounding core inventory was used.

Appendix Acceptable assumptions regarding core inventory and the release of Conforms A-l radionuclides from the fuel are provided in Regulatory Position 3 of

. this guide.

Appendix Ifthe sump or suppression pool pH is controlled at values of 7 or Conforms The equilibrium pH in the sump stays above 7. See FSAR Section A-1.1

~ greater, the chemical form of radioiodine released to the 6.5.2.3.

containment should be assumed to be 95% cesium iodide (Cs]),

4.85% elemental iodine, and 0.15% organic iodide. Iodine species, including those from iodine re-evolution, for sump or suppression pool pH values less than 7 will be evaluated on a case-by-case basis. Evaluations of pH should consider the effect of acids and bases created during the LOCA event, e.g., radiolysis products.

With the exception of elemental and organic iodine and noble With the exception of elemental and organic iodine and noble gases, fission products are assumed to be in particulate form.

gases, fission products should be assumed to be in particulate form.

Chemical form Iodine released to Containment atmosphere is 95% aerosol as cesium iodide (CsI),

4.85% elemental, and

- 0.15% organic

NUMERICAL Attachment A

to NAS-2357-007 NAS-2357-007 ADVISORY SOLUTIONS Evaluation of Changes for AST Re-analysis Revision 0

a Zachry Group company for an Increase in Fuel Burnup Page A-22 of A-35 Table A-2.

Conformance with Regulatory Guide 1.183 Rev.

1 Appendix A

(Loss-of-Coolant Accident) l RG Section Section l

7

A-2 Transport in Primary Containment Regulatory Guide 1.183 Rev.

1, Position Analysis Basis of Conformance Appendix The radioactivity released from the fuel should be assumed to mix Conforms Based on relative

volumes, the release from the fuel is split A-2.1 instantaneously and homogeneously throughout the free air volume between the sprayed and unsprayed regions in containment.

of the primary containment in PWRs as it is released.

This While operation of 2

of 4

containment air coolers promotes distribution should be adjusted if there are internal compartments mixing between the two

regions, the exchange rate is that have limited ventilation exchange.

The release into the conservatively limited to two turnovers of the unsprayed region containment should be assumed to terminate at the end of the per hour.

This is in accordance with section A-2.3 (below).

early in-vessel phase.

The release to containment is assumed to terminate at the end of the early in-vessel phase.

Appendix Reduction in airborne radioactivity in the containment by natural i

Conforms Natural deposition is not credited in this analysis.

Spray removal A-2.2 deposition within the containment may be credited.

Acceptable coefficients are calculated in accordance with Chapter 6.5.2 of the models for removal of iodine and aerosols are described in Chapter SRP.

6.5.2, Containment Spray as a

Fission Product Cleanup

System, of the Standard Review Plan (SRP),

NUREG-0800 (Reference A-1) and in NUREG/CR-6189, A

Simplified Model of Aerosol

Removal by Natural Processes in Reactor Containments (Reference A-2).

The latter model is incorporated into the analysis code RADTRAD (Reference A-3).

The prior practice of deterministically assuming that a

50%

plateout of iodine is released from the fuel is no longer acceptable to the NRC staff as it is

_inconsistent_with the characteristics of the revised source terms.

RG Section Appendix A-2.3 Attachment A to NAS-2357-007 Evaluation of Changes for AST for an Increase in Fuel Burnup NW NUMERICAL a

=

ADVISORY SOLUTIONS aZachry Group company Re-analysis NAS-2357-007 Revision 0 Page A-23 of A-35 Table A-2. Conformance with Regulatory Guide 1.183 Rev. 1 Appendix A (Loss-of-Coolant Accident)

Regulatory Guide 1.183 Rev. 1, Position Analysis Reduction in airborne radioactivity in the containment by

- containment spray systems that have been designed and are maintained in accordance with Chapter 6.5.2 of the SRP may be credited. Acceptable models for the removal of iodine and aerosols are described in Chapter 6.5.2 of the SRP and NUREG/CR-5966.

This simplified model is incorporated into the analysis code RADTRAD.

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

- unsprayed compartments can be shown.

The SRP sets forth a maximum decontamination factor (DF) for elemental iodine based on the maximum iodine activity in the primary containment atmosphere when the sprays actuate, divided by the activity of iodine remaining at some time after decontamination. The SRP also states that the particulate iodine removal rate should be reduced bya factor of 10 when a DF of 50 is reached. The reduction in the removal rate is not required if the removal rate is based on the calculated time-dependent airborne aerosol mass.

There is no specified maximum DF for aerosol removal bysprays.

Conforms

\ In ace The spray removal a factor of 10 if and when a DF of 50 is reached.

Basis of Conformance ordance with Position 5.1.2, Containment spray is an ESF system, classified as safety related, required to be operable by technical specifications powered by emergency power sources, and automatically actuated.

. The mixing rate between sprayed and unsprayed regions in containment is conservatively assumed to be two turnovers of the unsprayed regions per hour.

coefficient for particulate iodine is reduced by

NUMERICAL Attachment A

to NAS-2357-007 NAS-2357-007 ADVISORY SOLUTIONS Evaluation of Changes for AST Re-analysis Revision 0

a Zachry Group company for an Increase in Fuel Burnup Page A-24 of A-35 Table A-2.

Conformance with Regulatory Guide 1.183 Rev.

1 Appendix A

(Loss-of-Coolant Accident)

Regulatory Guide 1.183 Rev.

1, Position Analysis Basis of Conformance

___Section A-2.4 Reduction in airborne radioactivity in the containment by in-Not used.

containment recirculation filter systems may be credited if these systems meet the guidance of Regulatory Guide 1.52 and Generic Letter 99-02 (Refs.

A-5 and A-6).

The filter media loading caused by the increased aerosol release associated with the revised source term should be addressed.

i Appendix Historically, reduction in airborne radioactivity in the containment Not A-2.5 by suppression pool scrubbing in BWRs should generally not be Applicable credited.

However, the staff may consider such reduction on an individual case basis.

The evaluation should consider the relative timing of the blowdown and the fission product release from the

fuel, the force driving the release through the

pool, and the potential for any bypass of the suppression pool (Reference 7).

Analyses should consider iodine re-evolution if the suppression pool liquid

__pH is not maintained greater than 7.

Appendixl Reduction in airborne radioactivity in the containment by retention Not A-2.6 in ice condensers, or other engineering safety features not addressed Applicable

above, should be evaluated on an individual case basis.

See Section 6.5.4 of the SRP (Reference A-1).

5 ee

Appendix The primary containment should be assumed to leak at the peak Conforms From Technical Specification 5.5.16.c.,

the maximum allowable A-2.7 pressure technical specification leak rate for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

For containment leakage

rate, La, at P,,

shall be 0.20%

of the

PWRs, the leak rate may be reduced after the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to 50%

containment air weight per day.

of the technical specification leak rate.

Leakage from subatmospheric containments is assumed to terminate when the containment is brought to and maintained at a

subatmospheric condition as defined by technical specifications.

After 24

hours, this is reduced to 0.10%

per day.

NUMERICAL Attachment A

to NAS-23 57-007 NAS-2357-007 rs ADVISORY SOLUTIONS valuation of Changes for A

T Re-analysis Revision 0

a Zachry Group company for an Increase in Fuel Burnup Page A-25 of A-35 Table A-2.

Conformance with Regulatory Guide 1.183 Rev.

1 Appendix A

(Loss-of-Coolant Accident)

RG Section

Regulatory Guide 1.183 Rev.

1, Position Analysis Basis of Conformance Appendix

{f the primary containment is routinely purged during power Conforms.

Only the Containment Mini-purge may be in use during power A-2.8 operations, releases via the purge system prior to containment operation.

isolation should be analyzed and the resulting doses summed with the postulated doses from other release paths.

The purge release evaluation should assume that 100%

of the radionuclide inventory l

in the reactor coolant system liquid is released to the containment at The mini-purge isolation valves automatically close within 11 the initiation of the MHA LOCA.

This inventory should be based

seconds, well before the onset of the gap release.

on the technical specification reactor coolant system equilibrium activity.

lodine spikes need not be considered.

If the purge system is not isolated before the onset of the gap release

phase, the release fractions associated with the gap release and early in-vessel phases should be considered as applicable.

100%

of the RCS maximum equilibrium activity is released to containment at initiation of the LOCA.

Section

Assumptions on Dual Containments Not A-3 oo, Applicable Section A

Assumptions on ESF System Leakage Appendix With the exception of noble

gases, all the fission products released Conforms In combination with item A-4.3

below, only iodine is released to A-4.1 from the fuel to the containment (as defined in Tables 1

and 2

of the environment.

this guide) should be assumed to instantaneously and homogeneously mix in the primary containment sump water (in PWRs) at the time of release from the core.

In lieu of this deterministic

approach, suitably conservative mechanistic models for the transport of airborne activity in containment to the sump water may be used.

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

NUMERICAL Attachment A

to NAS-2357-007 NAS-2357-007 ADVISORY SOLUTIONS Evaluation of Changes for AST Re-analysis Revision 0

a Zachry Growp company for an Increase in Fuel Burnup Page A-26 of A-35 Table A-2.

Conformance with Regulatory Guide 1.183 Rev.

1 Append A

(Loss-of-Coolant Accident)

Ro a

l

)

a l

Regulatory Guide 1.183 Rev.

1, Position Analysis Basis of Conformance Section

=

Appendix The leakage should be taken as two times the sum of the forms The operational limit of 1

gpm is doubled to 2

gpm as the basis for A-4.2 simultaneous leakage from all components in the ESF recirculation ECCS leakage to the Auxiliary.

Building.

Instead of waiting the systems above which the technical specifications, or licensee full 11.8 minutes as the earliest time to begin recirculation, commitments to item III.D.1.1 of NUREG-0737 (Reference A-8),

recirculation is conservatively assumed to start just after control would require declaring such systems inoperable.

The leakage room isolation at 62 seconds.

should be assumed to start at the earliest time the recirculation flow occurs in these systems and end at the latest time the releases from these systems are terminated.

Consideration should also be given Isolation valve seat leakage to the RWST is analyzed as a

separate to design leakage through valves isolating ESF recirculation case with a

total of 4

gpm of back-leakage to the RWST (3

gpm systems from tanks vented to atmosphere, e.g.,

emergency core below water

line, 1

gpm above water line).

cooling system (ECCS) pump miniflow return to the refueling water storage tank.

ge tank.

Appendix With the exception of

iodine, all radioactive materials in the Conforms The release from leakage to the environment is limited to iodine.

A-4.3 recirculating liquid should be assumed to be retained in the liquid

_phase.

Appendix If the temperature of the leakage exceeds 212°F, the fraction of Conforms l

With a

maximum sump temperature of approximately 265°F after A-4.4 total iodine in the liquid that becomes airborne should be assumed the beginning of recirculation in FSAR Figures 6.

.1-7 and

-8 equal to the fraction of the leakage that flashes to vapor.

This flash (Rev.

18),

the calculated flashing fraction is less than 10%.

fraction, FF, should be determined using a

constant

enthalpy, h,

process, based on the maximum time-dependent temperature of the sump water circulating outside the containment:

hy, why Ng

where, hp is the enthalpy of liquid at system design temperature l

and pressure; hm is the enthalpy of liquid at saturation conditions (14.7

psia, 212°F);

and hrg is the heat of vaporization at 212°F.

FF

NUMERICAL Attachment A

to NAS-2357-007 NAS-2357-007 Gy ADVISORY SOLUTIONS Evaluation of Changes for AST Re-analysis Revision 0

a Zachry Group company for an Increase in Fuel Burnup Page A-27 of A-35 a

ee Table A-2.

Conformance with Regulatory Guide 1.183 Rev.

1 Appendix A

(Loss-of-Coolant Accident) l 2

ae Regulatory Guide 1.183 Rev.

1, Pos ion Analysis Basis of Conformance l

If the temperature of the leakage is less than 12°F or the calculated Conforms l

The analysis assumes that 10%

of the iodine activity in the leakage flash fraction is less than 10%,

the amount of iodine that becomes becomes airborne and is available for filtration by the Aux.

airborne should be assumed to be 10%

of the total iodine activity in Building vent/exhaust system.

the leaked

fluid, unless a

smaller amount can be justified based on the actual sump pH history and area ventilation rates.

The back-leakage into the RWST represents much more controlled and well-defined environment that allows the ultimate release from the leakage to be more directly evaluated.

As avery conservative treatment of RWST

liquid, the analysis assumes the RG 1.183 conservative airborne release of 10%

of the liquid activity for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of the event to cover any potential flashing or elemental iodine regeneration within the piping.

After 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months a

very conservative airborne release of 8%

of the iodine is assumed despite the calculated flashing fraction of 5.5%

(based on sump saturated liquid enthalpy).

In accordance with NUREG/CR-5950, the temperature and pH dependent iodine re-evolution from the liquid space inside the tank and incoming sump leakage is conservatively bounded by an 8%

release of sump iodine directly to the airspace.

Additionally, the vent at the top of the tank restricts the ventilation rate of the vapor space.

l

ae ee ee Appendix The radioiodine that is postulated to be available for release to thel Conforms 97%

elemental and 3%

organic iodine was specified as the A-4.6 environment is assumed to be 97%

elemental and 3%

organic.

appropriate chemical form.

Reduction in release activity by dilution or holdup within buildings, or by ESF ventilation filtration

systems, may be credited where applicable.

Filter systems used in these applications should be Compliance with RG 1.52 is presented in FSAR Table 9.4-2.

evaluated against the guidance of Regulatory Guide 1.52

_(Reference A-5) and Generic Letter 99-02 (Reference A-6)._

Assumptions on Main Steam Isolation Valve Leakage in BWRs Not applicable

_t Section A-5

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NUMERICAL Attachment A to NAS-2357-007 NAS-2357-007

="

ADVISORY SOLUTIONS Evaluation of Changes for AST Re-analysis Revision 0 2 Zachry Growp company for an Increase in Fuel Burnup Page A-29 of A-35 A.3 RG 1.183 Rev. 1, Appendix B FHA

[

Table A-3. Conformance with Regulatory Guide 1.183 Rev. 1 Appendix B (Fuel Handling Accident)

RG

~

l a

Section l

7 Regulatory Guide 1.183 Rev. 1, Position Analysis Basis of Conformance Appends l Assumptions for Evaluating the Radiological Consequences of a Fuel Handling Accident

a a

i Section Source Term a

a nn

ten i

tener ny

ewe, Appendix Acceptable assumptions regarding core inventory and the release of Conforms The inventory of fission products in the reactor core and available Bl radionuclides from the fuel are provided in Regulatory Position 3 of

' for release to the containment was based on the maximum full this guide.

power operation with a core thermal power of 3636 MWt (102% of

3565 MWt nominal power). Core design parameters (enrichment, l

- burnup, and MTU loading) are based on the cycles 29 through 31 with conservative increases in enrichment and burnup. Margin is added to the EOC core inventory, calculated with ORIGEN-S, to

_ account for potential core design differences in future cycles. For the FHA in containment, 1.2 fuel assemblies were assumed to be affected and aconservatively bounding source term was used.

NUMERICAL ADVISORY SOLUTIONS a

Zachry Group company Table A-3.

Conformance with Regulatory Guide 1.183 Rev.

1 RG Section

é 5.

v Appendix

[

The number of fuel rods damaged during the accident should be B1.1 based on a

conservative analysis that considers the most limiting case.

This analysis should consider parameters such as the weight of the dropped heavy load or the weight of a

dropped fuel assembly (plus any attached handling grapples),

the height of the

drop, and the compression,

torsion, and shear stresses on the irradiated fuel rods.

Damage to adjacent fuel assemblies, if applicable (e.g.,

events over the reactor vessel),

should be considered.

Regulatory Guide 1.183 Rev.

1, Position Conforms Attachment A

to NAS-2357-007 Evaluation of Changes for AST Re-analysis for an Increase in Fuel Burnup Analysis NAS-2357-007 Revision 0

Page A-30 of A-35 Appendix B

(Fuel Handling Accident)

Basis of Conformance The number of damaged fuel rods is carried over from the current design basis FHA analyses.

For the FHA in the Fuel Handling

Building, 1.0 fuel assemblies were assumed to be affected and a

conservatively bounding source term was used The current design basis value of 1.0 damaged fuel assemblies for a

FHA in the FHB will be maintained for the AST analysis.

As noted on FSAR page 15.7-10, only one fuel assembly can be handled at a

time.

Likewise, FSAR page 9.1-9 explains that the spent fuel storage racks are designed to withstand the impact resulting from a

falling fuel assembly under normal loading and unloading conditions.

Unlike the free-standing conditions applicable toa partially loaded/unloaded core in an open reactor vessel in the Containment

Building, the configuration shown in FSAR Figure 9.1-2 illustrates how the spent fuel rack structure protects the fuel assemblies it contains.

For the FHA in Reactor Containment

Building, 1.2 fuel assemblies were assumed to be affected and a

conservatively bounding source term was used.

Instead of a

formal mechanical impact

analysis, the number of damaged fuel rods is carried over from the current design basis.

The number of damaged fuel rods is not directly affected by other changes to the source term or decontamination factors and so the previous value of 1.2 fuel assemblies remains applicable.

Use of this prior value is in accordance with Section 5.1.4 of the main body of RG 1.183

RG Section l

Appendix B

1.2

~

Appendix B13 NUMERICAL ADVISORY SOLUTIONS a@Zachry Group company ee Attachment A

to NAS-2357-007 Evaluation of Changes for AST Re-analysis for an Increase in Fuel Burnup NAS-2357-007 Revision 0

Page A-31 of A-35 ne ee Table A-3.

Conformance with Regulatory Guide 1.183 Rev.

1 Appendix B

(Fuel Handling Accident) l Analysis

+

Conforms Regulatory Guide 1.183 Rev.

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

All the gap activity in the damaged rods is assumed to be instantaneously released.

Radionuclides that should be considered include

xenons, kryptons,

halogens, cesiums, and rubidiums.

(The chemical form of radioio ine released from the fuel to the spent fuel pool should be assumed to be 95%

cesium iodide (CsI),

4.85 percent elemental

iodine, and 0.15 percent organic iodide.

Conforms All the gap activity is assumed to be released over two phases:

Phasel

the instantaneous release from the rising bubbles (from the start of accident to 2

hours).

Elemental iodine and organic iodine are conservatively assume to be in vapor form.

Phase 2

the protected release due to re-evolution as elemental iodine (starts at 2

hours and ends at 30 days).

Csl is conservatively assumed to completely dissociate into the pool water.

Csl (as well as Phase 1

absorbed elemental iodine within the pool) slowly re-evolve as elemental iodine into the building atmosphere.

he radioactive material available for release is assumed to be from the assemblies with the peak inventory.

The fission product inventory for the peak assembly represents an upper limit value.

The

inventory should be calculated assuming the maximum achievable operational power history and burnup.

These parameters should be examined to maximize fission product inventory.

This inventory

_l calculation should include appropriate assembly peaking factors.

l Conforms

2 The chemical form specified for the radioiodine was ee Basis of Conformance i

<iate The release fractions specified in Table 4

PWR Steady-State Fission Product Release Fractions Residing in the Fuel Rod Plenum and Gap are applied to damaged fuel.

The fuel rod is a

currently approved Westinghouse full length UOz l

fuel rod design operating up to a

maximum rod-average burnup consistent with the limits of RG 1.1 83 Rev.

l at power levels below the burnup-dependent power envelopes depicted in Figure l,

Maximum allowable power operating envelope for steady-state release fractions.

eT 95 percent

CsI, 4.85 percent elemental and 0.15 percent organic.

For Phase 1,

the elemental iodine DF is computed using Equations B-1, B-2, and B-3 and the DF for organic iodine is assumed to be 1.

For Phase 2,

the evolution removal/escape coefficient, he, is computed using Equations B-4 through B-12.

The calculation starts at 2

hours and runs to at least 30 days The inventory of fission products in the reactor core (see item B-1 above) are divided by 193 as the number of fuel assemblies in the

core, multiplied by the release fractions specified in Table 4

above, conservatively multiplied by 1.65 as the nuclear enthalpy rise hot channel

factor, Fran

Attachment A

to NAS-2357-007 NAS-2357-007 Evaluation of Changes for AST Re-analysis Revision 0

for an Increase in Fuel Burnup Page A-32 of A-35 a

Zachry Group company es SH Table A-3.

Conformance with Regulatory Guide 1.183 Rev.

1 Appendix B

(Fuel Handling Accident)

RG OT a

ee Regulatory Guide 1.183 Rev.

1, Position Analysis Basis of Conformance Section Section B-2 nn a

Appendix Ifthe depth of water above the damaged fuel is between 19 and 23 Conforms B2

feet, the decontamination factor for the elemental iodine can be A

minimum of 23 of water is required by Technical Specifications computed based ona best-estimate rod pin pressure for the limiting during fuel movement and the minimum decay time is 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

fuel rods in the reactor core at the most limiting time in life.

The time period between reactor shutdown and the movement of fuel A

conservatively high estimate for fuel Rod Internal Pressure is may be used to compute radioactive decay and reduced decay power.

1500 psig at 120°F as a

reasonably high Spent Fuel Pool water The internal gas temperature, and thus the pin pressure may be temperature, e.g.,

accommodating a

full-core offload.

determined using the limiting pool water temperature near the fuel rods and biasing these values ona full-core offload.

The elemental iodine DF is computed using Equations B-1, B-2, and B-3 and the DF for organic iodine is assumed to be 1.

Phase 1

Release

Initial Gaseous Release and Water Depth For water depths between 19 and 23

feet, the elemental iodine DF

based on pin pressure is computed using Equations B-1, B-2, and B-3.

The DF for organic iodine is assumed to be 1.

Section B-3 Appendix l

The evolution removal/escape coefficient, Ac, is computed using Conforms

_B-3 Equations B-4 through B-12.

° a

a Section Noble Gases B-4 Appendix The retention of noble gases in the water in the fuel pool orreactor

No holdup or scrubbing of noble gases is credited.

a ee Conforms B-4 cavity is negligible (i.e.,

decontamination factor of 1).

Particulate radionuclides are assumed to be retained by the water in the fuel pool or reactor cavity (i.e.,

infinite decontamination factor).

Phase 2

Release

Re-evolution Release Conservatively calculated using Equations B-4 through B-12.

2N NUMERICAL Attachment A to NAS-2357-007 NAS-2357-007 ADVISORY SOLUTIONS Evaluation of Changes for AST Re-analysis Revision 0 s

aZachry Group company for an Increase in Fuel Burnup Page A-33 of A-35

eeee Table A-3. Conformance with Regulatory Guide 1.183 Rev. 1 Appendix B (Fuel Handling Accident)

_e

__ ooo ee ro l

l a

Section Regulatory Guide 1.183 Rev. 1, Position Analysis

Basis of Conformance eo ee oT

+.

Section B-5 oot

Appendix The radioactive material that escapes from the fuel pool to the fuel Conforms For Phase 1, atwo hour release period is specified for activity B-5.1 building is assumed to be released to the environment over a 2-hour escaping the fuel pool for Phase 1. Holdup and dilution within the the initial fuel gap gas release, which accounts for FHB are not credited.

t releases from the fuel pool. The release rate is be a linear oF exponential function over this For Phase 2, the time-dependent releases from the fuel pool due to from the fuel pool due to the the re-evolution of iodine are directly from the pool to the re-evolution i

ed releases directly from the environment outside the fuel building.

Fuel Handling Accidents within the Fuel Building

ool to the environme Appendix [A reduction in the amount of radi l released from the

B-5.2 fuel pool by engineered safety feature (ESF) filter systems may be Applicable taken into account provided these systems meet the guidance of Regulatory Guide 1.52, Revision 4. Delays in radiation detection, actuation of the ESF filtration system, or diversion of ventilation flow to the ESF filtration system 1 should be determined and

__accounted for in the radioactivity release analyses.

=

a Appendix The radioactivity release from the fuel pool should be assumed to be Conforms Holdup and dilution in the FHB are not credited.

B53 ESF filtration system without mixing OF dilution in xing can be demonstrated, credit for mixing nsidered on a case-by-case basis. This magnitude of the building volume and exhaust rat i

to the environment, the

_ location of exh i

e of the pool,

recirculation ventilation system i

ors that impede stream flow between the surface of the poo plenums.

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NUMERICAL Attachment A to NAS-2357-007 NAS-2357-007 iG ADVISORY SOLUTIONS Evaluation of Changes for AST Re-analysis Revision 0 a Zachry Group company for an Increase in Fuel Burnup Page A-35 of A-35 Table A-3. Conformance with Regulatory Guide 1.183 Rev. 1 Appendix B (Fuel Handling Accident)

RG l

Section Regulatory Guide 1.183 Rev. 1, Position Analysis Basis of Conformance l l Appendix Credit for dilution or mixing of the activity released from the reactor Not

~ T Dilution and mixing within the containment are not credited.

B6.5 cavity by natural or forced convection inside the containment may be _ Applicable considered on a case-by-case basis. Such credit is generally limited to 50% ofthe containment free volume. This evaluation should consider the magnitude of the containment volume and exhaust rate, the potential for bypass to the environment, the location of exhaust plenums relative to the surface of the reactor cavity, recirculation ventilation systems, and internal walls and floors that impede stream flow between the surface ofthe reactor cavity and the exhaust plenums.

ULNRC-06922, Enclosure 8: Evaluation of Changes for AST RE-Analysis for an Increase in Fuel Burnup

Text

Title:

~

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