ML18191B015

Attachment 6: HI-2177599, Revision 0, HI-STORE Cis Facility Site Boundary Dose Rates Calculations for HI-STORM Umax System

ML18191B015
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Site:	Consolidated Interim Storage Facility
Issue date:	03/28/2017
From:	Tinker R Holtec
To:	Office of Nuclear Material Safety and Safeguards
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ML18191B009	List: ML18191B005 ML18191B006 ML18191B007 ML18191B008 ML18191B012 ML18191B013 ML18191B014 ML18191B015
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5025029 HI-2177599, Rev. 0
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2177599 Holtec Report No:

Holtec Project No:

5025 HI-5025 FOR Report Class : SAFETY RELATED NPD Sponsoring Holtec Division:

HI-STORE CIS Facility Site Boundary Dose Rates Calculations for HI-STORM UMAX System H O L T E C Holtec Center, One Holtec Drive, Marlton, NJ 08053 Telephone (856) 797- 0900 Fax (856) 797 - 0909 I N T E R N A T I O N A L ATTACHMENT 6 TO HOLTEC LETTER 5025029 Page 1 of 17

HOLTEC INTERNATIONAL DOCUMENT ISSUANCE AND REVISION STATUS1 DOCUMENT NAME:

DOCUMENT NO.:

PROJECT NO.:

CATEGORY:

GENERIC PROJECT SPECIFIC Rev.

No. 2 Date Approved Authors Initials VIR #

DOCUMENT CATEGORIZATION In accordance with the Holtec Quality Assurance Manual and associated Holtec Quality Procedures (HQPs), this document is categorized as a:

Calculation Package3 (Per HQP 3.2)

Technical Report (Per HQP 3.2)

(Such as a Licensing Report)

Design Criterion Document (Per HQP 3.4)

Design Specification (Per HQP 3.4)

Other (Specify):

DOCUMENT FORMATTING The formatting of the contents of this document is in accordance with the instructions of HQP 3.2 or 3.4 except as noted below:

DECLARATION OF PROPRIETARY STATUS Nonproprietary Holtec Proprietary Privileged Intellectual Property (PIP)

Notes ATTACHMENT 6 TO HOLTEC LETTER 5025029 Page 2 of 17 HI-STORE CIS Facility Site Boundary Dose Rates Calculations for HI-STORM UMAX System HI-2177599 5025 0

3/28/2017 623731 R.Tinker







1. This document has been subjected to review, verification and approval process set forth in the Holtec Quality Assurance Procedures Manual. Password controlled signatures of Holtec personnel who participated in the preparation, review, and QA validation of this document are saved on the company"s network. The Validation Identifier Record (VIR) number is a random number that is generated by the computer after the specific revision of this document has undergone the required review and approval process, and the appropriate Holtec personnel have recorded their password-controlled electronic concurrence to the document.

2. A revision to this document will be ordered by the Project Manager and carried out if any of its contents including revisions to references is materially affected during evolution of this project. The determination as to the need for revision will be made by the Project Manager with input from others, as deemed necessary by him.

3. Revisions to this document may be made by adding supplements to the document and replacing the "Table of Contents", this page and the "Revision Log".

Report: HI-2177599 Page: 1 Revision 0 Summary of Revisions Revision 0 - Original issue ATTACHMENT 6 TO HOLTEC LETTER 5025029 Page 3 of 17

Report: HI-2177599 _____________________________ Page: 2 Revision 0 Table of Contents Safety Analysis Summary............................................................................................................. 3

1.

Introduction........................................................................................................................... 9

2.

General Methodology........................................................................................................... 9

3.

Acceptance Criteria.............................................................................................................. 9

4.

Assumptions......................................................................................................................... 10

5.

Input Data.......................................................................................................................... 10

6.

Computer Codes.................................................................................................................. 10

7.

Analysis and Results........................................................................................................... 10

8.

Computer Files.................................................................................................................... 11

9.

Summary.......................................................................................................................... 11

10. References.......................................................................................................................... 11 10.1 Drawings....................................................................................................................... 11 Appendix A:

SAS2H/ORIGEN-S Source Terms (total of 1 page)................................... A-1 Appendix B:

MCNP Filenames and Tally Specifications (total of 2 pages)................... B-1 Appendix C:

MCNP Modeling of Casks (total of 6 pages).............................................. C-1 Attachment A: [Proprietary Information withheld per 10CFR2.390]

ATTACHMENT 6 TO HOLTEC LETTER 5025029 Page 4 of 17

Report: HI-2177599 ___________________________ Page: 3 Revision 0 Safety Analysis Summary (Source Document ID: HI-2177599R0)

Project title HI-STORE CIS Facility Report title HI-STORE CIS Facility Site Boundary Dose Rates Calculations for the HI-STORM UMAX System Name of the Site HI-STORE CIS Facility Name of the SSC (acronym for system, structure or component)

HI-STORM UMAX Version C System HI-TRAC CS Applicable NRC docket number 72-1040, 72-1051 Ref Holtec FSAR report number HI-2167374, Revision 0 Ref. Purchasing Spec. ID (if applicable)

NA Pertinent technical disciple Shielding

[Proprietary Information withheld per 10CFR2.390]

This Safety analysis summary is intended to provide the necessary information to demonstrate that the SSC identified above will render its intended safety function established within the purview of the above-referenced technical discipline under all applicable normal, off-normal and extreme environmental (accident) conditions. This safety summary is limited to demonstrating compliance in the specific area (technical discipline) of evaluation noted above and does not purport to cover other safety considerations that may apply to the subject SSC. The principal objective of this summary document is to provide a concise input to a safety significant Plant document such as a design modification package" or a multi-disciplinary comprehensive safety evaluation (such as that required under 72.212) needed to implement a planned Plant initiative in accordance with the Plant's established safety confirmation protocol.

The summary information provided below is shared with the client and is archived as a part of the Holtec proprietary document (viz., a Calculation package" which has imbedded Holtec intellectual property and is hence prohibited from external dissemination) in the Company's configuration control system. The main body of the parent document may make reference to this summary document, as appropriate, to prevent the need to repeat the same information. This document is QA validated along with its parent report and is subject to revision (and re-submittal to the client) if any significant change in the input parameters to the analysis so warrant. To serve its role as an authoritative input to the Plant, it is intended to be self-contained and entirely focused on safety. For this purpose, this summary document is organized in a series of sections to provide a succinct and concise safety assessment, as follows:

ATTACHMENT 6 TO HOLTEC LETTER 5025029 Page 5 of 17

Report: HI-2177599 ___________________________ Page: 4 Revision 0

1. Scope of Analysis This report calculates the dose rates as a function of distance for 500 loaded UMAX VVMs containing fully loaded MPC-37 canisters at the HI-STORE CIS Facility. The dose rates as a function of distance are calculated for the source term case which have the burnup, cooling time, and initial enrichment combination shown in Table 7.1.1 of Reference [8]. Using a single UMAX VVM model, dose rates as a function of distance, with occupancy factors, are calculated at distances in the range of 10 meters to 1000 m to demonstrate compliance with 10CFR72.104 [9].

Additionally a fire accident with the transfer cask, the HI-TRAC CS is considered with the density of the concrete reduced to account for degradation from the fire. The dose is calculated for a 30-day accident duration and a 100 meter distance to demonstrate compliance with 10CFR72.106 [9].

2. Acceptance Criteria The acceptance criteria for this analysis are dictated by 10CFR72.104 and 10CFR72.106 [9].

3. Computer codes and their benchmarking status Holtec International maintains an active list of QA validated computer codes on the Companys network that are approved for use in Safety significant projects.

[

Proprietary Information withheld per 10CFR2.390

]

ATTACHMENT 6 TO HOLTEC LETTER 5025029 Page 6 of 17

Report: HI-2177599 _________________________ Page: 5 Revision 0

4. Principal references

[1]

I.C. Gauld, O.W. Hermann, "SAS2H: A Coupled One-Dimensional Depletion and Shielding Analysis Module," ORNL/TM-2005/39, Version 5.1, Vol. I, Book 3, Sect. S2, Oak Ridge National Laboratory, November 2006.

[2]

I.C. Gauld, O.W. Hermann, R.M. Westfall, "ORIGEN-S: SCALE System Module to Calculate Fuel Depletion, Actinide Transmutation, Fission Product Buildup and Decay, and Associated Radiation Source Terms," ORNL/TM-2005/39, Version 5.1, Vol. II, Book 1, Sect. F7, Oak Ridge National Laboratory, November 2006.

[3]

X-5 Monte Carlo Team, MCNP - A General Monte Carlo N-Particle Transport Code, Version 5, LA-UR-03-1987, Los Alamos National Laboratory April 2003 (Revised in 2008).

[4]

USNRC Docket 72-1040, Final Safety Analysis Report on The HI-STORM UMAX Canister Storage System, Holtec Report No. HI-2115090, Revision 3.

[5]

Shielding Analysis of the HI-STORM UMAX. HI-2125194 Latest Revision. Holtec International.

[6]

USNRC Docket 71-9373, Safety Analysis Report on the HI-STAR 190 Package, Holtec Report No. 2146214, Revision 0.

[7]

Final Safety Analysis Report for the HI-STORM FW Cask MPC Dry Storage System, HI-2114830 Latest Revision. Holtec International.

[8]

USNRC Docket 72-1051, Licensing Report on The HI-STORE CIS Facility HI-2167374 R0. Holtec International.

[9]

10CFR72, Code of Federal Regulations, Licensing Requirements for the Independent Storage of Spent Nuclear Fuel, High-Level Radioactive Waste, and Reactor Related Greater than Class C Waste, USNRC, Washington, DC.

[10] Thermal Analysis of HI-TRAC CS Transfer Cask. HI-2177553 Revision 0. Holtec International.

ATTACHMENT 6 TO HOLTEC LETTER 5025029 Page 7 of 17

Report: HI-2177599 _________________________ Page: 6 Revision 0 4.1 Drawings The following drawings were used to develop the MCNP models in this report.

[ Proprietary Information withheld per 10CFR2.390 ]

5. Approach and major assumptions to insure conservative results The site boundary analysis of the HI-STORE CIS Facility with the ISFSI containing 500 loaded UMAX VVMs, can be separated into two distinct parts. The first is the generation of the radiation source terms to represent the spent nuclear fuel at the appropriate burnup and cooling time. The second part is the radiation transport simulation to calculate the dose rates near and far from a single cask.

The source terms are calculated using the computer codes SAS2H and ORIGEN-S from SCALE 5.1 [1], [2]. These codes are a widely accepted means of generating radiation source terms from spent nuclear fuel. The dose rates are calculated using the computer code MCNP5 Version 1.51 [3]. MCNP5 is a state of the art Monte Carlo code that offers coupled neutron-gamma transport using continuous energy cross sections in a full three-dimensional geometry.

The distance dose rate calculations use the same methodology employed in the site boundary MCNP cask model in Reference [4] with tallies at greater distances in the 10 meter to 1000 meter range.

The HI-STORM UMAX VVM model includes several key assumptions, as follows:

1. A standard 17x17 PWR design basis fuel assembly as defined in the HI-STORM UMAX FSAR [4], as shown in Table 5.2.1 of [4] is used for both the source term and MCNP calculations.

2. The cobalt-59 impurity level is assumed to be 0.8 g/kg for the hardware above and below the active fuel region. [

Proprietary Information withheld per 10CFR2.390

].

3. The fuel enrichment is conservatively assumed to be 5.0 wt% in the MCNP models. [

Proprietary Information withheld per 10CFR2.390

].

4. It is conservatively assumed that each loaded HI-STORM UMAX VVM contains 37 fuel assemblies each with a design basis BPRA present consistent with Section 5.2 of Reference

[4].

5. Two separate occupancy factors are assumed: 2000 hours0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> (40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> per week, 50 weeks per year) and 8760 hr, which is full time occupancy for the entire year.

ATTACHMENT 6 TO HOLTEC LETTER 5025029 Page 8 of 17

Report: HI-2177599 _________________________ Page: 7 Revision 0

6. The dose rates as a function of distance for the HI-STORM UMAX casks are calculated for the source term case presented in Table 7.1.1, which has a heat load of approximately 32 kW, consistent with the maximum allowable heat load of the HI-STAR 190 (Table 7.C.7 of Reference [6]).

7. No credit is taken for self-shielding of the UMAX VVM closure lids (i.e. one UMAX VVM closure lid above ground is not credited for shielding another UMAX VVMs dose contribution to the site boundary).

Additional modeling assumptions, modeling deviations and discussion can be found in references

[4] and [5].

The HI-TRAC CS fire accident model includes several key assumptions, as follows:

1. The material composition of the HI-TRAC CS fire accident degraded concrete density is as described in Table 7.3.1 [8] and conservatively bounds the calculated degradation due to the fire accident calculated in Appendix B of Reference [10].

2. The accident duration is assumed to last 30 days, consistent with Reference [4].

3. The distance of the HI-TRAC CS under accident conditions to the Controlled Area Boundary is assumed to be 100 meters, consistent with Reference [4].

4. The height of the tally at 100 meters is 5 feet (1.524 meters) above ground level.

6. Input data & source The input data and material compositions for SAS2H, ORIGEN-S and MCNP models are from references [4] and [5]. The UMAX VVM pitch in both the x and y directions of 17 feet (5.1816 meters) is shown in Table 1.1.1 of [8]. The ISFSI array for 500 loaded UMAX VVMs is modeled as a 20 x 25 array, crediting the additional x and y distances to the site boundary.

The distance from the nearest loaded UMAX VVM to the site boundary is a distance of 400 meters as is shown in Table 1.0.1.

The material concrete and soil compositions and densities are as described in Table 7.3.1 [8].

Additional material compositions and material properties of the storage system are provided in Subsection 5.3.2 and Table 5.3.2 in [4].

ATTACHMENT 6 TO HOLTEC LETTER 5025029 Page 9 of 17

Report: HI-2177599 _________________________ Page: 8 Revision 0

7. Results and Safety Findings The HI-STORM UMAX VVM ISFSI site boundary shielding analysis at the HI-STORE CIS Facility is presented in this report. The dose rates and annual dose (for occupancy factors of 2000 hrs/yr and 8760 hrs/yr) as a function of distance for the HI-STORM UMAX VVM ISFSI are presented in Table 7.4.3 of Reference [8]. The dose rates and annual doses are calculated at distances ranging from 100 meters to 1000 meters from the casks. Figure 7.4.3 of Reference [8]

shows ISFSI dose rates as a function of distance.

The maximum controlled area boundary dose rate (assuming an occupancy of 2,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> per year) is below the 25 mrem annual dose limit of 10CFR72.104 [9].

The nearest residence is 1.5 miles from the HI-STORE CIS Facility. The dose calculations conservatively assume a full-time resident (8760 hour0.101 days <br />2.433 hours <br />0.0145 weeks <br />0.00333 months <br />s/year) is only 1000 meters from the nearest loaded UMAX VVM. In the case of this nearest residence, the dose is calculated to be below the 25 mrem annual dose limit prescribed in 10CFR72.104 [9].

The HI-TRAC CS fire accident results are presented in Table 7.4.4 of Reference [8], with the resulting accident dose (assuming a 30-day accident duration) at 100 m from the cask showing compliance with the requirements of 10CFR72.106 [9].

ATTACHMENT 6 TO HOLTEC LETTER 5025029 Page 10 of 17

Report: HI-2177599 _________________________ Page: 9 Revision 0

1. Introduction The Introduction and scope of the analysis is presented in Section 1 of the Safety Analysis Summary.

This report is organized with a very short front section followed by detailed appendices. The actual calculations and evaluations are presented in the appendices.

2. General Methodology The analysis of the HI-STORE CIS Facility UMAX VVM ISFSI can be separated into two distinct parts. The first is the generation of the radiation source terms to represent the spent nuclear fuel at the appropriate burnup and cooling time. The second part is the radiation transport simulation to calculate the dose rates near and far from a single cask.

The radiation source terms are calculated using the SAS2H and ORIGEN-S modules from the SCALE 5.1 [1] and [2] code system from Oak Ridge National Laboratory. This is a widely accepted means of generating radiation source terms from spent nuclear fuel.

The radiation transport simulation in the HI-STORM UMAX shielding models is performed with MCNP5 1.51 [3] from Los Alamos National Laboratory. MCNP is a Monte Carlo code that offers coupled neutron-gamma transport using continuous energy cross sections in a full three-dimensional geometry.

The reader is also referred to references [4], [5] for additional discussion of the methodology and calculation of the source terms.

3. Acceptance Criteria The acceptance criteria for this analysis are dictated by 10CFR72.104 and 10CFR72.106, and are summarized here.

Normal condition requirements from 10CFR72.104 [9].

1. During normal operations and anticipated occurrences, the annual dose equivalent to an individual who is located beyond the controlled area, must not exceed 0.25 mSv (25 mrem) to the whole body, 0.75 mSv (75 mrem) to the thyroid and 0.25 mSv (25 mrem) to any other critical organ.

2. Operational restrictions must be established to meet as low as reasonably achievable (ALARA) objectives for radioactive materials in effluents and direct radiation.

ATTACHMENT 6 TO HOLTEC LETTER 5025029 Page 11 of 17

Report: HI-2177599 _________________________ Page: 10 Revision 0 Accident condition requirements from 10CFR72.106 [9]

Any individual located on or beyond the nearest boundary of the controlled area may not receive from any design basis accident the more limiting of a total effective dose equivalent of 0.05 Sv (5 Rem), or the sum of the deep-dose equivalent and the committed dose equivalent to any individual organ or tissue (other than the lens of the eye) of 0.50 Sv (50 rem). The lens dose equivalent shall not exceed 0.15 Sv (15 rem) and the shallow dose equivalent to skin or to any extremity shall not exceed 0.50 Sv (50 rem). The minimum distance from the spent fuel or high level radioactive waste handling and storage facilities to the nearest boundary of the controlled area shall be at least 100 meters.

As discussed in Subsection 5.1.2 of the HI-STORM UMAX FSAR [4], the dose requirement of 10CFR72.106 is satisfied for the design basis accident, and no additional site specific analyses are required.

4. Assumptions The major assumptions are listed in Section 5 of the Safety Analysis Summary.

5. Input Data Input data are presented in Section 6 of the Safety Analysis Summary, and appropriately referenced within each Appendix.

6. Computer Codes Computer Codes are listed in Section 3 of the Safety Analysis Summary.

7. Analysis and Results The steps to determine dose rate as a function of distance from the HI-STORM UMAX VVM are outlined in Chapter 5 of Reference [4]. Results and safety findings are presented in Section 7 of the Safety Analysis Summary.

This section provides a brief description of the contents of each Appendix. Please note that the results sections in the appendices might be page numbered separately from the Appendix.

Appendix A. SAS2H/ORIGEN-S Source Terms: This Appendix provides the neutron, fuel gamma, and Cobalt-60 hardware source terms of SAS2H and ORIGEN-S calculations.

Appendix B. MCNP Filenames: This Appendix briefly describes the nomenclature used for MCNP calculations. With that, the content of each file can be derived from the filename.

ATTACHMENT 6 TO HOLTEC LETTER 5025029 Page 12 of 17

Report: HI-2177599 _________________________ Page: 11 Revision 0 Appendix C. MCNP Modeling of Casks: The MCNP modeling methodology and tally descriptions are provided. The results for a single cask dose vs. distance, and single cask 1 meter dose rates at the top, bottom, and side mid-height various sources (neutrons, gammas, etc.) are also provided.

8. Computer Files All files are stored on the Holtec computer server in Camden, NJ [

Proprietary Information withheld per 10CFR2.390

]

9. Summary The HI-STORE CIS Facility shielding analysis of the UMAX VVM ISFSI is presented in this report.

The hourly and annual dose rates from the UMAX VVM ISFSI are presented in Table 7.4.3 of Reference [8]. The annual dose rates are specified for occupancy factors of 2000 hrs/yr and 8760 hrs/yr. The doses are calculated at distances ranging from 100 meters to 1000 m from the casks.

The results of this calculation demonstrate compliance with 10CFR72.104 [9].

Additionally, a HI-TRAC CS fire accident condition was considered in which the concrete is considered to be degraded. The dose at 100 meters from a single HI-TRAC CS for the fire accident is presented in Table 7.4.4 of Reference [8]. The results of this calculation demonstrate compliance with 10CFR72.106 [9].

10. References The references are listed in Section 4 of the Safety Analysis Summary.

10.1 Drawings The drawings are listed in Section 4.1 of the Safety Analysis Summary.

ATTACHMENT 6 TO HOLTEC LETTER 5025029 Page 13 of 17

Report: HI-2177599 _________________________

Revision 0 Appendix A: SAS2H/ORIGEN-S Source Terms (total of 1 page)

[

Proprietary Information withheld per 10CFR2.390

]

ATTACHMENT 6 TO HOLTEC LETTER 5025029 Page 14 of 17

Report: HI-2177599 _________________________

Page: B-1 Revision 0 Appendix B: MCNP Filenames and Tally Specifications (total of 2 pages)

[

Proprietary Information withheld per 10CFR2.390

]

ATTACHMENT 6 TO HOLTEC LETTER 5025029 Page 15 of 17

Report: HI-2177599 _________________________

Revision 0 Page: C-1 Appendix C: MCNP Modeling of Casks (total of 6 pages)

[

Proprietary Information withheld per 10CFR2.390

]

ATTACHMENT 6 TO HOLTEC LETTER 5025029 Page 16 of 17

Report: HI-2177599 _________________________

Page: Att. A-1 Revision 0 Attachment A

[

Proprietary Information withheld per 10CFR2.390

]

ATTACHMENT 6 TO HOLTEC LETTER 5025029 Page 17 of 17