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Attachment 3 to NWMI-2013-021, Rev. 2, Chapter 13.0 - Accident Analysis Construction Permit Application for Radioisotope Production Facility and Chapter 14.0 - Technical Specifications
	ML17221A203
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Site:	Northwest Medical Isotopes
Issue date:	08/05/2017
From:	Haass C; Northwest Medical Isotopes
To:	; Office of Nuclear Reactor Regulation
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ML17221A370	List: ML17221A199; ML17221A200; ML17221A201; ML17221A202; ML17221A203;
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	NWMI-LTR-2017-011 NWMI-2013-021, Rev. 2
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* * * * * * * * ****** * * ** * * * ** * ** * * * ** * ** * * ** * * . *. *. * . NORTHWEST MEDICAL ISOTOPES Prepared by: *

Chapter 13.0 -Accident Analysis Construction Permit Application for Radioisotope Production Facility NWMl-2013-021, Rev. 2 August2017 Northwest Medical Isotopes, LLC 815 NW gth Ave, Suite 256 Corvallis , Oregon 97330 This page intentionally left blank.

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NOflTMWHT MEDICAL ISOTDftlS NWMl-2013-021 , Rev. 2 Chapter 13.0 -Accident Analysis Chapter 13.0 -Accident Analysis Construction Permit Application for Radioisotope Production Facility NWMl-2013-021, Rev. 2 Date Published:

August 5, 2017 Document Number. NWMl-2013-021 I Revision Number. 2 Title: Chapter 13.0 -Accident Analysis Construction Permit Application for Radioisotope Production Facility Approved by: Carolyn Haass Signature:

T hi s p age int e n t i o n a ll y l eft bl a nk. NWMl-2013-021 , Rev. 2 Chapter 13.0 -Accident Analysis

.. ;.:;**NWMI .** .. ... * * *

NOtliTifWEST MEDfCAl ISOTOPES Rev Date 0 6/29/2015 1 6/26/2017 2 8/5/2017 NWMl-2013-021 , Rev. 2 Chapter 13.0 -Accident Analysis RE V I S ION HI S TORY Reason for Revision Revised By Initial Application Not requ i red Incorporate changes based on responses to NRC C. Haass Requests for Addit i onal Information Mod i fications based on comments from NRC staff C. Haass Thi s page int e ntionally left blank. NWMl-2013-021 , Rev. 2 Chapter 13.0 -Accident Analysis CONTENTS NWMl-2013-021 , Rev. 2 Chapter 13.0 -Accident Analysis 13.0 RADIOISOTOPE PRODUCTION FACILITY ACCIDENT ANALYSIS ...............................

I 3-1 13. l Accident Analysis Methodology and Preliminary Hazards Analysis .............................

13-3 13.1.1 Methodologies Applied to the Radioi s otope Production Facilit y Integrated Safety Analysis Process ..........................................

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........ I 3-3 13.1.1.1 Accident Likelihood Categories, Consequence Severity Categories , and Risk Matrix ..............

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I 3-5 13.1.1.2 Accident Conseq uence Analysis .........................................

............ 13-7 13.1.1.3 What-If and Structured What-If ...................

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.. 13-7 13.1.1.4 Hazards and Operabilit y Study Method ..........................................

13-8 13.1.1.5 Event Tree Analysis ......................................

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13-8 13.1.1.6 Fault Tree Ana l ys i s ...................................................

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13-8 1 3.1.1.7 Failure Modes and Effects Analysis .................................

.............. 13-8 13.1.2 Accident-Initiating Events ...................................................

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.... 13-8 13.1.3 Preliminary Hazards Analysis Result s ............

.............................................. 13-12 13.1.3.1 Hazard Criteria ..........

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13-12 13.1.3.2 Radioisotope Production Fac ilit y Accident Sequenc e Eval uation ...................

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.... 13-1 3 13.2 Analysis of Accidents with Radiological and Critica lit y Safety Consequences

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13-38 13.2. l Reserved .........................................

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13-39 13.2.2 Liquid Spills and Sprays with Radiolo g ical and Criticality Safety Consequences

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13-39 13.2.2.1 Initial Conditions

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13-39 I 3 .2.2.2 Identification of Event Initiating Conditions

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........ 13-44 I 3.2.2.3 Description of Accident Seque nc es ..............................................

I 3-44 13.2.2.4 Function of Components or Barriers .............................................

13-44 13.2.2.5 Unmitigated Likelihood

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........ 13-45 I 3.2.2.6 Radiation Source Term ........................................................

........ 13-45 1 3.2.2.7 Eva lu ation of Potential Radiological Consequences

...................... 13-47 13.2.2.8 Identification ofltems Relied on for Safety a nd Associated Functions

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13-50 13 .2.2.9 Mitigated Estimates

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13-54 13.2.3 Target Dissolver Offgas Accidents with Radiolo g ical Consequences

............ 13-54 13 .2.3.1 Initial Conditions

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................. I 3-55 1 3 .2.3.2 Identification of Event Initiating Conditions

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..................... 13-56 13.2.3.3 Description of Accident Sequences

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............................... 13-56 13.2.3.4 Function of Compo n ents or Barriers ...............

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....... 13-56 13.2.3.5 Unmitigated Like lihood ...............

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13-56 I 3.2.3.6 Radiation Source Term .................................

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.......... I 3-57 13.2.3.7 Eva luation of Potential Radiological Conseq u ences ......................

13-57 13.2.3.8 Identification ofltems Relied on for Safety and Associated Funct ion s ...................

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13-58 1 3.2.3.9 Mitigated Estimates

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NORTHWUT MEDtCAl ISOTOPE S 13.2.4 13.2.5 13.2.6 13.2.7 NWMl-2013-021 , Rev. 2 Chapter 13.0 -Accident Analysis Leaks into Auxiliary Services or Systems with Radiological and Criticality Safety Consequences

................................................................... 13-59 13.2.4.1 Initial Conditions

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13-59 13.2.4.2 Identification of Event Initiating Conditions

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13-63 13.2.4.3 Description of Accident Sequences

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........... , .............. 13-64 13.2.4.4 Function of Components or Barriers ..................

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13-64 13 .2.4.5 Unmitigated Likelihood

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.......... 13-64 13.2.4.6 Radiation Source Term ...............................

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13-65 13.2.4.7 Evaluation of Potential Radiological Consequences

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..... 13-65 13.2.4.8 Identification ofltems Relied on for Safety and Associated Functions

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..... 13-65 13.2.4.9 Mitigated Estimates

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13-69 Loss of Power ..................................

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13-69 13.2.5.1 Initial Conditions

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13-69 13.2.5.2 Identification of Event Initiating Conditions

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13-69 13.2.5.3 Description of Accident Sequences

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13-69 13.2.5.4 Function of Components or Barriers .............................................

13-70 13.2.5.5 Unmitigated Likelihood

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13-70 13.2.5.6 Radiation Source Term ................................................................

13-70 13.2.5.7 Evaluation of Potential Radiological Consequences

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13-70 13.2.5.8 Identification ofltems Relied on for Safety and Associated Functions

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....... 13-70 Natural Phenomena Events ............

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13-71 13.2.6. l Tornado Impact on Facility and Structures, Systems, and Components

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.. 13-71 13.2.6.2 High Straight-Line Winds Impact the Facility and Structures, Systems, and Components

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13-72 13 .2.6.3 Heavy Rain Impact on Facility and Structures, Systems, and Components

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13-72 13.2.6.4 Flooding Impact to the Facility and Structures, Systems , and Components

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........... 13-73 13.2.6.5 Seismic Impact to the Facility and Structures, Systems, and Components

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13-73 13.2.6.6 Heavy Snow Fall or Ice Buildup on Facility and Structures, Systems, and Components

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13-74 Other Accidents Analyzed .............................

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............. 13-75 13.2.7.1 Items Relied on for Safety for Radiological Accident Sequences (S .R.) .................................

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13-85 13.2. 7.2 Items Relied on for Safety for Criticality Accident Sequences (S.C.) ...........................

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13-87 13.2.7.3 Items Relied on for Safety for Fire or Explosion Accident Sequences (S.F.) .................

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13-93 13.2.7.4 Items Relied on for Safety for Natural Phenomena Accident Sequences (S.N.) ............

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... 13-93 13.2.7.5 Items Relied on for Safet y for Man-Made Accident Sequences (S.M.) ..................................................

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13-94 13.2.7.6 Items Relied on for Safety for Chemical Accident Sequences (S.CS.) ................

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........... 13-94 13-ii NWMl-2013-021 , Rev. 2 Chapter 13.0 -Accident Analysis 13.3 Analysis of Accide nts w i t h Hazar d o us C h emica l s .............................................

......... 13-95 13.3.1 C h e m ical Bu m s from Co n ta min ate d So lu t i o n s D u r ing Sa m p l e Ana l ys i s ....... 13-95 13.3.1.1 C h e m ica l Acci d e nt D escriptio n ........................................

............ 13-95 1 3.3.1.2 C h e mi ca l Acc id e nt Conseq u e n ces ................................................

13-95 1 3.3.1.3 C h e mi cal P rocess Co n tro l s .......................................

.................... 13-95 1 3.3.1.4 C h e mi cal P rocess S ur vei ll a n ce R e qu ireme nt s ...............................

13-9 5 13.3.2 N i tr i c Acid F um e R e l ease ....................

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......... 13-96 1 3.3.2.1 C h e mi ca l Acci d e nt D escript i on ..........................................

.......... 13-96 1 3.3.2.2 C h e mi ca l Acc id e nt Conseq u e n ces ..........................

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13-96 1 3.3.2.3 C h e mi ca l Process Co n trols ...........................

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13-96 1 3.3.2.4 C h e mi ca l Process S ur ve i lla n ce R e qu iremen t s ...............................

13-96 1 3.4 R efe r e n ces ...............................................................................

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13-9 7 13-iii Figure 13-1. Figure 13-2. FIGURES NWMl-2013-021, Rev. 2 Chapter 13.0 -Accident Analysis Integrated Safety Analysis Process Flow Diagram .....................................................

13-4 Unmitigated Off-Site Dose of Dissolver Product Spray Leak Accident ....................

13-49 TABLES Table 13-1. Likelihood Categories

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13-5 Table 13-2. Qualitative Likelihood Category Guidelines

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13-5 Table 13-3. Radioisotope Production Facility Consequence Severity Categories Derived from 10 CFR 70.61 ............................................................................................................

13-6 Table 13-4. Radioisotope Production Facility Risk Matrix .................

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13-6 Table 13-5. Radioisotope Production Facility Preliminary Hazard Anal ys is Accident Sequence Category Designator Definitions

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13-9 Table 13-6. Crosswalk ofNUREG-1537 Part l lnterim Staff Guidance Accident Initiating Events versus Radioisotope Production Facility Preliminary Hazards Analysis Top-Level Accident Sequence Categories

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.................. 13-9 Table 13-7. Radioisotope Production Facility Preliminary Hazards Analysis Primary Process Nodes and Subprocesses (2 pages) ...............

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13-10 Table 13-8. Crosswalk of Radioisotope Production Facility Preliminary Hazards Analysis Process Nodes and Top-Level Accident Sequence Categories

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...... 13-12 Table 13-9. Adverse Event Summary for Target Fabrication and Identification of Accident Sequences Needing Further Evaluation (4 pages) .....................................................

13-14 Table 13-10. Adverse Event Summary for Target Dissolution and Identification of Accident Sequences Needing Further Evaluation (3 pages) .....................................................

13-18 Table 13-11. Adverse Event Summary for Molybdenum Recovery and Identification of Accident Sequences Needing Further Evaluation (3 pages) ...............................

....... 13-21 Table 13-12. Adverse Event Summary for Uranium Recovery and Identification of Accident Sequences Needing Further Evaluation (4 pages) .....................

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13-24 Table 13-13. Adverse Event Summary for Waste Handling and Identification of Accident Sequences Needing Further Evaluation (2 pages) .....................................................

13-28 Table 13-14. Adverse Event Summary for Target Receipt and Identification of Accident Sequences Needing Further Evaluation (2 pages) .....................................................

13-30 Table 13-15. Adverse Event Summary for Ventilation System and Identification of Accident Sequences Needing Further Evaluation

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13-32 Table 13-16. Adverse Event Summary for Node 8.0 and Identification of Accident Sequences Needing Further Evaluation (5 pages) ................................

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13-33 Table 13-17. Bounding Radionuclide Liquid Stream Concentrations (4 pages) .............................

13-40 Table 13-18. Source Term Parameters

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13-46 Table 13-19. Release Consequence Evaluation RASCAL Code Inputs .........................................

13-48 13-iv

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HOITifWEST MEDICAL ISDTl>n.S NWMl-2013-021 , Rev. 2 Chapter 13.0 -Accident Analysis Table 13-20. Uranium Separations Feed Spray R e lea se Consequence Summary at 100 Meters ..... 13-49 Table 13-21. Maximum Boundin g Inventor y ofRadioiodine

[Proprietar y Information]

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13-55 Table 13-22. Target Dissolver Offgas Accident Total Effective Dose E quivalent..

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........ 13-58 Table 13-23. Bounding Radionuclide Liquid Stream Concentrations (4 pages) ......................

....... 13-60 Table 13-24. Analyzed Accidents Sequences (9 page s) ..........................................................

....... 13-75 Table 13-25. Summary of Item s Relied on for Safety Identified by Accident Analyses (2 pages) ............

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....... 13-84 Table 13-26. Accident Sequence Category Definition s ..........................

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13-85 13-v TERMS Acronyms and Abbreviations 99 Mo molybdenum-99 99 mTc technetium-99m 235 U uranium-235 2 41 Am americium-241 AAC augmented administrative control AC administrative control ACI American Concrete Institute AEC active engineered contro l AEGL Acute Exposure Guideline Level AISC American Institute of Steel Construction ALARA as low as reasonably achievable ALOHA areal locations of hazardous atmospheres ARF airborne re l ease fraction ASCE American Society of Civil Engineers CDE committed dose equiva l ent CEDE committed effective dose eq u ivalent CFR Code of Federa l Regulations DAC derived air concentration DOE U.S. Department of Energy DOT U.S. Department of Transportation DR damage ratio EDE effective dose equivalent EOI end of irradiat ion ET A event tree ana l ysis FEMA Federal Eme r gency Manage m ent Agency FMEA fai l ure modes and effects ana l ysis FT A fault tree ana l ysis HAZOP hazards an d operability HEGA high-efficiency gas adsorption HEPA high-efficiency particulate air HIC high-integrity canister HN0 3 nitric acid HV AC heating, venti l ation, and air conditioning IBC Internationa l Building Code IROFS items relied on for safety IRU iodine remova l unit ISA integrated safety analysis ISG Interim Staff Guidance IX ion exchange LEU low enriched uranium LPF l eak path factor MAR material at risk Mo molybdenum MURR University of Missouri Research Reactor NaOH so dium hydroxide NDA nondestructive assay NWMl-2013-021 , Rev. 2 Chapter 13.0 -Accident Analysis NIOSH National Institute for Occupational Safety and Hea lth NO x nitrogen oxide 13-vi NOAA NRC NWMI NWS OSTR osu P&ID PEC PFD PHA PMP QRA RASCAL RF RPF RSAC SNM SSC ST TCE TEDE u U.S. UN NWMl-2013-021 , Rev. 2 Chapter 13.0 -Accident Analysis National Oceanic and Atmospheric Administration U.S. Nuclear Regulatory Commission Northwest Medical Isotopes , LLC Nationa l Weather Service Oregon State University TRIGA Reactor Oregon State University piping and instrumentation drawing passive e n gineered control process flow diagram preliminary hazards ana l ysis probable maximum precipitation quantitative risk assessment Radiological Assessment System for Consequence Analysis respirab l e fraction Radioisotope Production Facility Radiological Safety Ana l ysis Code special nuclear materia l structures, systems, and components source term trichl oroethy lene total effective dose equiva l ent uranium United States uranyl n i trate 13-vii NWM I ...... *

NORTHWEST MlDtCAL ISOTOPE S Units o c O F Ci Cm ft ft 3 g hr in.2 kg km km 2 L lb m M m 3 mg m1 mi2 mil mm mrem oz ppm rem sec Sv wk wt% yr degrees Celsius degrees Fahrenheit cune centimeter feet cubic feet gram hour square inch kilogram kilometer square kilometer liter pound meter molar cubic meter milligram mile square mile thousandth of an inch minute millirem ounce parts per million roentgen equivalent man second sievert week weight percent year 13-viii NWMl-2013-021, Rev. 2 Chapter 13.0 -Accident Analysis NWMl-2013-021, Rev. 2 Chapter 13.0 -Accident Analysis 13.0 RADIOISOTOPE PRODUCTION FACILITY ACCIDENT ANALYSIS The proposed action is the issuance of a U.S. Nuclear Regulatory Commission (NRC) Construction Permit and Operating License under Title 10, Code of Federal Regulations , Part 50 (10 CFR 50) "Domestic Licensing of Production and Utilization Facilities," and provisions of 10 CFR 70, "Domestic Licensing of Special Nuclear Material," and 10 CFR 30, "Rules of General Applicability to Domestic Licensing of Byproduct Material ," that would authorize Northwest Medical Isotopes , LLC (NWMI) to construct and operate a molybdenum-99 (99 Mo) Radioisotope Production Facility (RPF) at a site located in Columbia, Missouri. The RPF is being designed to have a nominal operational processing capability of one batch per week of up to [Proprietary Information]. The primary mission of the RPF will be to recover and purify radioactive 99 Mo generated via irradiation of low-enriched uranium (LEU) targets in off-site non-power reactors.

The purified 99 Mo will be p ac kaged and transported to medical industry users where the radioactive decay product, technetium-99m (99 m Tc), can be employed as a valuable resource for medical imaging. This section analyzes potential hazards and accidents that could be encountered in the RPF during operations involving special nuclear material (SNM) (irradiated and unirradiated), radioisotope recovery and purification, and the use of hazardous chemicals relative to these radiochemical processes.

Irradiation services and tran spor tation activities are not analyzed in thi s chapter. This chapter evaluates the various processing and operational activities at the RPF , including:

Receiving LEU from U.S. Department of Energy (DOE) Producing LEU target materials and fabrication of targets Packagin g and shipping LEU targets to the universit y reactor network for irradiation Returning irradiated LEU targets for dissolution , recovery , and purification of 99 Mo Recovering and recycling LEU to minimize radioactive, mixed , and ha za rdous waste generation Treatin g/packag ing wastes generated b y RPF process steps to enable transport to a disposal site Chapter Organization Section 13.1 describe s hazard and acc ident analysis methodologies applied to the RPF integrated safety analysis (ISA) (Section 13.1.1). Section 13.1.2 identifies the accident initiating events, and Section 13.1.3 summarizes the results of the RPF preliminar y hazards analysis (PHA) (NWMI-2015-SAFETY-001, NWMI Radioi sotope Production Facility Preliminary Ha zards Analysis).

The PHA discu ss ion in Section 13.1.3 identifies the accident scenarios that required further evaluation. Section 13.2 pres ents analyses of radiological and criticality accidents, including:

Section 13.2. l (Reserved)

Section 13.2.2 discusses spills and spray accidents Section 13.2.3 discusses dissolver offgas accidents Section 13.2.4 discusses leaks into auxiliary systems accidents Section 13.2.5 discusses loss of e lectr ical power Section 13.2.6 discusses natural phenomena accidents Section 13.2. 7 identifie s the additional accident sequences evaluated and associated items relied on for safety (IROFS) 13-1 NWMl-2013-021, Rev. 2 Chapter 13.0 -Accident Analysis Section 13.3 presents bounding accidents involving hazardous chemicals.

The data presented in the following subsections are based on a comprehensive PHA , conservative assumptions, draft quantitative risk assessments (QRA), and scoping calculations.

These items provide an adequate basis for the construction application. 13-2 NWMl-2013-021, Rev. 2 Chapter 13.0 -Accident Analysis 13.1 ACCIDENT ANALYSIS METHODOLOGY AND PRELIMINARY HAZARDS ANALYSIS 13.1.1 Methodologies Applied to the Radioisotope Production Facility Integrated Safety Analysis Process This section describes methodologies applied to the RPF ISA. The ISA process comprises the PHA and the follow-on development and completion of QRAs to address events and hazards identified in the PHA as requiring further evaluation.

T he ISA process flow diagram is provided Figure I 3-I. The ISA process (being adapted for this application) consists of conducting a PHA of a system using a combination of written process descriptions, process flow diagrams (PFD), process and instrument drawings (P&ID), and supporting calculations to identify events that could lead to adverse consequences.

Those adverse consequences are evaluated qualitatively by the ISA team members to identify the likelihood and severity of consequences using guidance on event frequencies and consequence categories consistent with the regu l atory guidelines.

Each event with an adverse consequence that involves licensed material or its byproducts is evaluated for r i sk using a risk matrix that enables the user to identify unacceptable intermediate-and high-consequence risks. For the unacceptable intermediate-and high-consequence risks events, the IROFS developed to prevent or mitigate the consequences of the events and an event tree analysis are used to demonstrate that the risk can be reduced to acceptable frequencies through preventative or mitigative IROFS. Fault trees and failure mode and effects analysis can be used to (I) provide quantitative fai lur e analysis data (fai lure frequencies) for use in the event tree ana l ysis of the IROFS , as necessary, or (2) quantitatively analyze an event from its basic initiators to demonstrate that the quantitative failure frequency is already highly unlikely under normal standard industrial conditions, thus not needing the application ofIROFS. Once the IROFS are developed, management measures are identified to ensure that the IROFS failure frequency used in the analysis is preserved and the IROFS are able to perform their intended function when needed. The following subsections summarize the RPF ISA methodologies.

13-3 NWMI ...... * * ! NORTHWEST MmtCAL ISOTOf'ES Design and Safety Functions ISATeam Deve l op p rocess descr i p t ions, P FDs, P&IDs Identify preliminary hazards and consequences (radiological, criticality, chemical, fire, e xte rn a l) using regulato ry g u ides where applicable l Develop CSAs, FHA, and other support documents Initiate ISA process by collecting preliminary data Perform PHA on facility operations Categorize events for likelihood, consequence, and risk Indeterminate, high, or intermediate risk? Yes+ Perform QRA to quantitatively evaluate risk and identify IROFS High or intermediate risk event? Yes Identify "accident sequence" and 1------++

develop IROFS and basis for each in complete QRA Develop PSAR, ISA summary, technical specifications Document identified low-risk events (no IROFS) No Start Phase 1 deve l opment of ...--. IROFS boundary definition packages for each IROFS Complete Phase 1 development of IROFS boundary definition packages I ISA team review and I recommendation for approval Management NWMl-2013-021 , Rev. 2 Chapter 13.0 -Accident Analysis Design and Engineering Functions Design function deve l opme n t of IROFS specifications/

conceptual drawings NRCReview approval of ISA basis NRC review of document r----------

+---------1* license submit to NRC application 1 cm_r Figure 13-1. Integrated Safety Analysis Process Flow Diagram 13-4 NWMl-2013-021 , Rev. 2 Chapter 13.0 -Accident Analysis 1 3.1.1.1 Accident Likelihood Categories, Consequence Severity Categories, and Risk Matrix Ta ble 13-1 shows the accident likelihood categories applied to the RPF ISA process. T able 13-2 shows qualitative guidelines for applying the likelihood categories from Table 13-1. Table 13-3 shows accident consequence severity categories from Table 13-1. Likelihood Categories 10 CFR 70.61, "Performance Requirements." Table 13-4 s ho ws the RPF risk matrix , which i s a product of the likelihood and consequence severity categories from Table 13-1 and Table 13-3, respectively.

Not unlik ely Unlikely Highly unlikely 3 2 Event frequency limit More than I 0-3 events per year Between I 0-3 and I 0-5 events per year Less than 10-5 per events per year Table 13-2. Qualitative Likelihood Category Guidelines 11.* Initiator 3 An event initiated by a human error 3 An event initiated by failure of a process system processing corrosive materials 3 An event initiated by a fire or explosion in areas where combustible s or flammable materials are present 3 An event initiated by failure of an active control system 3 A damaging seismic event 3 A damaging high wind event 3 A spill of material 3 A failure of a proce ss variable monitored or unmonitored by a control system 3 A valve out of position or a valve that fai l s to seat and isolate 3 Most standar d industrial component failures (valves, senso rs , safety devices, gauges, etc.) 3 An adverse chemical reaction ca u sed by improper quantities ofreactant s, o ut-of-date reactants, of-specification reaction e nvir onment, or the wron g reactants are u se d 3 Most external man-made events (until confirmed using an approved method) 2 An event initiated by the failure of a robust passive design feature with no significant internal or externa l cha ll e n ges applie d (e.g., spontaneo u s rupture of an a ll-welded dry nitr ogen system pipe operating at or below design pres s ure in a clean, vibratio n-fr ee environment) 1-2 An adverse chemical reaction when proper quantities of in-date chemicals are reacted in the proper environment Natural phenomenon s uch as tsunami, volcanos, and asteroids for the Missouri facility site 13-5 NWMl-2013-021 , Rev. 2 Chapter 13.0 -Accident Analysis Table 13-3. Radioisotope Production Facility Consequence Severity Categories Derived from 10 CFR 70.61 *iii Consequence category Workers Off-site public Environment High consequence Intermediate consequence Low co n sequence 3 2

Radiological dose* > I Sv (I 00 rem)

Airborne , radiologically contaminated nitri c acid

Radio l ogical dose* > 0.25 Sv (25 rem)

Toxic intake> 30 mg so lubl e U > 170 ppm nitri c acid (AEGL-3 ,

10-min exposure limit) Airborne, contaminated nitric acid > 24 ppm nitric acid (AEGL-2, 60-min exposure limit)

Unshie ld edb nuclear critica li ty . Radiological dose* between . Radiological dose* 0.25 Sv (25 rem) and I Sv between 0.05 Sv (5 rem) (100 rem) and 0.25 Sv (25 rem) . Airborne, radiologically

Airborne, contaminated contaminated nitric acid nitric acid > 0.16 ppm > 43 ppm nitric acid (AEGL-2, nitric acid (AEGL-1, 10-min exposure limit) 60-min exposure limit) Accidents with l ower Accide nt s with lower radiological, c h emical, and/or radio l og i ca l , chemica l , toxicological exposures than those and/or toxicological above from lic e n sed material and exposures than those above byproducts of lic ensed material from li censed material and byproducts of licensed material 24-hr radioactive release > 5,000 x Table 2 of 10 CFR 20, 0 Appendix B Radiological releases producing lower effects than those li sted above from licensed material Source: I 0 CFR 70.61 , " Performance Requirements

," Cod e of F e d e ral R egu l a tions, Office of the Federal Register, as amended.

As total effective dose equ i va l ent. b A shiel d e d cr i t i cality acc id ent is a l so cons id ered a high-consequence event. c IO CFR 20, "Standards for Protection Against Radiation ," Cod e of F e d e ral Regulati o n s, Office of the Federa l Register , as amended. AEGL Ac ut e Exposure Gu id eline Leve l. u = uranium. Tab l e 13-4. Radioisotope Production Facility Risk Matrix Severity of consequences High consequence (Consequence category 3) Intermediate consequence (Consequence category 2) Low consequence (Consequence category 1) Highly unlikely (Likelihood category 1) Risk index = 3 Acceptable risk Risk index = 2 Acceptable ri sk Risk index = 1 Acceptable risk Likelihood of occurrence Unlikely (Likelihood category 2) Ri sk index = 6 U n accepta bl e ri sk Risk index= 4 Acceptab l e risk Risk index = 2 Acceptable risk 13-6 { Not unlikely (Likelihood Category 3) Risk index = 9 Unaccepta bl e risk Risk index = 6 Unacceptab l e risk Risk index = 3 Acceptable risk

... ;. NWMI *::**:*** 0 0 NORTHWEST MEDICAL ISOlWES NWMl-2013-021, Rev. 2 Chapter 13.0 -Accident Analysis 13.1.1.2 Accident Consequence Analysis T he ISA process requires an understanding of the source terms and consequences of an adverse event to determine if the event is low , intermediate, or high consequence, as compared with the hazard criteria identified in Table 13-4. NUREG/CR-6410, Nuclear Fuel Cycle Facility Accident Analysis Handbook , offers methodologies to calculate the quantitative consequences of events. For simplicity and prudent expenditure ofresources, the RPF ISA assumes a worst-case approach using a few bounding evaluations o f events that are identified through either: Calculations (e.g., the source term and radiation doses caused by contained material in the system) Studies of representative accidents (e.g., comparison of accidental criticalities in industry with processes similar to those at the RPF) Bounding release calculations using approved methods (e.g., using RASCAL [Radiological Assessment System for Consequence Analysis]

to model bounding facility releases that affect the public) Reference to nationally recognized safety organizations (e.g., u se of Acute Exposure Guideline Levels [AEGL] from the U.S. Environmental Protection Agency to identify chemical exposure limits for each consequence category)

Approved methods for evaluation of natural and man-made phenomenon and comparison to the design basi s (e.g., calculation of explosive damage potential from the nearest railroad line on the facility)

Accident consequence analysis results are identified before or during the ISA process following preliminary reviews of the processes , and as the process hazard identification phase identifies new potential hazards. Initial hazards identified by the preliminary reviews include: High radiation dose to workers and the public from irradiated target material during processin g High radiation dose due to accidental nuclear criticality Toxic uptake of licensed material by workers or the public durin g processing or accidents Fires and explosions associated with chemical reactions and use of combustible materials and flammable gases Chemical exposures associated with chemicals used in processing the irradiated target material External events (both natural and man-made) that impact the facility operations 13.1.1.3 What-If and Structured What-If RPF activities that will be mainly conducted by personnel using a sequence of actions to affect a process were evaluated using what-if or structured-what-if techniques to identify process hazards that can lead to unacceptable risk. These methods allow free-form evaluation of the activity by ISA team members , which can be enhanced by using a list of key guidewords addressing the specific hazards identified in the facility (e.g., the deviations to normal condition criticality safety controls like spacing, mass , moderation

material spills; wrong materials, place, or time for activities; etc.). The key words for each structured what-if evaluation are documented in the PHA. 13-7 13.1.1.4 Hazards and Operability Study Method NWMl-2013-021, Rev. 2 Chapter 13.0 -Accident Analysis For processes that are part of a processing system and have well-defined PFDs and/or P&IDs , the more structured hazards and operability (HAZOP) approach was used. This method systematically evaluates each node of a process using a set of key words that enables the team to systematically identify adverse changes in the process and evaluate those changes for adverse consequences.

The key words for each evaluation are documented in the PHA. 13.1.1.5 Event Tree Analysis An event tree analysis (ET A) is a bottoms-up , logical modeling technique for both success and failure that explores responses through a single initiating event and lays a path for assessing probabilities of the outcomes and overall system analysis.

ET A uses a modeling technique referred to as an event tree , which branches events from one single event using Boolean logic. The ISA uses ETA in two primary ways. For those initiating events where the ISA team is uncertain of the likelihood of reaching the adver s e consequence , the method can be used during the QRA to follow the sequence of events leading to an adverse consequence and thus quantify the adverse event's frequenc y given the initiator.

ETA is also used in the QRA process to demonstrate that the IROFS, s elected to prevent an adverse event , reduce the failure frequency to a level that satisfies the performance requirements (e.g., the frequency of a high-consequence event is reduced to highly unlikely).

13.1.1.6 Fault Tree Analysis Fault tree analysis (FT A) is a top-down , deductive failure anal y sis in which an undesirable system state is analyzed with Boolean logic to combine a series of lower-level initiating events. The process enables the user to understand how systems can fail , identify the best ways to reduce risk , and/or determine event rates of an accident or a particular system-l evel functional failure. This analysis method is mainly used in QRAs when a failure frequency or probability is needed for a specific component, an IROFS, or some other complex process. 13.1.1.7 Failure Modes and Effects Analysis Failure modes and effects analysis (FMEA) is an inductive reasoning (forward logic) single point of failure analysis that is also quantitative in nature. FMEA involves reviewing as many components , assemblies , and subsystems as possible to identify failure modes , along with associated causes and effects. For each component, the failure modes and associated effects on the rest of the system are recorded in a FMEA worksheet.

This is an exhaustive analysis technique that can be used to evaluate the reliability of a complex , active engineered control (AEC) type ofIROFS. 13.1.2 Accident-Initiating Events Each of the following accident initiating events was included in the PHA. Loss of power as an accident event is discussed further in Section 13.2.5. Criticality accident Loss of electrical power External events (meteorological, seismic, fire , flood) Critical equipment malfunction Operator error Facility fire (explosion i s included in this category)

Any other event potentially related to unique faci lit y operations 13-8 The PHA (NWMl-2015-SAFETY-OOI) identifies and categorizes accident sequences that require further evaluation. Table 13-5 defines t h e level accident sequence notation u se d in the RPF PHA. Table 13-6 provide s a crosswalk between the PHA top-level accident sequence categories and the NUREG-1537 , Guidelines for Preparing and R ev i ew in g Applications for the Licensing of No nPo wer R e actor s -Format and Co ntent, Part 1 Interim Staff Guid a nce (ISG) accident initiatin g events listed above. As noted at the bottom of Table 13-6 , PHA accident sequences invo l ve one or more of the NUREG-1537 Part 1 ISG accident initiating event categories, as noted b y ./ in the corresponding table ce ll , but the PHA accident sequences themsel ves are not necessari l y initiated b y the ISG accident initiating event. Table 13-6 NWM l-2013-02 1 , R ev. 2 C h a p t er 13.0 -A c ciden t Ana lys i s Tab l e 13-5. Radio i soto p e Pro du c ti o n Fac ili ty Pre limin a r y Hazar d Analys i s Acci d e n t Seq u e n ce Category Designator De fi n i tions PHA top-level accident sequence categorya S.C. S.F. S.R. S.M. S.N. S.CS. Definition Criticality Fire or explosion Radiological Man-made Natural phenom e n a Chemica l safety

The a lpha category d es ign a t o r is fol low e d in th e PHA b y a two-digit number "XX" that r efe r s t o the s p ec ific accident se qu e n ce (e.g., S.C.01 , S.F.07). Specific accident se qu ences a re di scusse d in Sect ion s 1 3.1.3 a nd 13.3. PH A = pr e limi nary h aza rd a n a l ys i s. s h ows how PHA accident sequences correspond with ISG accident initiatin g events, and demonstrat es that the PHA considers the full range of accident events identified in the ISG. Table 13-6. Crosswa l k ofNUREG-153 7 Part I I n teri m Staff G u idance Acci d ent Init i ating Events versus Ra d ioisotope Pro du ction Fac ili ty Pre limin ary Hazar d s Analysis Top-Leve l Accide n t Seq u ence Ca t egories NUREG-1537a Part 1 ISG accident initiating event category C riticality accident Loss of electrical power Exte rnal eve nt s (meteorological , se i s mi c, fire, flood) Critical equipment malfunction Operator error Facility fire (explosion is included in this category)

Any other event potentially related to unique faci li ty operations PHA Top-Level Accident Sequence Categoryb

,/ ,/ ,/ ,/ ,/ ,/ ,/ ,/ ,/ ,/ ,/ ,/ ,/ ,/ ,/ ,/ ,/ ,/ ,/ ,/ ,/ ,/

NURE0-1537 , Guide lin es for Pr e paring and R ev i ew in g App li ca tions for the Li ce nsin g of No n-P ower R e actors -Forma t and Con t e nt , Part I , U.S. Nuclear R egula t ory Co mmi ss ion , Office ofNuclear R eac tor R egu l at i o n , Wa s hin gto n , D.C., Fe b ru ary 1996. h PHA acc ident seque nces inv o l ve o n e or more of the NURE0-1 537 Part I I SO accident initiatin g eve nt ca te go ri es, as noted by a n ./ in th e cor r es pondin g tabl e ce ll , but the PHA se quence s th emse l ves a r e not necessarily initiat ed b y th e IS O accident initiatin g event. I SO = lnt e rim StaffO uidanc e. PHA = pr e limin ary h aza rd a n a l ysis. 13-9 NWM I ...... *

NOmfWlST MEOtC.Al ISOTOP£S NWMl-2013-021, Rev. 2 Chapter 13.0 -Accident Analysis The RPF PHA subdivides the RPF process into eight primary nodes based on facility design documentation.

Table 13-7 lists the RPF primary nodes and corresponding subprocesses , as identified in the PHA. Table 13-7. Radioisotope Production Facility Preliminary Hazards Analysis Primary Process Nodes and Subprocesses (2 pages) Node no. Node name Subprocesses encompassed in node 1.0.0 2.0.0 3.0.0 4.0.0 Target fabrication process Target dissolution process Molybdenum recovery and purification proce s s Uranium recovery and recycle process

Fresh uranium receipt and storage

Fresh uranium dissolution

Uranyl nitrate blending and feed preparation

Nitrate extraction

Recycled uranyl nitrate concentration

[Proprietary Information]

[Proprietary Information]

[Proprietary Information]

[Proprietary Information]

[Proprietary Information]

[Proprietary Information]

Uranium s crap recovery

Target assembly, l oading , inspection , quality checking, verification, packaging and storage * [Proprietary Information]

[Proprietary Information]

Primary process offgas treatment

Fission gas retention

Feed preparation

First stage recovery

First stage purification preparation

First stage purification

Second stage purification preparation

Second stage purification

Final purification adjustment

9 9 Mo preparation for shipping

Impure uranium lag storage

First-cycle uranium recovery

Second-cycle uranium purification

Product uranium lag storage

Other support (storage vesse ls , transfer lines, solid waste handling for resin bed replacement) 13-10 NWM I ...... *

NORTHWEST MH>>CAl ISOTOPES NWMl-2013-021, Rev. 2 Chapter 13.0 -Accident Analysis Table 13-7. Radioisotope Production Facility Preliminary Hazards Analysis Primary Process Nodes and Subprocesses (2 pages) Node no. Node name Subprocesses encompassed in node 5.0.0 6.0.0 7.0.0 8.0.0 99 Mo HEPA Waste handling system process Target receipt and disassembly proces s Ventilation system Natural phenomena, man-made external events, and other facility operations

Liquid waste s tora ge

High dose liquid waste volume r eduction

Co nd ensate storage and recycling

Concentrated high dose liquid waste storage/preparation

Low dose liquid waste vo lum e reduction and s t orage

Liquid waste solidification

So lid waste hand! in g

Waste encapsu l ation

TCE solvent re c lamati on

Mixed waste accumulation

Cask receipt and target unloading

Target Inspection

Target disassembly

[Proprietary Information]

Target disassembly stations

Gaseous fission product control * [Proprietary Information]

Empty target hardware handling * (No subprocesses identified in PHA. Ventilation system provid es cascading pressure zones , a common air s uppl y system with makeup air as n ecessary , heat recovery for preconditioning incoming air, and HEPA filtratio n.)

Natural phenomena

Man-made external events

Chemical storage and preparation areas

On-site vehicle operation

General storage, utilities , and maintenance activities

Laboratory operations

Hot cell s upport activities

Waste storage operations including packaging and shipment molybdenum-99 high-efficiency particulate air. PHA preliminary hazards anal ys is. TCE = trichloroethylene. Table 13-8 shows a crosswalk that identifies the applicability of RPF PH A top-level accident sequence categories to the primary proces s nodes. The information in this table is referenceable to Table 13-6 and ultimatel y shows the relationship between the PHA process nodes and the NUREG-1537 Part 1 ISG accident initiating event categories via the PHA top-level accide nt sce nario categories.

13-11 NWMl-2013-021, Rev. 2 Chapter 13.0 -Accident Analysis Table 13-8. Crosswa l k of Radioisotope Production Facility Preliminary Hazards Analysis Process Nodes and Top-Leve l Accident Sequence Categories PHA Top-Level Accident Sequence Category *-m1111* Tar get fabrication (Node 1.0.0) ,, ,, ,, Target dissolution (Node 2.0.0) ,, ,, ,, Molybdenum recovery and ,, ,, ,, purification (Node 3.0.0) Uranium recovery and recycle ,, ,, ,, (Node 4.0.0) Waste handling system ,, ,, ,, (Node 5.0.0) Target receipt and disassembly

,, ,, (Node 6.0.0) V e ntilation system (Node 7.0.0) ,, ,, ,, Natural phenomena, man-made ,, ,, ,, external events, and other facility operations (Node 8.0.Q) Note: The ../ in a table cell indicates t hat the accident se quen ce categor y applies to th e proc ess nod e. If it does not , th e cell is bl a nk. PHA = preliminar y h aza rds a n a l ys is. 13.1.3 Preliminary Hazards Analysis Resu l ts This section presents the radiological, criticality , and chemical hazards that could result in high or intermediate consequences.

13.1.3.1 Hazard Criteria Methodologies and hazard criteria are identified in Section 13.1.1. Numerou s hazards are present during the handling and processing the materials in the RPF. The target material is fissile LEU consisting of uranjum enriched up to 19.95 weight percent (wt%) uranium-235 (2 35 U). Tills material presents a criticality accident hazard in the processes that involve high concentrations of uranium. Both 10 CFR 50 and 10 CFR 70 require that accidental nuclear criticalities be prevented using the double-contingency principle , as defined in ANSI/ ANS-8.1 , Nu cl ear Criticality Safety in Op eratio n s with Fissionable Material Outside Rea cto rs. The RPF separates 99 Mo from among the fission products in the irradiated LEU target material.

The fission products , including 99 Mo , present a high-dose hazard that must be properly contained and shielded to protect workers and the public. Radiation protection standards are given in 10 CFR 20 , " Standards for Protection Against Radiation ," and its appendices.

The RPF also uses high concentrations of acids , caustics, and oxidizers, both se parate from and mixed with licensed material , that present chemical hazards to workers. The Nationa l Institute for Occupational Safety and Health (NIOSH) provides acute exposure guidelines (CDC, 2010) that evaluate chemical exposure hazards to workers and the public from chemicals and toxic l icensed material.

The facility can also be impacted b y various internal and external man-made and natural phenomena events that have the potential to damage structures , systems , and component s (SSC) that control the l icensed material , thereby leading to intermediate-and high-consequence events. 13-12 Known and credited safe ty features for normal operations include: NWMl-2013-021 , Rev. 2 Chapter 13.0 -Accident Analysis The hot cell shielding boundary, credited for shielding workers and the public from direct exposure to radiation (an expected operational hazard) The hot cell confinement boundaries, credited with confining fissile and high-dose solids, liquids , and gases, and controlling gaseous releases to the environment Administrative and passive engineered design features that control uranium batch size, volume, geometry and interaction are credited for maintaining critically safe (i.e., subcritical) configurations during normal operations with fissile material.

The RPF PHA identifies abnormal operation event initiators that require further evaluation for IROFS to ensure that the double-contingency principle is satisfied.

13.1.3.2 Radioisotope Production Facility Accident Sequence Evaluation A structured what-if analysis was used to evaluate RPF system nodes where operators are primarily involved with licensed material manipulations.

All proce ss syste m nodes were analyzed using a HAZOP approach with special emphasis on criticality, radiological, and chemical safety hazards. Fire safety i ssues a re addressed in every node and addressed generally in Node 8.0.0. Fire safety issue s include the explosive hazard associated with hydrogen gas generation via radiolytic decomposition of water in p roce ss solutions and due to certain chemical reactions encountered during dissolution processes.

Most hot cell processing areas contain ve r y few combustible materials , either transient or fixed. The RPF PHA has identified adverse events listed in Table 13-9 through Table 13-16. Adverse events are identified as: Standard industrial events that do not involve licensed material Acceptable accident sequences that satisfy performance criteria by bein g low consequence and/or low frequenc y Unacceptable accident sequences that require further evaluation via the QRA process An accident sequence number was assigned to each accident initiator that re s ult s in the sa me , or similar , bounding accident se quence result and consequence.

The same accident se quence d esignator can appear in multiple nodes. (Ta ble 13-5 provides definitions of accident sequence category designators.)

13-13

i*:h-NWMI ...... ' *

NOmtWlST M£DtcAl ISOTOPf:S NWMl-2013-021, Rev. 2 Chapter 13.0 -Accident Analysis Table 13-9. Adverse Event Summary for Target Fabrication and Identification of Accident Sequences Needing Further Evaluation (4 pages) Bounding accident PHA item numbers description Consequence Accident sequence 1.1. l.1 , 1.1.1.2, 1.6.1.1, Operator double batches Accidental criticality S.C.02, Failure of 1.8.1.l , 1.8.2.1 , and 1.8.3.1 allotted amount of material issue -Too much fissile administrative control on (fresh U, scrap U, [Proprietary mass in one location mass (batch limit) during Information], target batch) may become critical handling of fresh U, into one location or container scrap U, [Proprietary during handling Information], and targets 1.1.1.3 Supplier ships greater than Accidental criticality S.C.01 , Failure of site 20 wt% m u to site issue-Too much mu enrichment limit put into a container or solution vessel, exceeding assumed amounts 1.1. l.6 , I. I. I. 7 , 1.6.1.2 , Operator handling various Accidental criticality S.C.03, Failure of 1.6.1.4 , 1.8.1.2, 1.8.1.3, containers of uranium or i ss ue -Too much administrative control on 1.8.1.6 , 1.8.2.2, 1.8.2.3, batches of uranium . . interaction limit durin g uranium mass m one 1.8.3.2, 1.8.3.3, 1.8.3.4 , and components brings two location handling of fresh U, 1.8.3.5 containers or batche s clo se r scrap U, [Proprietar y together than the approved Information], and targets interaction control di s tance 1.2.1.1, 1.2.1.11, 1.2.1.14, Failure of safe geometry Accidenta l criticality S.C.04, Spill of fissile 1.2.1.25, 1.3.1.1, 1.3.1.6, confinement from fissi l e solution not material from safe 1.3.1.ll, 1.3.1.17, 1.4.1.19, confined in safe geometry system 1.4.1.20, 1.4.1.2 I, I .4.1.23, geometry confinement I .4.2.6, I .4.2. l 0, 1.4.2.15, 1.4.3.14, 1.4.3.26, 1.4.3.31, 1.4.4.1, 1.4.4.6, 1.4.4.10, 1.4.4.15, 1.5.1.21, 1.5.1.23, 1.5.1.26, 1.5.2.16, 1.7.1.l, 1.7.1.ll, 1.7.1.14, 1.7.1.25 , 1.9.1.1, 1.9.1.6, 1.9.1.10, and 1.9.1.15 1.2.1.2 and 1.7 .1.2 Uranium-containing so lution Accidental criticality S.C.05, Leak of fissile leak s out of safe geometry from fissile solution not solution into heatin g/ confinement into the confined in safe coolin g jacket on vessel heatin g/cooling jacketed space geometry 1.2.1.3, 1.4.3.33, 1.4.3.34, Uranium solution is Accidental criticality S.C.07 , Leak offissile and 1.7.1.3 transferred via a leak between from fissile solution not solution across auxiliary the process system and the confined in safe system boundary (chilled heater/cooling jackets or coils geometry water or steam) on a tank or in an exchanger 13-1 4 NWMl-2013-021 , Rev. 2 Chapter 13.0 -Accident Analysis Table 13-9. Adverse Event Summary for Target Fabrication and Identification of Accident Sequences Needing Further Evaluation ( 4 pages) Bounding accident PHA item numbers description Consequence Accident sequence 1.2.1.8 , 1.3.1.4, 1.4.1.15, Failure of safe geometry Accidenta l criticality S.C.19, Fai l ure of 1.4.2.4, 1.4.3.18, 1.4.4.4, dimension caused by from fissile solution not passive design feature -1.5.1.20 , 1.5.2.11, 1.7.1.8 , configuration management confined in safe Component safe and 1.9.1.4 (insta ll ation, maintenance), geometry geometry dimension internal or external event 1.2.1.12, 1.3.1.9, 1.4.2.8, Tank overflow into process Accidental criticality S.C.06, Overfill ofa tank 1.4.4.8, 1.4.5.4, 1. 7.1.12, and ventilation system issue -Fissile so lution or component causing 1.9.1.8 entering a system not fissile so lution entering necessarily designed for the process vessel fissile solutions ventilation system 1.3.1.2, 1.4.2.2 , 1.4.4.2, and Uranium precipitate or other Accidenta l criticality S.C.20, Fai lur e of 1.9.1.20 high uranium solids from fissile solution not concentration limits -accumu l ate in safe geometry confined to safe Precipitation of uranium ve s sel geometry and in safe geometry tank interaction controls within a ll owable concentrations 1.2.1.26, 1.3.1.7, 1.5.1.3, and Uranium solution backflows Accidental criticality S.C.08, Fissile solution 1.5.2.5 into an auxiliary support issue -Fissile solution backflow into an sys tem (water line , purge line , entering a system not auxiliary system at a fill chemical addition line) due to nece ssa rily designed for point boundary various causes fissile solutions 1.4.1.6 , 1.4.1.12, and 1.4.1.16 Failure of safe geometry Accide nt a l criticality S.C.11, Fissi l e material confinement due to from fiss il e solution not contamination of inadvertent transfer to confi n ed in safe contactor regeneration U-bearing solution across a geomet r y aqueous waste stream -boundary into non-favorable boundary to un safe geometry geometry system 1.4.3.1, 1.4.3.9 , 1.4.3.19 , Failure of safe geometry Accidental criticality S.C.09, Fissile material 1.4.3.21, 1.4.5.9, and 1.4.5.11 confinement due to from fissile solution not contamination of inadvertent transfer to confi ned in safe evaporator condensate

-U-bearing solution across a geometry boundary to unsafe boundary into non-favora ble geometry sys tem geometry 1.6.1.3 Failure of safe geometry Accidental criticality S.C.12, Wash of confinement du e to from fissile so lu tion not [Proprietary Information]

inadvertent transfer to confined in safe with wrong reagent U-bearing so luti on across a geometry co ntamin ating wash boundary into non-favorable so luti on with fissile U; geometry boundary to un safe geometry sys tem 1.1.1.11 Dusty surface generated Potential exposure to S.F.01, Pyrophoric fire during s hipping on uranium workers due to airborne in uranium metal pieces spontaneously ignites uranium generation due to pyrophoric nature of uranium 13-15 NWMI ...... ! *

NomtWtST MEDtcAL ISOTDPH NWMl-2013-021, Rev. 2 Chapter 13.0 -Accident Analysis Table 13-9. Adverse Event Summary for Target Fabrication and Identification of Accident Sequences Needing Further Evaluation ( 4 pages) Bounding accident PHA item numbers description Consequence Accident sequence 1.2.1.6 , 1.2.1.11 , 1.7.1.6 , and H y dr og en buildup in tanks or 1. 7 .1.11 sys t e m , l ea din g t o ex pl os i ve co n ce ntr at i o n s 1.4.1.17 , 1.4.1.21 , and 1.4.1.23 1.6.1.6 , 1.6.1.9, and 1.6.1.1 2 1.6.1.8 1.2.1.l I , 1.2.1.14 , 1.4.1.1 7, 1.4.1.1 9, 1.4.1.2 0 , 1.4.1.2 I , 1.4.1.23 , 1.4.2.6 , 1.4.3.1 4, 1.4.3.2 6 , 1.4.3.3 I , 1.4.3.32, 1.7.1.11 , 1.7.1.14 , an d 1.9.1.6 1.2.1.1 I , 1.2.1.12 , 1.2.1.14, 1.2.1.25 , 1.3.1.1 , 1.3.1.6 , 1.3.1.11 , 1.3.1.17 , 1.4.1.17 , 1.4.1.18 , 1.4.1.19 , 1.4.1.21 , 1.4.2.1 , 1.4.2.6 , 1.4.2.8 , 1.4.2.10 , 1.4.2.15 , 1.4.3.14 , 1.4.3.26, 1.4.3.31, 1.4.4.6 , 1.4.4.10 , 1.4.4.15, 1.5.1.21 , 1.7.1.11, 1.7.1.14, 1.7.1.25 , 1.9.1.1 , 1.9.1.6, 1.9.1.8 , 1.9.1.10 , and 1.9.1.15 Fire in process s ystem containing high concentration uranium spreads the uranium A ir inl ea k age int o th e r e du c t io n furn ace durin g H z pur ge cy cl e or H z in l eakage int o r ed u c ti o n furn ace b efo r e inertin g wi th nitr og en c an l ea d t o an ex pl os i ve mi x tur e in th e pr ese n ce of an igniti o n s ourc e Loss of cooling of exhaust or fire in the reduction furnace lead s to high temperatures in downstream ventilation component and accelerated release of adsorb radionuclides Hi g h co n ce ntr a ti o n ur an ium s oluti on is s pra ye d fro m the system, ca u s in g high a irb o rn e radi oac ti v ity High concentration uranium solution i s spilled from the system 13-16 Ex pl osion l ea din g to radi o l og i c al and critic a li ty c onc e rn s Radiological and criticality issue -Radiological airborne relea s e of uranium and uncontrolled s pread of uranium outside sa fe g eometry confinement Acc id e ntal critic a li ty i ss u e -U n co ntr o ll ed s pr ea d of uranium o ut s id e sa f e geo m etry co n fi n e ment Radiological issue -Potential accelerated release of high-dose radionuclides to the stack (worker and public exposure)

R a di o l og i cal r e l ease of uranium s olution s pra y th at r e mains s u s p e nd e d in the a ir , ex po s in g wo rk e rs o r th e publi c Potential radiological e x posure to workers from contaminat e d solution S.F.0 2, Acc umulation of flamm a ble g a s in tan ks or sys t e m s S.F.07 , Fire in nitrate extraction system -flammable solvent with uranium S.F.0 3 , Hy dro ge n d e t o n a ti o n in r e duc t i o n furna ce S.F.04 , High temperature damage to process ventilation system due to loss of cooling in reduction furnace exhaust or fire in reduction furnace S.R.0 3, So lution s p ray rel ease p o t e ntiall y cr ea ting a irborn e ur an ium a bo ve D AC limit s S.R.01 , contaminated s olution spill NWM I ...... *

NOWTtfWUT 11£DtCAl tSOTOPU NWMl-2013-021 , Rev. 2 Chapter 13.0 -Accident Analysis Table 13-9. Adverse Event Summary for Target Fabrication and Identification of Accident Sequences Needing Further Evaluation ( 4 pages) Bounding accident PHA item numbers description Consequence Accident sequence 1.2.1.21, 1.2.1.22, 1.4.5.13, 1.7.1.21, and 1.7.1.22 1.3.1.16 and 1.4.1.24 1.8.3.7 uranium-235. Boiling or carryover of steam or high concentration water vapor into the primary ventilation system, affecting retention beds from partial or complete lo ss of cooling system capab ili ties High-dose solution (fai lure of the uranium recovery process) results in high-do se radionuclides entering th e first stage of proc essi n g uranium [Proprietary Information] (eventually handled by the worker) Loading limit s are not adhered to by the operators or the closure requirements are no t satisfied, and the cask does not provide the containment or shielding function that it is designed to perform m u DAC H 2 !RU deriv e d air concentration. h ydrogen gas. iodine removal unit. 13-17 Radiological release from retention bed s Potentially high radiological exposure to wo rkers Hi g h-dose to workers or the public from improperly s hi elded cask S.R.04 , Liquid enters process vesse l ventilation system damaging IRU or retention beds relea s ing retained radionuclides S.R.05, High-dose solution enters the UN blending and storage tank S.R.28 , Target or waste shipping cask not loaded or secured accordin g to procedure, leadin g to per so nnel exposure PHA u UN process h aza rds a n alys is. uranium. uranyl nitrate.

NWMl-2013-021 , Rev. 2 Chapter 13.0 -Accident Analysis Table 13-10. Adverse Event S ummary for Target Dissolution and Identification of Accident Sequences Needing Further Evaluation (3 pages) Bounding accident PHA item numbers description Consequence Accident sequence 2.1.1.1, 2.1.1.11 , Fai lur e of safe geometry Accidental critica li ty from S.C.04 , Failure of 2.1.1.13, 2.1.1.17' confinement fissile so lu tion not confined in confinement in safe 2.2.1.5 , 2.2.1.12, safe geometry geometry; spill of fissile 2.2.1.15, 2.3.6.5 , material solution 2.3.6.12, and 2.3.6.1 3 2.1.1.2 Uranium-containing Accidental criticality from S.C.05, Leak of fissile so lution leak s out of safe fissile so lution not confined in solution in to geo metry confinement into safe geometry heating/cooling jacket the heatin g/coo ling jacketed on vessel space 2.1.1.3 Uranium so lution is Accidental criticality from S.C.07, Leak of fissile transferred via a leak fissile so lution not confined in so lut ion across auxiliary between the process system safe geometry system boundary and the heater/cooling (chilled water or steam) jackets or coils on a tank or in an exchanger 2.1.1.8, 2.2.1.11, and Fai lure of safe geometry Accidental criticality from S.C.19, Failure of 2.3.6.11 dimen s ion fissile solution not confined in passi ve d esign feature; safe geo m etry component safe-geo m etry dimension 2.1.1.12, 2.1.1.15 , and Fai lur e of safe-geometry Accidental critica li ty from S.C.13, Fissile solution 2.3.1.4 confinement fissile solution not confined in enters the NO x scrubber safe geo metr y where high uranium solution is not intended 2.1.1.14 and 2.3.4.14 Tank overflow into process Accidental criticality issue -S.C.06, System ventilat ion system Fissile solution entering a overflow to process sys tem not necessarily de signe d ventilation involvin g for fissile so lutions fissile material 2.3.4.11 Uranium enters carbon Accidental critica li ty from S.C.24 , Bui ld-up of high retention bed dryer where it fissi l e material or so lu tion not uranium particulate in can mix with co nd e n sate to confined in safe geometry the carbon retention bed form a fissile solutio n dryer system 2.1.1.33 and 2.1.1.34 Uranium so lution backflows Accidental criticality and high S.C.08, Sys tem into an auxiliary su pp ort radiological dose -High-dose backflow into auxiliary syste m (wate r lin e, pur ge and fissile so lution entering a s upport sys tem line, chemical addition lin e) system not necessarily de signe d du e to vario u s causes for fissile so lution s that exist outside of hot ce ll walls 13-18 NWM I ...... *

NOITHWEST MEDM:.Al tsOTOP£S NWMl-2013-021 , Rev. 2 Chapter 13.0 -Accident Analysis Table 13-10. Adverse Event Summary for Target Dissolution and Identification of Accident Sequences Needing Further Evaluation (3 pages) Bounding accident PHA item numbers description Consequence Accident sequence 2.1.1.18, 2.3.1.21, Hydrogen build-up in tanks Explosion leading to S.F.02, Accumulation of 2.3.2.21, 2.3.3.24 , or system leading to radiological and criticality flammable gas in tanks 2.3.4.3, and 2.3.5.5 exp l osive concentrations co n cerns or systems 2.3.4.20, 2.3.5.2, A fire develops through Radiological issue -Potential S.F.05, Fire in a carbon 2.3.5.6, 2.3.5.10, and exothermic reaction to accelerated release of high-dose retention bed 2.3.5.13 contaminants in the carbon radionuclides to the stack retention bed and rapidly (worker and public exposure) releases accumulated gaseous high-d ose radionuclides 2.1.1.1 , 2.1.1.2, High-do se and/or high-Potential radio lo g ic al ex posure S.R.01, Radiological 2.1.1.11 , 2.1.1.13, concentration uranium to workers from high-dose release in the form of a 2.1.1.17, 2.2.1.5, so luti on is s pilled from the and/or high uranium-liquid spi ll of high-do se 2.2.1.12 , 2.2.1.15, system co nt aminated so lution and/o r high uranium 2.3.6.5 , 2.3.6.12, and concentration so luti on 2.3.6.1 3 2.1.1.3 High-dose solution is Radiological exposure to S.R.13, High-dose transferred via a leak workers and the public from solution leaks to chilled between the process system high-radiological dose not water or steam and the heater/cooling contained in the hot cell condensate system jackets or coils on a tank or containment or confinement in an exchanger boundary 2.1.1.11 , 2.1.1.1 7, Sp ill l eading to spray-typ e Radiological d ose from S.R.0 3, Spray of product 2.2.1.1 5, and 2.3.6.13 relea se, causing airborn e airborne s pray of produ ct so lut ion in hot ce l l area radioactivity above DA C so luti on from sy stem s limit s for exposure 2.1.1.23, 2.1.1.26, Carryover of high vapor High airborne radionuclide S.R.04 , Carryover of 2.1.1.27, 2.3.4.1, content gases or entrance of r elease, affecting workers and hea vy vapo r or solution 2.3.4.12, and 2.3.4.17 so lu tions into the process the public into the process ventilation header can cause ventilation header poor performance of the causes downstream retention bed materials and failure of retention bed, re l ease radionuclides relea sing radionuclides 2.3.1.17, 2.3.1.22, A spi II of l ow-dose Potential radio l ogical do se to S.R.02, Spi ll oflow-2.3.1.24 , 2.3.2.17, condensate occurs for a workers and the public from do se conde n sate 2.3.2.22, 2.3.2.24, variety of rea so ns from the sp ill ed liquid 2.3.3.8, 2.3.3.20 , confinem e nt tanks or vessels 2.3.3.27 , 2.3.4.3 , 2.3.4.5 , 2.3.4.6, and 2.3.4.8 2.3.3.1, 2.3.3.2, 2.3.3.3 , High flows through the IRU Potential radiological dose to S.R.06, High flow 2.3.3.6 , 2.3.3.12, increases the release of the workers and the public from through IRU causes 2.3.3.13, 2.3.3.16, retained iodine and iodine above regulatory limits premature release of 2.3.3.17 , 2.3.3.23, increases the high-dose high-dose iodine gas 2.3.4.13, 2,3.5.1 , concentration ofthis gas in 2.3.5.6, 2.3.5.8 , and the stack 2.3.5.l 0 13-19

i*:h NWMI ...... * * ! NORTHWtST MEIHCAL ISOTOP£S NWMl-2013-021, Rev. 2 Chapter 13.0 -Accident Analysis Table 13-10. Adverse Event Summary for Target Dissolution and Identification of Accident Sequences Needing Further Evaluation (3 pages) Bounding accident , PHA item numbers description Consequence Accident sequence 2.3.3.15 and 2.3.5.8 2.3.3.22 and 2.3.5.8 2.3.4.4 , 2.3.4.5, and 2.3.4.6 2.3.4.11 2.3.3.28, 2.3.4.19, 2.3.5.9 , 2.3.4.15, and 2.3.5.11 2.3.4.16, 2.3.5.5, and 2.3.5.12 2.1.1.33 and 2.1.1.34 Low temperatures in the IRU inlet gas stream drives release of iodine from the unit Liquid and water vapor in the IRU inlet gas stream drives release of iodine from the unit Loss of vacuum pumps in the dissolver offgas treatment system leads to pressure buildup inside the process and potential release of radionuclides from the sys t em up stream Uncontrolled loss of media and contact with a liquid with potential for premature release of the adsorbed iodine Using the wrong retention media (IRU or carbon beds) or u sing saturated media with potential for ineffective adsorp tion of high-d ose gaseous radionuclides An event causes damage to the structure holding the retention media, and retention media is released to an uncontrolled environment High-do se process so luti on backflows into an auxiliary s upport system (water line , purge lin e, chemical ad dition line) due to various ca u ses DAC IRU derived air co ncent ration. iodine remova l unit. Potential radiological dose to workers and the public from iodine above regulatory limits Potential radiological do se to workers and the public from iodine above regulatory limits Potential radiological dose to workers and the public from spi ll ed liquid Potential radiological dose to workers and the public from iodine above regu l atory limits Potential radio lo gica l dose to workers and the public from radionuclides above regulatory limit s Potential radiological dose to workers and the public from radionuclides above regulatory limits High radiological dose -High dose process so lution enters a system that exits outside of the hot ce ll walls nitrogen oxide. S.R.07, Loss of temperature control on the IRU l eads to premature release of high-dose iodine S.R.04, Liquid/high vapor in the IRU leads to premature release of high-dose iodine S.R.08, Loss of vacuum pumps S.R.09, Loss ofIRU media to downstream dryer S.R.10, Wrong retention media added to b ed or saturated retention media S.R.09, Breach of an IRU or retention bed resulting in release of the media S.R.11, System backflow of high-dose solution into an auxiliary s upport system and outside the h ot cell boundary NO x PHA process hazards ana l ysis. 13-20 NWMl-2013-021 , Rev. 2 Chapter 13.0 -Accident Analysis Table 13-11. Adverse Event Summary for Molybdenum Recover y and Identification of Accident Sequences Needing Further Evaluation (3 pages) Bounding accident PHA item numbers description Consequence Accident sequence 3.3.1.24 Higher radiation dose due to Hig h er l ocalized dose in N I A hold-up accumulation or hot ce ll boundary transient batch differences (unocc upi ed by workers) 3.2.3.7, 3.2.4.7, 3.4.3.7, 3.4.4.7, C hemical spills of Standard industrial N I A 3.6.3.7, and 3.6.4.7 nonradiological ly accident -Chemical conta minated bulk exposure (not involving che mical s licen se d material) to workers 3. 7.4.5 and 3. 7.4.6 Dropped cask or cask Standard industrial N I A component during loading accident -Worker injury or handling 3.7.4.2, 3.7.5.2, and 3.7.5.3 Mo product is ex posed with Potential dose to the N I A -Addressed by no shielding as the result of public and/or enviro nment DOT packaging and an accident, shipme nt due to relea se o r transportation mishap, or shipment mishandling of Mo regulations mishandling after leaving product durin g transit (10 CFR 71*) the site 3.1.1.9, 3.1.1.14 , 3.1.1.23 , 3.1.2.4 , Failure of safe-geometry Accidental critica l ity from S.C.04 , Fai lur e of 3.1.2.7 , 3.1.2.13 , 3.1.2.16 , confinement fissi l e so luti on not confineme nt in safe 3.1.2.17, 3.2.1.6 , 3.2.1.10, confi ned in safe geo m etry geometry; spi ll of 3.2.1.20, 3.2.1.22, 3.2.1.23 , fi ss ile material 3.2.2.9, 3.2.2.1 3 , 3.2.3.6 , 3.2.3.8 , solution 3.2.5.9, 3.2.5.14, 3.2.5.23, 3.8.1.9 , 3.8.1.13, and 3.8.1.22 3.1.1.4, 3.1.1.16, 3.2.5.4, 3.2.5.16 , Tank overflow into proc ess Accidental criticality issue S.C.06, System and 3.8.1.4 ve ntilation system -Fiss ile so lution entering overflow to proc ess a syste m not necessarily ventilation involving designed for fissile fissile material sol ution s 3.1.1.23, 3.2.1.23 , 3.2.5.23, and Uran ium so lu tion is Accide nt a l criticality from S.C.07 , Leak of 3.8.1.22 transferred via a leak fissi l e so luti on not fissile so luti on betwee n the process system confined in safe geometry across auxiliary and the h ea t er/coo l ing sy s tem boundary jackets or coil s on a tank or (chi ll ed water or in an exc han ger steam) 3.2.1.4 , 3.2.1.5, 3.2.2.3, 3.2.2.4, F issile product solution Criticality safety issue -S.C.10, Inad ver tent 3.2.2.5, 3.2.3.6, and 3.2.4.6 transferred to a system not Fissile solution directed to tran sfe r of solution d es igned for safe-geometry a sys tem not intended for to a system not confinement fissile solution d esigne d for fissile so lutions 13-21

.; .. ;. NWMI ...... ..* .. ........ *. . * * ! NORTIIW'En MEOtCAL ISOTOP£S NWMl-2013-021, Rev. 2 Chapter 13.0 -Accident Analysis Table 13-11. Adverse Event Summary for Molybdenum Recovery and Identification of Accident Sequences Needing Further Evaluation (3 pages) Bounding accident PHA item numbers description Consequence Accident sequence 3.1.1.1 3 , 3.1.2.9, 3.2.1.15, 3.2.5.13 , and 3.8.1.12 3.1.1.25, 3.2.5.25, 3.3.1.25, 3.5.1.25, and 3.8.1.24 3.7.1.1 , 3.7.1.2, 3.7.2.1 , 3.7.3.1 , 3.7.3.2, and 3.7.4.l 3.1.1.9, 3.1.1.14, 3.1.1.23, 3.1.2.7, 3.1.2.13, 3.1.2.16, 3.1.2.17, 3.2.1.6, 3.2.1.20, 3.2.1.22, 3.2.1.23, 3.2.2. 7 , 3.2.2.9, 3.2.2.13, 3.2.3.6, 3.2.3.8, 3.2.3.l 0, 3.2.4.10, 3.2.5.9, 3.2.5.14, 3.2.5.23, 3.3.1.9, 3.3.1.14, 3.3.1.18, 3.3.1.22, 3.3.1.23, 3.3.2.4, 3.3.2. 7, 3.3.2.13, 3.3.2.16, 3.3.2.17, 3.4.1.5, 3.4.1.9, 3.4.1.19, 3.4.1.21, 3.4.1.22, 3.4.2.6, 3.4.2.7, 3.4.2.12, 3.4.3.6, 3.4.3.8, 3.4.3. l 0, 3.4.3.14, 3.4.4.6, 3.4.4.10, 3.4.4.14, 3.5.1.9, 3.5.1.14, 3.5.1.16, 3.5.1.23, 3.5.2.4, 3.5.2. 7 , 3.5.2.13, 3.5.2.16, 3.5.2.17, 3.6.1.5, 3.6.1.6, 3.6.1.10, 3.6.1.20 , 3.6.1.20, 3.6.1.23, 3.6.2.7, 3.6.2.9, 3.6.2.13, 3.6.3.8, 3.6.3.10, 3.6.3.14, 3.6.4.10 , 3.6.4.14, 3.8.1.9, 3.8.1.13, and 3.8.1.22 Fai lur e of safe-geometry dimension Hydrogen buildup in tanks or sys tem, leading to explosive concentrations Operator spi ll s Mo product s olution during remote handling operations Spill of product solution in the hot cell area 13-22 Accide ntal criticality from fissile so lution not confined in safe geometry Explosion leading to radiological and criticality concerns Radiological spill of hi g hdose Mo solution Radiological dose from spi ll of product solution from systems S.C.1 9 , Fai lur e of passive d esign feature; component safe-geometry dimension S.F.02, Accumulation of flammable gas in tanks or systems S.R.O 1, Radiological spill of Mo product during remote handling S.R.01, Spill of product solution in hot cell area NWMl-2013-021 , Rev. 2 Chapter 13.0 -Accident Analysis Table 13-11. Adverse Event Summary for Molybdenum Recover y and Identification of Accident Sequences Needing Further Evaluation (3 pages) Bounding accident PHA item numbers description Consequence Accident sequence 3.1.1.9, 3.2.1.10 , 3.2.1.22, 3.2.2.7, 3.2.2.9 , 3.2.3.8, 3.2.3.10, 3.2.4.10 , 3.2.5.9, 3.3.1.9, 3.3.1.18, 3.3.1.22, 3.3.2.7, 3.4.1.10, 3.4.1.22, 3.4.2.7 , 3.4.3.8, 3.5.1.9, 3.5.1.23 , 3.6.1.10 , 3.6.2. 7 , 3.6.3.8, and 3.8.1.9 3.1.1.7 , 3.1.1.22, 3.2.5.7 , 3.2.5.22, 3.3.1.4, 3.3.1.7 , 3.3.1.16 , 3.5.1.4, 3.5.1. 7 , 3.5.1.16 , 3.5.1.22, 3.8.1.7, and 3.8.1.13 3.7.4.3 3.3.1.23, 3.3.2.16, 3.4.1.22, 3.5.1.23, and 3.6.1.23 Spi ll le ading to spray-type release, ca u s ing airborne radioactivity a bo ve DAC limits for exposure Boiling or carryover of stea m or high-concentration water vapor into the primary process offgas ventilation system affecting retention beds with partial or complete los s of cooling s ystem capabilities A Mo product cask is remo ve d from the hot cell boundar y with improper s hield plug installation High-dose radionuclide so lution leak s through an interface between the process system and a heating/cooling jacket coil into a secondary syste m (e.g., chilled water or steam condensate) r e lea s in g radionuclides to workers, the public , and environment Radiological dose from airbo rn e s pray of product so luti on from syste ms Radiological release from retention beds Pote ntial do se to workers, the public , and/or e nvironment du e to r e lea se or mi s handlin g of Mo product durin g transit High-dose radionuclide sol ution that leak s to the environment through another system to expose workers or the public S.R.03 , Spray of product so luti o n in h o t ce ll area S.R.04, Loss of cooli ng, leadin g to liquid or steam carryove r into the primary offgas treatment train S.R.12, Mo product i s re l eased during sh ipment S.R.13, High do se radionuclide containing solution leak s to chilled water or steam condensate system

10 CF R 7 1 , " Packagi n g a nd Transportat i on of R a dio ac tiv e M ate ri a l ," Code of F edera l Regulations , Office of t h e Feder a l Re giste r , as a mended. D AC deriv e d ai r co n ce ntr ation. DOT U.S. Department of Transportat i on. Mo mol ybde num. N I A PHA 13-23 not applicab l e. pro cess hazards a n a l ysis.

.; .. ;. NWMI ...... .. .. .. : ... * * *

NORTHWlST MEDICAL ISOTOPE S NWMl-2013-021 , Rev. 2 Chapter 13.0 -Accident Analysis Table 13-12. Adverse Event Summary for Uranium Recovery and Identification of Accident Sequences Needing Further Evaluation (4 pages) Bounding accident PHA item numbers description Consequence Accident sequence 4.1.1.4 , 4.1.1.18, 4.2.1.4, 4.2.1.6 , 4.2. I. I 7, 4.2. I .18, 4.2.3.6, 4.2.8.4 , 4.2.8.18, 4.2.10.4, 4.3.1.4, 4.3.1.6, 4.3. I .1 8, 4.3. I .19 , 4.3.3.6 , 4.3.8.4 , 4.3.8. I 8, 4.3. I 0.4 , 4.4. I .4, 4.4.1.17, 4.5.1.4 , 4.5.1.17, 4.5.2.4, 4.5.2.17 , 4.5.3.4, and 4.5.3.14 4.1.1.6, 4.2.1. 7, 4.2.2.4 , 4.2.3.4, 4.2.3.7 , 4.2.3.8, 4.2.8.7 , 4.3.1.7 , 4.3.2.4 , 4.3.3.4, 4.3.3.7 , 4.3.3.8, 4.3.8. 7, 4.4.1.6, 4.5.2.6 , and 4.5.3.6 4.1.1.14 , 4.2.1.14 , 4.2.3.16 , 4.2.8.15 , 4.3.1.15, 4.3.3.16 , 4.3.8.15, 4.3.9.20 , 4.4.1.14 , 4.5.1. I 4 , 4.5.2. I 4 , and 4.5.3.11 4.1.1.8, 4.1.1.9, 4.1.1.12, 4.1.1.13, 4.1.1.16 , 4.2.1.9, 4.2.1.13, 4.2.5.11, 4.2.8.10, 4.2.8.13, 4.2.8.14, 4.2.8.17, 4.2.9.18, 4.3.1.10, 4.3.1.11, 4.3.1.14, 4.3.1.17, 4.3.1.18, 4.3.5.11, 4.2.8.10 , 4.3.8.13 , 4.3.8.14 , 4.3.8.17, 4.3.9.18, 4.4.1.8, 4.4.1.9 , 4.4.1.12 , 4.4.1.13 , 4.4.1.16, 4.5.1.16, 4.5.2.8, 4.5.2.9, 4.5.2.12 , 4.5.2.13 , and 4.5.2.16 4.1.1.10 , 4.1.1.15, 4.1.1.23, 4.2.1.11 , 4.2.1.15, 4.2.1.24 , 4.2.2.1 , 4.2.3.11, 4.2.3.13 , 4.2.3.18 , 4.2.3.22, 4.2.3.23 , 4.2.3.24 , 4.2.4.10, 4.2.5.10, 4.2.7.8, 4.2.8.11 , 4.2.8.16, 4.2.8.23, 4.2.9.16 , 4.2.9.29 , 4.2.9.34, 4.3.1.12, 4.3.1.16, 4.3.1.25, 4.3.2.1 , 4.3.3.11, 4.3.3.13 , 4.3.3.18 , 4.3.3.2 2 , 4.3.3.23, 4.3.3.24 , 4.3.4.10, 4.3.5. l 0 , 4.3.7.8, 4.3.8.1 1 , 4.3.8.J 6 , 4.3.8.23, 4.3.9. I 6 , 4.3.9.28, 4.3.9.34, 4.4. I. I 0, 4.4. I .15 , 4.4. I .23 , 4.5.1.23, 4.5.2.10, 4.5.2.15 , 4.5.2.23 , 4.5.3.8, 4.5.3.12, and 4.5.3.19 Tank overflow into Accidental criticality S.C.06 , System overflow process ventilation sys tem i ss u e -Fissi l e sol ution to pro cess ventilation Uranium solution backflows into an auxiliary support system (water line, purge line, chemical addition line) due to various causes Failure of safe geo m e try dimension caused by co nfi g uration man age m e nt (insta ll at ion , maintenan ce) or exte rnal event Uranium precipitate or other high uranium solids accumulate in geometry vessel Failure of safe-geo metry co nfin e ment du e to s pill of uranium s olution from the syste m 13-24 enters a system n ot involvin g fissi l e m a ter ia l n ecessar ily de signe d for fissile solutions Accidental criticality S.C.08 , System backflow issue -Fissile solution into auxiliary support enters a system not system necessarily designed for fissile solutions Accidental criticality from fissile so luti o n not confined in safe geo metry Accidental criticality from fissile solution not confined to safe geometry and interaction controls within allowable concentrations Acc idental criticality from fissile so luti o n n o t co nfined in sa fe geome try S.C.19 , Fa ilur e of p ass i ve d esign feature; component geo m etry dimen sio n S.C.20 , Failure of concentration limits S.C.04 , Failure of co nfin eme nt in safe geo m etry; s pill of fissile materi a l so lution NWMl-2013-021 , Rev. 2 Chapter 13.0 -Accident Analysis Table 13-12. Adverse Event Summary for Uranium Recovery and Identification of Accident Sequences Needing Further Evaluation (4 pages) Bounding accident PHA item numbers description Consequence Accident sequence 4.2.3.21, 4.2.4.11 , 4.2.6.12, Failure of safe-geometry Accidental criticality S.C.14, Failure of 4.3.3.21, 4.3.4.11 , and 4.3.6.12 confinement due to from fissile solu tion confinement in safe inadvertent transfer to not confined in safe geometry; transfer of U-bearing re sin to the U geo metry U-bearing resin to U IX IX waste collection tanks waste collection tanks through a broken retention eleme nt 4.2.5.5 , 4.3.1.9, 4.3.5.5, and Fai lur e of safe-geo m etry Accide nt a l critica lity S.C.14, Failure of 4.5.1.5 co nfin ement due to from fissile solution confinement in safe inadvertent transfer to not confined in safe geometry; transfer of U-bear in g solution to the geometry U-b earing so lution to U IX waste collect ion U IX waste co ll ection tanks tank s 4.2.7.7, 4.3.7.7, and 4.5.3.10 Inadvertent transfer of high Accidental criticality S.C.15, Too high of uranium-concentration too high of uranium uranium mass in spent so lu tion or resins to spent mass in waste stream re sin was te strea m resin tanks 4.2.9.10 , 4.2.9.19 , 4.2.9.21 , Uranium is inadvertently Acc id e nt a l criticality S.C.09, Carryover of 4.2.9.23, 4.2.10.10 , 4.2.10.1 2 , carr i e d over from th e from fissile so l ution uranium to the condenser 4.3.9.10 , 4.3.9.19 , 4.3.9.2 1 , concentrator (I or 2) t o the not co nfin ed in safe or condensate tanks 4.3.9.23, 4.3.10.10 , and 4.3.10.12 condenser and geometry subseq u ent l y, the condenser conde n sate collection tanks 4.2.9.36 and 4.3.9.36 Uranium so lution is Accidenta l criticality S.C.07, Uranium-transferred via a leak from fissile so lution containing so lution leak s between the process not confined in safe to chilled water or steam system and heater/cooling geometry condensate syste m jackets or coils on a tank or in an exchanger 4.1.1.8 , 4.1.1.22 , 4.2.1.9 , 4.2.1.17, Carryover of high-vapor High airborne S.R.04 , Carryover of 4.2.1.23 , 4.2.9.11 , 4.2.9.14, content gases or entrance radio nuclid e release , h eavy vapor or solution 4.2.9.17, 4.2.9.23, 4.2.9.30, of so lution s into the affecti n g workers and into the process 4.2.9.32, 4.2.10.14, 4.3.1.10, process ventilation h eader the public ventilation h ea d e r causes 4.3.1.18, 4.3.1.24 , 4.3.9.11 , can ca u se poor down s tream fai lur e of 4.3.9.14 , 4.3.9.17 , 4.3.9.23, performance of the r etention bed, releasing 4.3.9.30 , 4.3.9.32 , 4.3. l 0.14 , retention bed materials radionuclides 4.4.1.8, 4.4.1.22 , 4.5.1.9 , 4.5.1.22 , and release rad i onuclides and 4.5.2.8 13-25 NWM I ...... *

NOmtMST MEDtcAL lSOTOPl S NWMl-2013-021 , Rev. 2 Chapter 13.0 -Accident Analysis Table 13-12. Adverse Event Summary for Uranium Recovery and Identification of Accident Sequences Needing Further Evaluation (4 pages) Bounding accident PHA item numbers description Consequence Accident sequence 4.1.1.10, 4.1.1.15, 4.1.1.23, 4.2.1.11, 4.2.1.15, 4.2.1.24, 4.2.2.1, 4.2.2.4, 4.2.3.11, 4.2.3.13, 4.2.3.18, 4.2.3.22, 4.2.3.23, 4.2.3.24, 4.2.4.10, 4.2.5.10, 4.2.6.11, 4.2.7.8, 4.2.8.11, 4.2.8.16, 4.2.8.23, 4.2.9.16, 4.2.9.28, 4.2.9.34, 4.3.1.12, 4.3.1.16, 4.3.1.25, 4.3.2.1, 4.3.2.4, 4.3.3.11, 4.3.3.13 , 4.3.3.18, . 4.3.3.22, 4.3.3.23, 4.3.3.24, 4.3.4.10, 4.3.5.10, 4.3.6.11, 4.3.7.8, 4.3.8.11, 4.3.8.16, 4.3.8.23, 4.3.9.16, 4.3.9.28, 4.3.9.34, 4.4.1.10, 4.4.1.15, 4.4.1.23 , 4.5.1.11, 4.5.1.15, 4.5.1.23 , 4.5.2.10, 4.5.2.15, 4.5.2.23, 4.5.3.8, 4.5.3.12, and 4.5.3.19 4.2.1.12, 4.2. 1.24, 4.2.2.1, 4.2.3.11 , 4.2.3.13, 4.2.3.18, 4.2.3.22, 4.2.3.23, 4.2.4.10, 4.2.5. I 0, 4.2.6.11, 4.2.8.11, 4.2.8.1 6 , 4.2.8.23, 4.2.9.16, 4.2.9.28, 4.2.9.34, 4.2.9.35, 4.3.1.12, 4.3.1.16, 4.3.1.12 , 4.3.1.25 , 4.3.2.1 , 4.3.3.11 , 4.3.3.13, 4.3.3.18 , 4.3.3.22 , 4.3.3.23 , 4.3.4. I 0 , 4.3.5.10 , 4.3.6.11 , 4.3.8.11 , 4.3.8.16 , 4.3.8.23, 4.3.9.16, 4.3.9.28 , 4.3.9.34, 4.3.9.35, 4.4.1. I 0, 4.4.1.15, 4.4.1.23, 4.5.1.11 , 4.5.1.23, 4.5.2.10 , 4.5.2.15, 4.5.2.23, and 4.5.3.19 4.2.9.37, 4.2.9.36, 4.3.9.36, and 4.3.9.37 High-dose radionuclide solution is spilled from the system High-do se radionuclide so lution is sprayed from the system, ca u sing high airbo rn e radioactivity High-dose radionuclide solution leaks through an interface between the process system and a heating/cooling jacket coil into a secondary system (e.g., chilled water or steam condensate), releasing radionuclides to workers, the public , and environment 13-26 Radiological relea se of high-dose solution with potential to impact workers, the public, or environment Radiological release of high-dose spray that r emains s u spe nd ed in the air , givi n g high dose to workers or the public High-dose radionuclide solution that leaks to the environment through another sys tem to expose workers or the public S.R.01, Spill of product solution in hot cell area S.R.03 , Spray of product so luti on in hot ce ll area S.R.13, High-dose, radionuclide-containing solution leaks to chilled water or steam condensate system NWMI *:*:**:*:* ...... ' *

NOATitWUT MEDtCAL ISOTOPE S NWM 1-2013-021 , Rev. 2 Chapter 13.0 -Accident Analysis Table 13-12. Adverse Event S ummar y for Uranium Recovery and Identification of Accide nt Sequences Needing Further Evaluation (4 pages) Bounding accident PHA item numbers description Consequence Accident sequence 4.1.1.25, 4.2.1.2 6 , 4.2.8.25, 4.3.1.27, 4.3.8.25 , 4.4.1.25, 4.5.1.25, 4.5.2.2 5, and 4.5.3.2 1 4.1.1.24 , 4.2.1.25, 4.2.8.24, 4.2.10.18 , 4.3.1.26 , 4.3.8.24, 4.3.10.18 , 4.4.1.24 , 4.5.1.24, 4.5.2.24, and 4.5.3.20 4.2.4.8 and 4.3.4.8 4.2.10.6 and 4.3.10.6 4.2.10.8, 4.2.10.11 , 4.2.10.1 7, 4.3.10.8, 4.3.10.11, and 4.3.10.17 Hydrogen buildup in tanks Ex pl os ion l eading to or sys tem , l ea din g to radio logic a l and ex plo s i ve concentrations critica lity concerns Higher do se than normal due to double-batching an activity or due to buildup of radionuclides in the system over time Radiation dose is elevated over normal operational levels, but does not exceed low consequence values for exposure to workers due to s hielding H i g h temperature Co n seq u e n ce is n ot pre-elution or regeneration fu ll y und e r stood reagent causes unknown im p ac t on I X res in Same as S.C.08 except Low consequence with low-dose solution resulting in from condenser condensate contaminated system Spi ll or spray of l ow-do se con d e n sate Low co n se qu ence res ultin g in conta minated surfaces and dose to worker below int ermediate co n se quen ce d ose l eve l s S.F.02, Accumulation of flammable gas in tanks or syste m s Hot cell shielding is credited as the normal condition, mitigating safety feature for this hazard (adverse condition does not represent failure of the safe ty function of th e IROFS) Tentatively S.R.14 N I A N I A IR OFS T X N I A it ems relied o n for safety. ion exchange.

PH A u p rocess hazards a n a l ysis. = u ra nium. n ot applicab l e. Uranium Recover y Open Item The following adverse eve nt needs to be further researched. PH A items 4.2.4.8 and 4.3.4.8 po stu l a te high-t emperat ure 2 molar (M) nitri c ac id (HN03) solution b e in g u se d on the uranium purification ion-e xc h a n ge (IX) media as a pre-elution rinse. The consequence of the bo u nding accident was not full y understood and n ee d s to b e further researched.

The lik e lihood was id e ntified as low , as there are no goo d causes of the high temperature from the supply tank other than an im p rop e r m1xmg seq u e nce. This upset would not ca u se extremely elevated temperatures nor go undet ecte d. 13-27

.; .. NWMI ...... .. .. .......... ' *

Nomtw£ST MEDICAL ISOTOPES NWMl-2013-021, Rev. 2 Chapter 13.0 -Accident Analysis Table 13-13. Adverse Event Summary for Waste Handling and Identification of Accident Sequences Needing Further Evaluation (2 pages) PHA item Bounding accident numbers description Consequence Accident sequence 5.1.1.13 High uranium content Solution from this tank is solidified S.C. l 0, Fissile solution in product solution is in a non-favorable geometry process high-dose waste collection directed to the high-dose with potential to result in accident tanks (a non-fissile solution waste collection tanks by nuclear criticality at the high boundary) accident uranium concentration 5.2.1.13 and High uranium content Solution from this tank is solidified S.C.10, Fissile solution is 5.2.2.13 product solution enters the in a non-favorable geometry process directed to the low-dose low-dose waste collection with potential to result in accidental waste collection tank tanks by accident nuclear criticality at the high uranium concentration 5.4.1.1 High uranium content The mass of uranium may exceed a S.C.22, High concentration accumulates in the TCE safe mass and result in an accidental of uranium in the TCE reclamation evaporator nuclear criticality without evaporator residue monitoring and controls 5.4.2.l Dissolved uranium The mass of uranium may exceed a S.C.23, High concentration products may accumulate safe mass and result in an accidental in the spent silicone oil in the silicone oil waste nuclear criticality without waste stream monitoring and controls 5.1.1.24 and Hydrogen buildup in Explosion leads to radiological and S.F.02, Accumulation of 5.1.4.23 tanks or system leads to criticality concern flammable gas in tanks or explosive concentrations systems 5.1.1.4, 5.1.1.16, Several tank or Radiological release may cause a S.R.04, High-dose solution 5.1.4.4, 5.1.4.15, components vented to the high-dose exposure to workers and from a tank or component and 5.1.4.17 process vessel ventilation the public overflows into the process system overflow and send ventilation system, high-dose solution into compromising the retention process ventilation system beds components that exit the hot cell boundary 5.1.1.6 and 5.1.4.6 The purge air system (an Radiological release may cause a S.R.16, High-dose solution auxiliary system that high-dose exposure to workers and backflows into the purge air originates outside the hot the public system cell boundary) allows high-dose radionuclides to exit the boundary in an uncontrolled manner 5.1.1.10, 5.1.1.14, Spills from multiple Radiological release may cause a S.R.01, High-dose solution 5.1.1.22, 5.1.2.26, sources; materials high-dose exposure to workers and spill in the hot cell waste 5.1.2.31, 5.1.4.10, originating from high-the public handling area 5.1.4.13, 5.1.4.21, dose process solutions are 5.1.5.16, 5.1.5.19, spilled from the system or 5.1.5.20, 5.3.1.14, process that normally 5.3.1.17, and confines them 5.3.1.18 13-28 NWMl-2013-021 , Rev. 2 Chapter 13.0 -Accident Analysis Table 13-13. Adverse Event Summary for Waste Handling and Identification of Accident Sequences Needing Further Evaluation (2 pages) PHA item Bounding accident numbers description Consequence Accident sequence 5.1.1.2 1 , 5.1.2.28 , Severa l tank s or R a diolo g ical release may cause a and 5.1.4.2 0 co mpon e nt s vente d to th e high-do se ex p osure to workers and process vessel venti lation the publi c 5.1.1.22 , 5.1.2.26 , 5.1.2.31 , 5.1.2.32 , 5.1.4.10 , and 5.1.4.21 5.1.2.9, 5.1.2.18, 5.1.2. I 9, and 5.1.2.21 5.1.2.3 3 5.1.5.8 5.5.1.1 sys tem evo l ve high liquid va p o r co n ce ntr ations, resulting in accelerated high-do se radionuclide release to th e stac k from wetted ret e nti on beds Catastrophic failure of a component (high pressure or detonation) leads to rapid release of solution and higher airborne levels A d ve r se events in the co ncentrator or eva porator syste m s lead to carryove r of high-dose so lution into the condenser, r es ulting in high-dose radionuclides in the low-do se waste co llection tanks Normally low-do se vapor in the condenser leaks through the boundary into the chilled water sys tem High-dose so lution i s inadvertently mi sfe d into th e so lidification hopper Due to several potential initiators, the payload container or the shipping cask of high-dose encapsulated waste is dropped during transfer from the storage location to the conveyance Radiological release may cause a high-dose exposure to workers and the public R ad i o l og ical ex p os ur e l eve l s on the l ow-do se encapsulated waste may excee d interm e di ate or high conse qu e n ce l eve l s Radiological release may cause a high-dose exposure to workers and the public R ad iolo g ic al r e l ease may ca u se a hi g h-d ose ex p osure to workers an d th e publi c Radiological issue -Depending on damage from the drop, worker s could receive high-dose radiation exposure.

Unshielded package may impact dose rates at the controlled area boundary. S.R.04, High-dose radi o nuclide r e lea se du e to hi gh vapor co nt e nt in exhaust S.R.03, High-dose solution spray events from equipment upsets may cause high airborne radioactivity S.R.17, Carryove r d ose so luti o n into condensate (a low-d ose waste stream) S.R.13 , Process vapor from the evaporator leaks acros s the condenser cooling coil s into the chilled water sys tem S.R.18 , High-dose so luti on flow s in to the so lidifi ca ti o n hopp e r S.R.32, Container or cask dropped during transfer P H A proce ss hazards analysis.

TCE trich 1 oroet h y l e ne. 13-29 NWMl-2013-021, Rev. 2 Chapter 13.0 -Accident Analysis Table 13-14. Adverse Event Summary for Target Receipt and Identification of Accident Sequences Needing Further Evaluation (2 pages) PHA item numbers Bounding accident description Consequence Accident sequence 6.1.2.4, 6.1.2.8, 6.1.2.9, Handling damage to th e target Accidenta l nuclear critica li ty S.C.21, Target basket 6.1.2.11 , 6.1.2.14 , and basket fixed-interaction passive l eads to high dose to workers passive design control 6.1.2.15 design feature leads to accidenta l and potential dose to the fa ilur e on fixed nuclear criticality public interaction spacing 6.1.2. 7, 6.1.2.l 0, Too much uranium mass is Accidental nuclear criticality S.C.02, Operator 6.2.1.1, 6.2.1.5, 6.2.2.1, handled at once either through leads to high dose to workers exceeds batch handlin g 6.2.2.2, 6.2.2.4, 6.2.2.5, operator error or inattention to and potential dose to the limits during target 6.2.3.3, 6.2.4.1, 6.2.4.2, housekeeping public disassembly operations 6.2.4.4, 6.2.6.1, 6.2.6.3, in the hot cell and 6.2.6.4 6.2.1.6 , 6.2.2.9, 6.2.3.4 , Operator accum ul ates more Accidental nuclear criticality S.C.03, Fai lur e of and 6.2.6.6 targets or [Proprietary leads to high dose to workers administrative control Information]

containers into and potential dose to the on interaction limit spec ifi c room than allowed and public during handling of vio l ates interaction control targets and irradiated

[Proprietary Information]

6.2.1.3, 6.2.1.4, 6.2.1.5, Too much uranium in the solid Accidental nuclear criticality S.C.17, [Proprietary 6.2.2.2, 6.2.2.4, 6.2.2.6, waste container (that is not safe-leads to high dose to workers Information]

residual 6.2.3.1, 6.2.3.2, 6.2.3.3 , geometry) entering the solid and potential dose to the determination fails, and 6.2.5.1, 6.2.5.3, 6.2.5.4, waste encapsulation process public used target housings 6.2.5.8, 6.2.6.1, 6.2.6.2, (where moderator will be added have too much uranium 6.2.6.3, and 6.2.6.5 in the form of water) in solid waste encapsulation waste stream 6.1.1.5, and 6.1.1.9 Cask involved in an in-transit High dos e to workers during S.R.28 , High dose to accident or improperly closed receipt inspection and workers during prior to shipment, leading to opening activities s hipm ent receipt streaming radiation inspection and ca s k preparation activities due to damaged irradiated target cask 6.1.1.10 Cask involved in in-tran sit High dose to workers during S.R.29, High dose to accident or targets failed durin g rec e ipt inspection and workers from re l ease of irradiation, leading to excessive opening activities gaseous radionuclides offgassing from damaged targets during cask receipt inspection and preparation for target basket removal 6.1.1.11, 6.1.1.12 , Sea l b etween cask and h ot ce ll High dose to workers from S.R.30, Cask docking 6.1.2.1 , 6.1.2.13 , and docking port fails from a number streaming radiation and/or port fai lure s lead to 6.1.2.16 of causes high airborne radioactivity high dose to worker s due to streaming radiation and/or high airborne radioactivity 13-3 0 NWMl-2013-021 , Rev. 2 Chapter 13.0 -Accident Analysis Table 13-14. Adverse Event Summary for Target Receipt and Identification of Accident Sequences Needing Further Evaluation (2 pages) PHA item numbers Bounding accident description Consequence Accident sequence 6.1.1.1 Cask involved in a crane High dose to workers during S.R.32, High dose to movement incident, leading to receipt inspection and workers during streaming radiation opening activities shipment receipt inspection and cask preparation activities due to damaged cask in crane movement incident 6.1.2.3 and 6.1.2.5 Improp er handling activities High external d ose to S.R.19, High target result in high external dose rates workers basket retrieval dose through the hot cell wa ll when rate removing the target basket and setting it in the target basket carousel shie ld ed well 6.1.2.10, 6.1.2.15, [Proprietary Information]

spilled High dose to workers or the S.R.20, Radiological 6.2. l .5, 6.2.2.2, 6.2.2.4, or ejected in an uncontrolled public may result from spill of irradiated 6.2.3.3, 6.2.4.2, 6.2.5.4, manner during various target and uncontrolled accumulation of targets in the hot cell 6.2.6. I, and 6.2.6.3 container-handling activities or irradiated

[Proprietary area during target-cutting activities Information]

6.1.2.15 Operations removing the target High dose to workers due to S.R.21, Damage to the basket (potentially in a h eavy degraded s hi e ldin g hot cell wall providing sh i e ldin g housing) with a hoi s t shielding leads to striking the wall and damaging the hot cell wall s hi e ldin g function 6.2.4.5 Delays in processing a batch of High dose to workers from S.R.22, Decay heat removed [Proprietary high airborne radioactivity buildup in unproce sse d Information]

results in long-term

[Proprietary heating outside of target housing Information]

remo ved from targets lead s to higher high dose radionuclide offgassing 6.2.4.6 and 6.2.4.7 Improper venting of the chamber High dose to workers from S.R.23, Offgassing or prem ature open in g of the hi gh airborne radioactivity from irradiated target valve during processing of a di sso lution tank occurs previously ad d ed b atc h results in when the upp er valve is release of high-dose opened radionuclides to the hot cell space 6.2.5.5, 6.2.5.6, and The seal on the bagless transport High dose to workers from S.R.24, Bagles s 6.2.5.7 door fails and leads to high dose high airborne radioactivity transport door failure radionuclides escaping the hot cell containment or confinement boundary PHA proce ss ha za rds ana l ys is. 13-31

.; .. NWMI ::.**.*.* .......... ' *

NomtW£ST MEDICAL ISOTOf'fS NWMl-2013-021, Rev. 2 Chapter 13.0 -Accident Analysis Table 13-15. Adverse Event Summary for Ventilation System and Identification of Accident Sequences Needing Further Evaluation PHA item numbers 7.1.1.7 and 7.1.1.8 7.1.1.2, 7.1.1.3, and 7.1.1.6 7.1.1.1 0 and 7.2.1.19 7.1.1.11 and 7.2.1.20 7.1.1.12, 7.1.1.14 , and 7.2.1.21 7.2.1.4 , 7.2.1.7 , 7.2.1.8, 7.2.1.9, 7.2.1.13, 7.2.1.14, 7.2.1.17 , and 7.2.1.22 7.2.1.12 and 7.2.1.1 7 Bounding accident description T oo much uranium acc umulated on the HEP A filter a llows an accidental criticality when left in th e wrong configuration Hydrogen buildup in the ventilation system, due to insufficient flow to sweep it away, leads to fire in the HEPA filters or carbon beds Ignition so urc e causes fire in th e carbon bed Overloading of HEPA filter lead s to failure and release of accumulated radionuclide particulate The acc umulat ed high-do se (and l ow-do se) radionuclides retained in the car bon b ed are r e l ease d thr o u g h a flow , heat , or chemical reaction from the media (or the media is r e l eased) Loss of the negative air balance between zones (a confinement feature that prevents migration of radionuclides from areas of high do se and high concentration to areas oflow concentration)

During an exte nd e d power outage, so me so luti o n sys t e m s freeze and cause failure of th e pipin g sys t e m , l eadi ng to radiological s pill s H EPA high-efficie n cy particulate air. Consequence Accidental nucl ear c riti ca li ty l ea ds to high dose to workers and pot e nti al dose to th e publi c A detonation or deflagration event in the ventilation system rapidly releases retained high-dose radionuclides , causing high airborne radioactivity Fire eve nt in the ve ntil ation syste m rapidly releases retained high-d ose ra dionuclid es , ca u s in g high air borne radioactivity High dose to workers from high airborne radioactivity High do se to workers fro m high airborne ra dioactivi ty High dose to workers from high airborne radioactivity High dose to worke r s from high airborne ra di oactivity Accident sequence S.C.2 4 , High uranium co nt e nt on HEPA filters S.F.06, Accumulation of flammable gas in ventilation system components S.F.0 5, F ire in the carbon b e d S.R.25, HEPA filter failure S.R.04 , Car bon bed ra di o nu c lid e retention failure S.R.26, Failed negative air balance from zone to zone or failure to exhaust a radionuclide buildup in an area S.R.27, Exten ded o uta ge of heat , le ading t o freezing, pip e failure , and release of radionuclides from liquid process sys t ems PHA process h azards ana l ysis. 13-32 PHA item NWMl-2013-021 , Rev. 2 Chapter 13.0 -Acci d e n t Analysis Table 13-16. Adverse Event Summary for Node 8.0 and Identification of Accide n t Sequences Needing Furt h er Evaluation (5 pages) numbers Bounding accident description Consequence Accident sequence 8.2.1.5 Large leak l eads to localized l ow Standar d industrial hazard -Localized Nitrogen storage or oxygen levels that adverse l y asphyxiant distribution system leak impact worker performance and may lead to death 8.5.1.1 and Operator doub l e-batches allotted Accidental criticality issue -Too much S.C.02, Failure of AC 8.5.1.5 amount of material (fresh U, scrap fissile mass in one location may become on mass (batch limit) U, [Proprietary Information], critical during handling of target batch) into one location or fresh U, scrap U, container during handling [Proprietary Information], and targets 8.5.1.3 and Operator handling various Accidenta l criticality issue -Too much S.C.03, Fai lur e of AC 8.5.1.5 containers of uranium or batches uranium ma s s in one location on interaction limit of uranium component s brings during handling of two containers or batches closer fresh U , scrap U, together than the approved [Proprietary interaction contro l distance Information], and targets 8.6.1.7 A liquid spi ll ofrecycle uranium Critica l ity issue -Fissi l e solution may S.C.04, A l iquid spill or target di sso lution solution collect in unsafe geometry of fissile so l ution occurs within the hot cell occurs boundary 8.6.1.9 Process solutions backflow Critica li ty issue -Fissi le solution may S.C.08 , Fissi l e process through chemical addition lin es to collect in unsafe geometry solutions backflow l ocations outside the hot cell through chemical boundary addition line s 8.6.1.13 Improper instalJation of HEPA Accidental nuclear criticality l eads to S.C.24, High uranium fi l ter s (and prefilters) leads to high dose to worker and potential dose content on HEPA transfer of fissile uranium to public filters particulate into downstream sections of the ventilation system with uncontrolled geometries 8.5.1.2 and Operator handling enric h ed Critica lit y h azard -Too much uranium S.C.27 , Fai lur e of AC 8.5.1.5 solutions pours so luti on into an mass in one place can l ead to accidenta l on volume limit during unapproved container nuclear cr iti cality sampling 8.4.1.8 and Drop of a hot cell cover block or Criticality is s ue -Structural damage S.C.28, Crane drop 8.6.1.12 other heavy object damages SSCs could adversely damage SSCs relied on accident over hot cell relied on for safety for safety, leading to accidents with or other area with SSCs intermediate or high consequence relied on for safety 13-33 PHA item NWMl-2013-021, Rev. 2 Chapter 13.0 -Accident Analysis Table 13-16. Adverse Event Summary for Node 8.0 and Identification of Accident Sequences Needing Further Evaluation (5 pages) numbers Bounding accident description Consequence Accident sequence 8.1.2.7 and A general facility fire (caused by Uncontrolled fire can lead to damage to S.F.08, General facility 8.1.2.12 vehicle accident inside or outside SSCs relied on for safety, resultin g in fire of the facility, wildfire, chemical, radiological, or criticality combustible fire in non-industrial hazard s that represent intermediate to areas, or fire in non-licensed high consequence to workers, the material processing areas) spreads public , and environment to areas in the building that contain licensed material 8.2.1.7 Leak of hydrogen in the facility May lead to an explosion (detonation or S.F.09, Hydrogen attains an explosive mixture and deflagration), depending on the location explosion in the facility finds an ignition source, leading in the facility where the hydrogen leaks due to a leak from the to detonation or deflagration of from. Explosion may compromise hydrogen storage or the mixture SSCs to various degrees and may lead distribution system to intermediate or high consequence events. 8.6.1.11 Electrical fire sparks larger Radiological and criticality issue -S.F.10, Combustible combustible fire in one of the hot Depending on the location and quantity fire occurs in hot cell cells of combustibles or flammables left in area the area , a fire in the hot cell area could rupture sys tem s with high-dose fission products and/or high uranium content, leading to spills and airborne releases 8.1.2.9 and A natural gas leak develops in the Potential explosion that could S.F.11, Detonation or 8.4.1.9 steam generator room and finds catastrophically damage nearby SSCs. deflagration of natural an ignition source, resulting in a Depending on the extent of the damage gas leak in steam detonation or deflagration that to SSCs , an accidental nuclear criticality generator room damages SSCs or an intermediate or high consequence exposure to workers could occur. 8.1.2. 7, Vehicle inside building strikes Accidental nuclear critica l ity l eads to S.M.01, Vehicle strikes 8.3.1.2, and fresh uranium dissolution system high dose to workers and potential dose SSC relied on for 8.6.1.5 component, leading to a spill or to public safety and causes accidental criticality due to damage or leads to an disruption of geometry and/or accident sequence of interaction intermediate or high consequence 8.4.1.6 TBD (impact must be evaluated TBD (impact must be evaluated after S.M.02, Facility after determining all IROFS that determining alJ IROFS that rely on evacuation impacts on rely on personne l action) personnel action) operation 13-34 PHA item NWMl-2013-021 , Rev. 2 Chapter 13.0 -Accident Analysis Table 13-16. Adverse Event Summary for Node 8.0 and Identification of Accident Sequences Needing Further Evaluation (5 pages) numbers Bounding accident description Consequence Accident sequence 8.1.2.13 Flooding from external eve nt s and Crit icality issue -Water accumulation S.M.03, Flooding internal events compromises the under safe geometry storage vesse l s or occurs in building due safe geometry s lab area under in safe interaction storage arrays , to internal system leak certain tanks. Depending on the causing inter spersed moderation. or fire s uppr ession liquid level, interspersed Flooding could compromise safe-system activation moderation of components may geome tr y storage capacity for (likely) be impacted. Floor storage arrays subsequent spills of fissile solution.

are subject to stored containers E ither event could compromise floating (loss of interaction control).

critica lity safety. 8.1.1.1 Large tornado strikes the facility Radiological, chemical, and criticality S.N.01, Tornado issue -Structural damage could impact on facility and adversely damage SSCs relied on for SSCs safety. Facility could lose all electrical distribution.

Facility could lose chilled water system :function (cooling tower outside of buildin g). 8.1.1.2 Straight-line winds strike the Radiological , chemical, and criticality S.N.02, High straight-facility issue -Structural damage could line wind impact on adverse l y dama ge SSCs relied on for facility and SSCs safe ty. Facility could lo se all electrica l distribution.

Facility could lo se chi ll ed water system function (cooling tower outside of building).

8.1.1.3 A 48-hr probab l e maximum Radiological, chemica l , and criticality S.N.03, Heavy rain precipitation event strikes the issue -Structural damage from roof impact on facility and facility collapse could adversely damage SSCs SSCs relied on for safety 8.1.1.4 Flooding occurs in the area in Radiological i ssue -Minor structura l S.N.04, F l oodi n g excess of 500-year return damage i s not anticipated to impact impact on facility and frequency SSCs relied on for safety except that the SS Cs facility could lo se all electrical distribution and/or chilled water system function (cooling tower outside of building) 8.1.1.6 Safe shutdown earthquake strikes -Radiological, chemical, and criticality S.N.05, Seismic impact Seismic shaking can lead to issue -Structural damage could on facility and SSCs damage of the facility and partial adversely damage SSCs relied on for to complete collapse.

This safety. Facility could l ose all electrical damage impacts SSCs inside and distribution.

Facility cou ld lose chilled outside the hot cell boundary.

water system :function (cooling tower Leaks of fissile solution, outside of building). compromise of safe-geometry, and safe interaction storage in solid material storage arrays and pencil tanks or vessels containing enriched uranium solutions.

13-35 NWMI

NOtmfWUTM£0tcAllSOTOP£S NWMl-2013-021, Rev. 2 Chapter 13.0 -Accident Analysis Table 13-16. Adverse Event Summary for Node 8.0 and Identification of Accident Sequences Needing Further Evaluation (5 pages) PHA item numbers Bounding accident description Consequence Accident sequence 8.1.1.9, Hea vy s n owfa ll or ic e buildup R a diolo g ical , ch e mical , and c riti ca li ty S.N.06 , H e a vy 8.1.1.10 ex c ee d s d es i gn lo a din g of the i ss u e -S tructur a l dama g e from roof s n owfa ll or i ce buildup roo f, r es ultin g in co ll a p se of the c oll a p se could ad ve r s el y dam age S S Cs o n faci li ty and SSCs roof a nd dam age t o S SCs (e.g., r e li e d o n fo r s af e t y. Lo ss of s it e tho se out s ide o f th e h o t ce ll s) el e ctri ca l pow er i s highl y lik e l y in h eavy i ce s torm eve nt. 8.6.1.8 Any s tored high-dose product Radiological issue -High-dose solution S.R.01 , A liquid spill solution spills within the hot c e ll is unconfined or uncontrolled and can of high-dose fis s ion boundary cause e x posures to workers , the public , product solution occur s and environment 8.5.1.5 Op e r a t o r s pill s dilut e d s ampl e R a di o l og ic al i ss u e -Pot e ntial s p ray o r S.R.01 , S pill of produ c t o ut s id e of th e hot ce ll area va p o ri za tion o f radi o nuclide co n ta inin g s oluti o n i n laborat o r y v ap o r-ca u s in g ad ve r se w o rk e r ex p os ur e (b ase d o n typi cal l ow quant i ti es handl e d in th e l a borat o r y, thi s i s po s tul a t ed t o b e an int e rm e di a t e co n se qu e nc e eve nt) 8.6.1.10 Recycle uranium transferred out Radiological issue -High radiation may S.R.05 , High-dose before lag storage decay complete occur in non-hot ce ll area s, impacting solution exits hot cell or with significant high-dose workers with higher than normal shielding boundary radionuclide contaminants external doses (destined for UN blending and storage tank) 8.6.1.9 Proc ess so lution s b ac kfl ow R a diolo g ical i ss u e -Hi g h radiati on m a y S.R.16 , Hi g h-do se throu g h c h e mical additi o n l in es to occ ur in non-h o t ce ll ar eas, imp ac tin g pro cess so lution s locati o ns o ut s id e th e h o t ce ll wor k e r s with h igher than norm a l b ac kfl ow through boundar y exte rn a l do ses ch e mi ca l addition lin es 8.6.1.2 and An improperly sealed cover block Radiological issue -Depending on S.R.21 , Damage to the 8.6.1.3 or transport door (e.g., for cask location of damage, s ome streaming of hot cell wall transfers) offer large opening high radiation may occur, impacting penetration , potential s for radiation streaming workers with higher than normal compromising external doses shielding 8.6.1.1 The sea l o n th e ba g l ess tran s p o rt R a di o l og ic al i ss u e -D eg rad e d o r lo ss of S.R.2 4 , Ba g l ess d o or fa il s and le a d s t o high-d os e casca d i n g n ega ti ve a ir pr ess ur e b e tw ee n tran s p o rt d oo r failur e radionuclid es esc a pin g th e hot zo n es may a ll o w high radiol og i ca l ce ll c o n fi n e m e nt boundar y airb o rn e contamin a ti o n to r e l ease wi th o ut proper filtr a tion and a d so rpti o n , l ea din g to high e r than allow e d ex p os ure rat es t o w ork e r s and th e publi c 8.6.1.13 Following process upsets and Radiological and criticality is s ue -S.R.25 , HEPA filter over long periods of operation , Following process upsets and over long failure contamination levels in periods of operation, contamination downstream components leads to levels in downstream components can high dose during maintenance and lead to high dose during maintenance to uncontrolled accumulation of and to uncontrolled accumulation of fissile material fissile material 13-3 6 NWM I ...... *

NOfliTHWHT MEDICAL ISOTOPfS NWMl-2013-021, Rev. 2 Chapter 1 3.0 -Accident Analysis Table 13-16. Adverse Event Summary for Node 8.0 and Identification of Accident Sequences Needing Further Evaluation (5 pages) PHA item numbers Bounding accident description Consequence Accident sequence 8.6.1.2, An improperly sealed cover b l ock Radiological issue -Degraded or loss of S.R.26 , Failed negative 8.6.1.3, and or transport door (e.g., for cask cascading negative air pressure between air balance from zone 8.6.1.6 transfers) compromises negative z ones may allow high radiological to zone or failure to air pressure balance airborne contamination to release exhaust a radionuclide without proper filtration and adsorption, buildup in an area leading to higher than allowed exposure rates to worker s and the public 8.5.l.7 and Laboratory technician is burned Radiological issue -Burns may lead to S.R.31, Chemical bums 8.5.1.8 by solutions containing intermediate consequence events if eyes from contaminated radiological isotopes during are involved solutions during sample sample analysis activities analysis 8.4.1.8 , Drop of a hot cell cover block or Radiological and criticality issue -S.R.32 , Crane drop 8.6.1.4 , and other heavy object damages SSCs Structural damage could adversely accident over hot cell 8.6.1.12 relied on for safety damage SSCs relied on for safety , or other area with SSCs leading to accidents with intermediate relied on for safety or high consequence 8.2.1.1 All nitric acid from a nitric acid Standard industrial accident with S.CS.01, Nitric acid storage tank is released in I hr potential to impact SSCs or cause fume release from the chemical preparation and additional accidents of concern storage room AC admini s trative control. SSC structure s , s y s tems , and components. H E PA high efficiency particul a te a ir. TBD to be d e termined.

IROFS items relied on for safet y. u uranium. PHA proce s s ha za rds anal ys i s. UN uran y l nitrate. The identified accident sequences are further evaluated in QRAs to continue the accident analysis and to identify IROFS for those accident sequences that exceed the performance criteria as specified in NWMI-2014-051, Int e grated Safety Anal y sis Plan for th e Radioisotope Production Fa c ility. 13-37 NWMl-2013-021, Rev. 2 Chapter 13.0 -Accident Analysis 13.2 ANALYSIS OF ACCIDENTS WITH RADIOLOGICAL AND CRITICALITY SAFETY CONSEQUENCES This section presents an analysis of accident sequences with radiological and criticality safety consequences.

In Section 13. 1 .3, a number of th e hazards and accident sequences identified in the PHA that require further evaluation are grouped and identified.

These accident sequences were evaluated using both qualitative and quantitative techniques.

Accidents for operations with SNM (including irradiated target processing , target material rec yc le , waste handling , and target fabrication), radiochemical , and hazardous chemicals were analyzed.

Initiating events for the analyzed sequences include operator error , loss of power , external events, and critical equipment malfunctions or failures.

Shielded and unshielded criticality accidents are assumed to have high consequences to the worker if not prevented. Most of the quantitati ve consequence estimates presented in these accident analyses were for releases to an uncontrolled area (public).

The worker safety consequence estimates are primarily qualitative. As the design matures , quantitative worker safety consequence analyses will be performed.

Updated frequency (likelihood) and the worker and public quantitative safety consequences will be provided in the Operatin g License Application. Sections 13.2.2 throu g h 13.2.5 pre se nt key representative sequences for radiological and criticality accidents.

Section 13.2.2 discusses s pill s and s pra y accidents with both radiological and criticality safety consequences Section 13.2.3 discusses dissolver offga s accidents with radiological consequences Section 13.2.4 discusse s leaks into auxiliary system accidents with both radiological and criticality safety consequences

Section 13 .2.5 discusses lo ss of electrical power These accidents cover fa ilure of primary vessels a nd piping in the proces s in g areas, lo ss of fission product gas removal efficiency , leaks into auxiliary syste ms , and lo ss of power to the RPF. Section 13.2.6 briefl y presents evaluations of natural phenomena events. The stringent de s ign criteria and requirements for the RPF structure, as discussed in Chapter 3.0, " Design of Structures, Systems, and Components," will require the RPF design to survive certain low-return frequency events. Therefore , the return frequency of most of the external events that the RPF will be designed to withstand are highly unlikel y per Table 13-1. The remainder of the accident sequences, identifi e d in the PHA as requirin g further evaluation, are summarized in Section 13.2.7. Each sequence i s identified and the associated IROFS (if any) listed. The IROFS not discus se d in Sections 13.2.2 through I 3.2.6 are also discussed in this section. Numerous accident sequences with both radiological and criticality safety consequences have been evaluated.

Some accident sequences are bounded or covered in th e prec e ding accident analysis; others , on further evaluation, have an unmitigated likelihood or consequence that does not require IROFS-level controls.

The discussions that follow form the basis for evaluating the accident sequences at thi s point in the RPF project development.

The additional required information will be provided in the Operating License Application. 13-38 NWM I ...... *

NORTHWEST MEDICAL lSOTOP£S NWMl-2013-021, Rev. 2 Chapter 13.0 -Accident Analysis 13.2.1 Reserved 13.2.2 Liquid Spills and Sprays with Radiological and Criticality Safety Consequences Liquid solution spill and spray events causing a radiological exposure hazard were identified by the PHA that represent a hazard to workers from direct exposure or inhalation and an inhalation exposure hazard to the public in the unmitigated scenario.

The PHA also identified fissile solution leaks with worker safety concerns from a solution-type accidental nuclear criticality.

This analysis addresses both of these hazards and identifies controls (in additional to the double-contingency controls identified in Chapter 6.0 , "Engineered Safety Features ," Section 6.3) to prevent an accidental criticality and reduce exposure from a spray or spill. 13.2.2.1 Initial Conditions Initial conditions of the process are described by a tank filled with process solution.

Multiple vessels are projected to be at initial conditions throughout the process , and the PHA reduced the variety of conditions to the following three configurations that span the range of potential initial conditions

A process tank containing low-dose uranium solutions , with no or trace quantities of fission product radionuclides located in a contact maintenance-type of enclosure typical of the target fabrication s ystems A process tank containing high-dose uranium solutions located in a hot cell-type of enclosure typical of the irradiated target dissolution system A process tank containing 99 Mo product solution located in a hot cell-type of enclosure typical of the molybdenum (Mo) purification system (this condition does not lead to a criticality safety concern) In each case , a vessel is assumed to be filled with process solution appropriate to the process location with th e process offgas v entilation s y stem operating. A level monitoring sy s tem is available to monitor tank transfers and stagnant storage volumes on all tanks processing LEU or fission product solutions. Bounding radionuclide concentrations in liquid s treams were developed for five region s of the proces s in NWMJ-2013-CALC-Ol l , Sour ce Term Calculations:

(1) target dissolution, (2) Mo recovery and purification, (3) uranium recovery and recycle , (4) high-dose liquid waste handling , and (5) low-dose li q uid waste handling.

The bounding radionuclide concentrations are based on material balances during the processing of MURR targets , which represent the highest target inventory of fission products entering the RPF due to a combination of high target exposure power and short decay time after end of irradiation (EOI). The predicted radionuclide concentrations are increased by 10 percent to address truncating the radioisotope list tracked by material balance calculations for calculation simplification.

Predicted batch isotope quantities were further increased by 20 percent a s a margin for the radionuclide concentration estimates.

This adds a 1.32 margin to the radionuclides stream compositions presented in Chapter 4.0 , " Radioisotope Production Facility Description

." Two high-dose uranium solutions located in hot cell enclosures have been evaluated for the Construction Permit Application:

Dissolver product in the target dissolution system -Based on a minimum radionuclide decay time of [Proprietary Information], representing the minimum time for receipt of targets at the RPF Uranium separation feed in the uranium recovery and recycle system -Based on a radionuclide decay time of [Proprietary Information], representing the minimum lag storage time required for impure uranium solution prior to starting separation of uranium from fission products 13-39 NWMI ...... *

frlO<<THW£ST MUMCAl tSOTOl'U NWMl-2013-021, Rev. 2 Chapter 13.0 -Acc i dent Analysis The source term used in this analysis is from NWMI-2013-CALC-O

11. The breakdown of the radionuclide inventory used in NWMI-2013-CALC-01 l is extracted from NWMI-2013-CALC-006 , Ov e rall Summary Material Balan ce -MURR Target Batch, using the reduced set of 123 radioisotopes. NWMI-2014-CALC-014 , Selection of Dominant Target Isotopes for NWMI Material Balances , identifies the 123 dominant radioisotopes included in the MURR material balance (NWMI-2013-CALC-006).

NWMI-2014-CALC-014 provides the basis for using the 123 radioisotopes from the total list of 660 radioisotopes potentially present in irradiated targets. The majority of omitted radioisotopes exist in trace quantities and/or decay swiftly to stable nuclides. The reduced set of 123 radioisotopes consists of those that dominate the radioactivity and decay heat of irradiated targets. Bounding solution concentrations from NWMI-2013-CALC

-011 are summarized in Table 13-17. Additional conservatism ha s been incorporated in the dissolver product radionuclide concentrations.

The nominal diluted dissolver product volume is [Proprietary Information]

dissolver batch. Predicted dissolver product concentrations are increased b y a factor of 2.4, to approximate a dissolver product volume of [Proprietary Information]

in a dissolver prior to dilution , producing a uranium concentration of [Proprietary Information] (creating a maximum radioactive liquid source term for the RPF). The criticality evaluations also bound the [Proprietary Information]

batch size. The uranium sepa ration feed composition reflects planned processing adjustments that reduce the solution uranium concentration to [Proprietary Information].

Note that while most of the radioisotopes concentration are noticeably lower in the uranium separation feed stream of Table 13-17 , some daughter isotopes (e.g., americium-241

[24 1 Am]) have increased due to parent decay. Ta b le 13-17. Bo un ding Ra dionu cl i de Liq u id Stream Concentrat i ons (4 pages) Unit operation Decay , hours after EO I Stream description Isotope 24 1Am 1 36 m Ba 1 3 7m Ba 1 3 9 Ba 140Ba 1 4 1ce 14 3 Ce 1 44 Ce 2 42 Cm z 4 3 Cm 244 Cm 1 3 4Cs I 34 m Cs 1 36 Cs 137 Cs 1 ss Eu 1s6Eu Target dissolution

[Proprietary Information]

Dissolver product Bounding concentration (Ci/L) [Proprietary Information]

[Proprietary Information]

[Propri e tary Information]

[Proprietary Information]

[Proprietar y Information]

[Proprietary Information]

[Proprietar y Information]

[Proprietary Information]

[Proprietar y Information]

[Proprietary Information]

13-4 0 Uranium recovery and recycle [Proprietary Information]

Uran iwn separation feed Bounding concentration (Ci/L) [Proprietary Information]

[Proprietary Information]

NWM I ...... ' * *

NOmtWtST MEO.CAL lSOTDPfS NWMl-2013-021, Rev. 2 Chapter 13.0 -Accident Analysis Table 13-17. Bounding Radionuclide Liquid Stream Concentrations (4 pages) Unit operation Decay , hour s after EO I Stream description Isotope 1 s1E u 1291 1 30 1 1 311 1 32 J 13 2 m I 1 33 1 1 33 m I 1 3 4 J 13S I 83 m Kr 85 Kr 8Sm Kr 87 Kr 88 Kr 140La 1 4 1La 14 2 La 99Mo 9 5 Nb 95m Nb 96Nb 97 Nb 9 7 mNb 14 1 Nd 236mN p 231 Np 23g Np 23 9Np 23 3 pa 234 P a 234m Pa 112p d 14 7 Pm 14 8 Pm 148mpm Target dissolution

[Proprietar y Information]

Dissolver product Bounding concentration (Cill) [Proprietary Information]

[Proprietary Information]

[Proprietar y Information]

[Proprietary Information]

[Proprietar y Information]

[Proprietary Information]

[Proprietar y Information]

[Proprietary Information]

[Proprietar y Information]

[Proprietary Information]

[Proprietar y Information]

[Proprietary Information]

[Proprietary lnfonnation]

[Proprietar y Information]

[Proprietary Information]

[Proprietar y Information]

[Proprietary Information]

[Proprietar y Information]

[Proprietary Information]

[Proprietar y Information]

[Proprietary Information]

13-41 ' Uranium recovery and recycle [Propri etary Information]

Uranium separation feed Bounding concentration (Ci/L) [Proprietary Information]

[Proprietary Information]

[Proprietary In formation]

[Proprietary Information]

NWMl-2013-021 , Rev. 2 Chapter 13.0 -Accident Analysis Table 13-17. Bounding Radionuclide Liquid Stream Concentrations (4 pages) Unit operation Decay, hours after EOI Stream description Isotope 14 9 Pm I SO Pm I S IPm 14 2 Pr 14 3 pr 144Pr J4 4m pr 14 5 Pr 23 8 pu 239p u 2 40 pu 24 1Pu 10 3 m Rh I OS Rh 1 06 Rh 106mRh 10 3 Ru 1osRu 106Ru 122 Sb 1 24 Sb 125 Sb 1 2 6 Sb 1 21 sb 1 2 s sb 1 2s msb 1 2 9 Sb 1 s1 sm 153 Sm 156 Sm s9s r 9o sr 9 1sr 92 Sr 99 Tc 99m rc Target dissolution

[Proprietary Information]

Dissolver product Bounding concentration (Ci/L) [Proprietary Information]

[Proprietary Information]

[Proprietary Inform a ti on] [Proprietary Information]

[Proprietary Information]

[Proprietar y Information]

[Proprietary Information]

13-42 Uranium recovery and recycle [Proprietary Information]

Uranium separation feed Bounding concentration (Ci/L) [Proprietary Information]

[Proprietary Information]

NWM I * * *

NORTKWtU MEDtcAL ISOTOPES NWMl-2013-021, Rev. 2 Chapter 13.0 -Accident Analysis Table 13-17. Bounding Radionuclide Liq u id Stream Concentrations (4 pages) Unit operation Decay, hours after EOI Stream description Isotope 1 2smre 127 Te 1 27mre 1 29 Te 12 9 mre n1re 131mre 1 32Te 133 Te 13 3 mre 1 34 Te 23 1Th 23 4Th 232u 23 4u 23su 236 u 237 u 23su 131mxe 133 Xe 1 33 mxe 135 Xe I 3s mxe 89my 90 y 90 my 9Iy 9 Jmy 92 y 93 y 93Zr 9s zr 9 7 Zr Totals Target dissolution

[Proprietary Information]

Dissolver product Bounding concentration (Ci/L) [Proprietary Information]

[Proprietary Information]

[Proprietary lnformati on] [Proprietary Information]

[Proprietar y Information]

[Proprietary Information]

[Proprietar y Information]

[Proprietary Information]

[Proprietar y Information]

[Proprietary Information]

[Proprietar y Information]

[Proprietary Information]

[Proprietar y Information]

[Proprietary Information]

[Proprietar y Information]

[Proprietary Information]

[Proprietary lnformati on] [Proprietary Information]

[Proprietary Information]

Uranium recovery and recycle [Proprietary Information]

Uranium separation feed Bounding concentration (Ci/L) [Proprietary Information]

[Proprietary Information]

[Proprietar y Information]

[Proprietary Information]

[Proprietar y Information]

[Proprietary Information]

[Proprietar y Information]

[Proprietary Information]

[Proprietar y Information]

[Proprietary Information]

[Proprietar y Information]

[Proprietary Information]

Source: Table 2-1 ofNWMI-2013-CALC-O 11 , Source T er m Cal c ulation s, R ev. A, Northwest M e dic al I soto pe s, LLC , Corvallis , Oregon , February 2015. EO I = end of irradiation.

13-43 NWMI ...**... * * *

NORTHWUT MEDtCAl ISOTDr£S 13.2.2.2 Identification of Event Initiating Conditions NWMl-2013-021, Rev. 2 Chapter 13.0 -Accident Analysis The accident initiating event is generally described as a process equipment failure, but also could be operator error or initiated by a fire/explosion.

Multiple mechanisms were identified during the PHA that resulted in the equivalent of a failure that spills or sprays the tank contents , resulting in rapid and complete draining of a single tank to the enclosure in the vicinity of the tank location.

13.2.2.3 Description of Accident Sequences The accident sequence for a tank leak is described as follows. 1. Process vessel fail or personnel error causes the tank contents to be emptied to the vessel enclosure floor in the vicinity of the leaking tank. 2. Tank liquid level monitoring and liquid level detection in the enclosure floor sump region alarms , informing operators that a tank leak has occurred.

3. Processing activities in the affected system are suspended based on location of the sump alarm. 4. Operators identify the location of the leaking vessel and take actions to stop additions to the leaking tank. 5. A final stable condition is achieved when solution accumulated in the sump has been transferred to a vessel availab l e for the particular sump material and removed from the enclosure floor. The accident sequence for a spray leak is similar to that of a tank leak and is described a s follows. 1. The process line , containing pressurized liquid, ruptures or develops a leak during a transfer , spraying solution into the s ource or receiver tank enclosure and transferring leaked material to an enclosure floor in the vicinity of the leak. 2. Transfer liquid level monitoring and liquid level detection in the enclosure floor s ump region alarms , informing operators that a leak has occurred. 3. Processing activities in the affected system are suspended based on location of the sump alarm. 4. Operators identify the location of the leaking vessel and take actions to ensure that the motive force of the leaking transfer line has been deactivated.

5. A final stable condition is achjeved when solution accumulated in the sump has been transferred to a vessel available for the particular sump material and removed from the enclosure floor. Maintenance activities to repair the cause of a tank or spray leak are initiated after achleving the final stab l e condition. 13.2.2.4 Function of Components or Barriers The process vessel enclosure floor , walls, and ceiling will pro v ide a barrier that prevents transfer of radioactive material to an uncontrolled area during a liquid spill or spray accident.

For accidents involving high-dose uranium solutions and 99 Mo product solution , the process vessel enclosure floor , walls, and ceiling will provide shielding for the worker. The enclosure structure barriers are to function throughout the accident until (and after) a stable condition has been achieved.

The process enclosure secondary confinement (or ventilation) system will provide a barrier to prevent transfer of radioactive material to an uncontrolled area during a liquid spill or spray accident from radioactive material in the airborne particulate and aerosols generated by the event. The s econdary confinement system is to function throughout the accident until a stab le condition has been achieved.

13-44 NWMl-2013-021, Rev. 2 Chapter 13.0 -Accident Analysis The process enclosure sump system represents a component credited (part of the double-contingenc y analysis) for preventing the occurrence of a solution-t ype accidental nuclear criticality due to spills or sprays of fissile material.

The sum p system is to function throughout the accident until a stable condition has been achieved.

13.2.2.5 Unmitigated Likelihood A spill or s pray can be initiated by operations or maintenance personnel error or equipment failures.

Failure rates for tanks, vessels, pipes, and pumps are estimated from WSRC-TR-93-262, Savannah Ri ver Site Generic Data Base Development.

Table 13-2 (Section 13.1.1.1) shows qualitative guidelines for applying the likelihood categories.

Operator error and tank failure as initiating events are estimated to have an unmitigated likelihood of " not unlikel y." Additional detailed information describing a quantitative evaluation , including assumptions , methodology, uncertainties , and other data , will be developed for the Operating License Application.

13.2.2.6 Radiation Source Term The following source term descriptions are ba sed on inform at ion developed for the Construction Permit Application.

Additional detailed information describing source terms will be developed for the Op erat ing License Application.

13.2.2.6.1 Direct Expos ure Source Terms Liquid spill source terms are dependent on the vesse l lo cation in the proce ss system. The following source terms describe the three configurations used to span the range of initial conditions:

Low-dose uranium solutions were bounded by the maximum projected uranium concentration solution in the target fabrication system. The primary attribute of low-dose uranium solutions used for consideration of direct exposure consequences is that fission products have been separated from recycled uranium to allow contact operation and maintenance of the target fabrication system within ALARA (as low as reasonably achievable) guidelines.

Chapter 4.0 , Section 4.2 , s hows that a pencil tank of this material would be less than 1 millirem (mrem)/hr; therefore, no radio l ogical IROFS are required for this st ream. High-dose uranium solutions were bounded by a spill from the irradiated target dissolver after dissolution is complete.

Dissolution of the targets produces an aqueous solution containing uran y l nitrate , nitric acid , and fission products. The primary attribute of high-dose uranium solutions used for consideration of direct exposure consequences is that equipment operation and maintenance must be conducted in a shielded hot cell environment due to the presence of fission products. 99 Mo product solution was bounded by a sma ll solution volume (less than 1 L) containing the weekly inventory of product from processing MURR targets. The product i s an aq ueous solution containing

-0.2 M sodium hydroxide (NaOH) with a total inventory of 1.3 x 10 4 curies (Ci) 99 Mo. 13.2.2.6.2 Confinement Release So urce Terms Confinement release so urce terms are based on the five-factor algebraic formula for calculating source terms for airborne release accidents from NUREG/CR-6410 , as shown b y Equation 13-1. where , ST MAR ST= MARxDRxARFxRFxLPF Source term (activity)

Material at risk (ac tivit y) 13-45 Equation 13-1 DR ARF RF LPF Damage ratio (dimens ionles s) Airborne release fraction (dimensionless)

Respirable fraction (dimensionless)

Leak path factor (d imen s ionl ess) NWMl-2013-021, Rev. 2 Chapter 13.0 -Accident Analysis Confinement release source terms for spray u sed the source term parameters listed in Table 13-18. Four source term cases were developed for evaluation based on the two boundin g liquid concentrations shown in Table 13-17 and the so urce term parameter alternatives. Parameter 3 Material at risk (MAR) Damage ratio (DR) Airborne release fraction (ARF) Respirable fraction (RF) Leak p at h factor (LPF) Table 13-18. Source Term Parameters Unmitigated spray release Table 1 3-17 1.0 0.0001 (I .0 for Kr , Xe, and iodine)b 1.0 1.0 Mitigated spray release Table 13-17 1.0 0.0001 (1.0 for Kr , Xe , and iodin e)h 1.0 0.0005 (1.0 for Kr , Xe; 0.1 for i o din e) So ur ce: Table 2-1 ofNWMI-2015-RPT-009 , Fi ss ion Produ ct R e l ease Evaluation , Rev. B , Northwest Medic a l I soto p es, LL C, Corvallis, Oregon , 20 1 5. a Parameter definiti ons derived from NUREG/CR-6410 , N ucl e ar Fuel Cy cl e Fa c ili ty Acc id e nt A nal ys i s Handbook , U.S. Nuclear R egu l ato r y Co mmi ss ion , Office of Nuclear M a terial Safet y and Safeg uards , W as hin gto n , D.C., March 1998. b Acc id e nt do se conseq uen ces wer e found to be se nsitive to iodin e so urce term param ete rs. Furth e r work may a llow for a low er iodine ARF. Kr = krypton. Xe = xe n on. The DR was set to 1.0 for all cases. The assumed v olume was 100 L of solution contained b y a vessel being affected by the spill or spray release. The ARF and RF values are functions of the rele ase mechanism and do not e nter into consideration for a mitigated versus unmitigated release. Thus , for both the unmitigated and mitigated cases , the ARF and RF were set to representative values based on the guidance in NUREG/CR-6410 and DOE-HDBK-3010 , DOE Handbook-Airborne Rel e ase Fractions/Rates and R espirab l e Fra c tions for Nonreacto r Nuclear Faciliti e s. A spray release due to rupture of a pressurized pipe (transfer line) is modeled as depressurization of a liquid through a leak below the liquid surface le ve l. Both NUREG/CR-6410 and DOE-HDBK-3010 report an ARF of 1 x 10-4 and a RF of 1.0 for a spray leak in vo l v in g a low temperature aqueous liquid. These values take into consideration upstream pressures as high as 200 pounds (lb)/square inch (in.2) gauge. The spray mechani s m is also bounded b y a droplet size distribution produced from commercial spray nozzles. This approach is conservative , as the effective nozzle created b y a pipe failure is unlikel y to be optimized to the extent of a manufactured s pray nozzle. Therefore, an ARF of 1 x 10-4 and a RF of 1.0 were used for all isotopes , except iodine and the noble gas fission products Kr and Xe. Radioisotopes of Kr , Xe, and iodine were assigned an ARF of I .0 for all cases. For the unmitigated evaluations , the LPF was set to 1.0 , since the unmitigated release sce nario credits no confinement measure s (i.e., no credit was taken for any aspect of the facility design or equipment performance).

The gravitational se ttling associated with flow throughout the faci lit y and the removal action of high-efficienc y particulate air (HEPA) filtration ma y be lumped into an effective va lue for LPF. The performance of different filtration systems is presented in Appendix F ofDOE-HDB K-3010. For scoping purposes , a HEP A filtration efficiency of 99. 95 percent was selected for all mitigated cases , which corresponds to an LPF of 0.0005. 13-46 NWMI ...... *

NCMtTHWEST MEDICAL ISOTOP{S NWMl-2013-021 , Rev. 2 Chapter 13.0 -Accident Analysis The HEPA filter LPF was applied to all isotopes except Kr, Xe, and iodine. An LPF of 1.0 was selected for Kr and Xe in the mitigated spray release evaluation, assuming these isotopes behave as a gas when airborne and are not removed by HEP A filtration or sufficiently retained on the high-efficiency gas adsorption (HEGA) modules. The mitigated analysis credits an iodine remova l capability in the facility ventilation exhaust gas equipment, with an iodine removal efficiency of 90 percent. The credited removal efficiency corresponds to an LPF of 0.1 for iodine due to the HEGA modules co-located with the HEPA filters. 13.2.2.7 Evaluation of Potential Radiological Consequences Confinement release consequence estimates for the Construction Permit Application are based on NUREG-1940 , RASCAL 4: Description of Models and Methods, and Radiological Safety Analysis Code (RSAC), Version 6.2 (RSAC 6.2). Additional detailed information describing validation of models , codes , assumptions, and approximations will be developed for the Operating License Application. 13.2.2. 7.1 Direct Expos ure Consequences The potential radiological exposure hazard of liquid spills depends on the vessel location in the process system. Low-dose uranium solutions are generally contact-handled, and direct radiation exposure to the worker is expected to be s lightl y elevated but well within ALARA guidelines.

Therefore , no IROFS are required to control radiation exposure from spilled low-dose uranium solutions.

Vessels located within hot cells require shielding to control worker radiation exposure independent of whether process solution is contained in the vessel or spilled to the enclosure floor. High-dose uranium solutions are assumed to require hot cell shielding.

Spills of 99 Mo solution from the Mo recovery and purification processes, and during handling prior to shipment of the product, involve product solution that contains high-dose 99 Mo. The direct whole-body exposure from radiation does not change from the normal case and must always be shielded to reduce the dose rate for workers to ALARA. As a pr e! iminary estimate using a point-source dose rate conversion factor for 99 Mo of 0.112924 roentgen equivalent man (rem)/hr at 1 meter (m) per Ci 99 Mo, the unshielded dose rate for the product is: MAR= J.3 x J0 4 Ci 99 Mo. 99 Mo dose rate at 1 m = l.30 x J0 4 Ci 99 Mo x 0.1129 rem/hr/Ci 99 Mo = J.5 x 10 3 rem/hr In a very short period of time, a worker can receive a significant intermediate or high consequence dose. Therefore, both high-dose uranium and 99 Mo product solution vessels must be located in hot cells for normal operations to control the direct exposure to workers. Based on the analysis of several accidental nuclear criticalities in industry , LA-13638 , A Revi e w of Criticality Accidents, ident ifi es that a uranium solution criticality can yield between 10 1 6 to 10 1 7 fissions.

Dose rates for anyone in the target fabrication area can have high consequences.

Consequences for a shielded hot cell criticality will be developed for the Operating License Application.

13.2.2. 7.2 Confinement Release Consequence Receptor dose consequences were originally evaluated in NWMI-2015-RPT-009 , Fission Product Release Evaluation , using the RASCAL code. Since the submission of the application, NWMI has selected RSAC 6.2 for off-site accident consequence modeling.

For the liquid spills and spray accident , NWMI has rerun the dissolver product off-site dose calculations using RSAC 6.2. Four release consequence estimates were prepared to support the Construction Permit Application based on unmitigated and mitigated spray release events using the two liquid radionuclide concentrations shown in Table 13-18. The RSAC inputs for the dissolver product accident are listed below , and the RASCAL input s for the high dose uranium solution are listed in Table 13-19. The uranium feed modeling will be rerun using RSAC 6.2 as part of the Operation License Application.

13-47 NWMl-2013-021, Rev. 2 Chapter 13.0 -Accident Analysis Table 13-19. Release Consequence Evaluation RASCAL Code Inputs Input Primar y tool Event type Facility location a County Time zo ne Latitude/longitude E le va tion Plume rise Meteorology Receptor distance Dose conversion factors Description STDose -Source term to dos e option se l ecte d as the primary tool in RASCAL for all cases. Other release -RASCAL includes separate models for nuclear power plant accidents involving spent fuel , accidents involving fuel cycle activities, and other radioactive material releases at non-reactor facilities.

The other radioactive material releases option was selected for all cases. Columbia, Mi sso uri Boone Centra l 38.9520° N/92.3290° W 231 m None -For scoping purposes, the enthalpy and momentum of the RPF stack exhaust was assumed negligible.

Summer-night-calm

-Selected for sco pin g purposes an d features w ind s peed of 6.4 km/hr (4 mi/hr), Pasquill Class F sta bility , no precipitation, relative humidi ty of 80%, and ambient temperature of l 2.8°C (55°F). Low wind spee d and stable conditions selected to provide maximum do se to near-field receptors. 100 m -Selected to approximate site boundary.

Input represents minimum value for RASCAL input. ICRP-72b -Selected as the mo s t current and authoritative set of dose conversion factors available. So urc e: Ta ble 2-1 ofNWMl-20 l 5-RPT-009 , Fission Produ ct R e l ease Evaluation, R ev. A, Northwest Medical I sotopes, LLC , Corva lli s, Oregon, February 20 1 5. a Locat ion informati on obtained from Wikipedia.

b I CRP-72, Age-Dependen t Dos es t o th e Memb e r s of th e Public from Intake of Radi onuclides

-Part 5 Co mpilation of In gestion and Inhal ation Coeffic i e nts , International Co mmi ssio n on Radi o lo gica l Protecti on, Ottawa, Canada , 1995. RASCAL = Radiolo gica l Assessment Syste m for RPF = R a dioi sotope Producti on Facility.

Co n sequence Analysis.

RSAC 6.2 was used to model the dispersion resulting from a spray leak. The following parameters were used for model runs: Mixing depth: 400 m (1 ,3 12 feet [ft]) (default)

Air densit y: 1 ,2 40 g/cubic meter [m 3] (1.24 ounce [oz]/cubic feet [ft 3]) (sea level) Pasquill-Gifford a (NRC Regulatory Guide 1.145 , Atmosp h er i c Dispersion Models for Pot entia l Acci dent Consequence Assessme nts at Nuclear Pow er Plants) No plume rise (i.e., buoyancy or stack momentum effects) No plume depletion (wet or dry deposition) 1-hr release (constant release of all activity) 1-hr exposure ICRP-30, Limits for Intak es of Radionuclides by Workers , inhalation model Finite cloud immersion model Breathing rate: 3.42E-4 m 3/second (sec) (1.2E-2 ft 3/sec) (ICRP-30 heavy activity) 13-48 NWMI * * *

NOM'HW£ST MEDICAL ISOTOKt NWMl-2013-021, Rev. 2 Chapter 13.0 -Accident Analysis Consequence evaluation re s ults a re shown in Figure 13-2 and Table 13-20 for a 100 L (26.4 ga l) s pra y release event. NWMI is considering the unmitigated spray release of di sso l ve r product so lution as an off-site public interm e diate consequence event (pending completion of the final safety analys i s). The nearest permanent resident, at 432 m (0.27 mil es [mi]), di sso lv e r product s pra y unmitigat e d dose estimate i s 300 mrem , while the maximum receptor loc at ion (1 , 100 m [0.68 mi]) has a total effective dose equivalent (TEDE) of 1.8 rem. The miti ga ted co n se quenc es are an order of magnitude lower due to the credited IROFS in the Zone I ex haust system. Therefore, the near est p erma nent r es ident ( 432 m [0.27 mi]) dissolver product s pra y mitigated do se es timat e is 30 mrem, w hil e the ma x imum receptor location (1, I 00 m [0.68 mi]) h as a TEDE of 0.18 rem. 2.0 1.8 1.6 1.4 E 1.2 Q) l-e) 1.0 t/l 0 0 0.8 0.6 0.4 0.2 _._Inhalation CEDE _._External EDE 100 200 300 400 500 600 700 800 900 1000110012001300140015001600 Distance, meters Figure 13-2. Unmitigated Off-Site Dose of Dissolver Product Spray Leak Accident Tabl e 13-20 shows that the uranium se parati o n feed solution s pra y r e lea se unmitigated do se i s below the imm e diate consequences thre s hold s of I 0 CF R 70.61. Even though thi s r ece ptor do se is at 100 m , th e uranium feed mod e ling will b e r erun usin g RS AC 6.2 as part of the Operation License Ap plication. Table 13-20. Uranium Separations Feed Spray Release Consequence Summary at 100 Meters Process stream Case Mitigation Receptor dose, total EDE Stack height Release mechanism Release duration Uranium separations feed Unmitigated 0.078 rem I 0 m (33 ft)" Spray leak, 100 L 1 hr Miti ga ted 0.006 rem 23 m (75 ft) Source: T ab l e 2-1 a nd Tab l e 2-7 ofNWM I-20 15-RPT-009 , Fissio n Produ c t R e l ease Eva l uation , R ev. A, Nort h west Medical I so t opes, LLC , Corvallis , Oregon , February 20 1 5. a Lowes t va lu e fo r p lum e h e ight avai l ab l e as input to RASCAL and reco mm e nd e d b y help tile as in put mod e ling a grounl eve l re l ease. EDE = effective dose e qui va l ent. RAS CA L = R a di o l o g ical Assessment Syste m for Conse qu e n ce Anal ys i s. 13-49 NWMl-2013-021 , Rev. 2 Chapter 13.0 -Accident Analysis 13.2.2.8 Identification of Items Relied on for Safety and Associated Functions Unmitigated spill and s pray relea s es have the potential to produce direct exposure and confinement releases with high consequence to workers and the public. Hot cell shielding is designed to provide protection from uncontrolled liquid spills and sprays that result in redistribution of high-dose uranium and 99 Mo product solution in the hot cell. From a direct exposure perspective , a liquid spill does not represent a failure or adverse challenge to the hot cell shielding boundary function.

However, the hot cell shielding boundary must also function to prevent migration of liquid spills to uncontrolled areas outside the shielding boundary.

Liquid spill and spray-type releases occur as a result of the partial failure of process vessels to contain either the fissile solution (for areas outside of the hot cell) or to contain fissile or high-dose radiological solutions (for areas inside the hot cell). In either case, the process vessel spray release results in an event that carries with it a higher airborne radionuclide release magnitude than a simple liquid spill. The type release also carries the extra hazard of potential chemical burns to eyes and skin, with the complication of radiologica l contamination. Consequent l y , spray protection is a secondary safety function needed to satisfy performance criteria.

The liquid spill and spray confinement safety function of the hot cell liquid confinement boundary is then credited for confining the spray to the hot cell and protecting the worker from sprays of radioactive caustic or acidic solution with the potential to cause intermediate or high consequences.

The airborne filtering safety feature of the hot cell secondary confinement boundary is credited with reducing airborne concentrations in the hot cells to levels outside the hot cell boundary , which are below intermediate consequence l evels for workers and the public during the event. Three IROFS are identified to control liquid spill and spray accidents from process vessels. * *

IROFS RS-01 , " Hot Cell Liquid Confinement Boundary" IROFS RS-03 , " Hot Cell Secondary Confinement Boundary" IROFS RS-04 , "Hot Cell Shielding Boundary" Liquid spill and spray events involving so luti on s containing fissile material have the potential for producing liquid nuclear criticalities that must be prevented.

The followin g IROFS are identified to control nuclear criticality aspects of the liquid spill and spray events. IROFS CS-07 , " Pencil Tank and Vessel Spacing Control Using Fixed Interaction Spacing of Individual Tanks or Vessels" IROFS CS-08 , "Floor and Sump Geometry Control on Slab Depth , Sump Diameter or Depth for Floor Spill Containment Berms" IROFS CS-09 , "Do uble-Wall Piping" Functions of the identified IROFS are described in the following sections.

13.2.2.8.1 IROFS RS-01, Hot Cell Liquid Confinement Boundary IROFS RS-01 functions to mitigate the impact of liquid spills from process vessels in the hot cells. As a passive engineered control (PEC) and safety feature , the hot cell liquid confinement boundary will provide an integrated system of features that protects workers and the public from the high-dose radiation generated during primary confinement releases of primarily liquid so lution s during the 99 Mo recovery process. The hot cell liquid confinement boundary will also protect the environment from releases of product solution from the primary confinement of the proce ssing vessels. In addition , the barrier will provide a function of confining spills of irradiated LEU target solid material in some of the irradiated target handling hot cells. 13-50 NWMl-2013-021 , Rev. 2 Chapter 13.0 -Accident Analysis The primary safety function of the hot cell liquid confinement boundary is to capture and contain liquid releases and to prevent those releases from exiting the boundary, causing high dose to workers or the public, or contaminating the environment.

A seco nd ary function of the liquid confinement boundar y is to pre vent contact chemical exposure to workers fro m acidic or caustic sol ution s contaminated with lic ensed material that exceeds the performance criteria esta bli s hed by NWMI for the RPF. As a PEC to contain sp ill s and sprays of high-do se product so lution , the hot cell liquid confi nement boundary will consist of sea l ed flooring with multiple la ye r s of protection from release to the environme nt. Various areas will be diked to contain specific releases, and sumps of appropriate de sign w ill be provided with r emote-op erate d pump s to mitigate liquid sp ills by captur ing the liquid in appropriate safe-geometry tanks. Additional IROFS apply to the flooring and sumps for criticality safety doubl e-co ntingenc y controls in some areas. In the 99 Mo purification product and sample hot cell, sma ll er confinement catch basins will be provided und er points of credible spill potential in addition to use of a sea led floor. Entryway doors into a de s ignat e d liquid confinement area will be sea led against credible liquid leaks to outside the boundar y. This continuous barrier is also credited to pr eve nt sp ill s or sprays of hi g h-dose product so lutions that are acidic or caustic from causing adverse exposure to personnel thr oug h direct contact with ski n , eyes , and mucu s membranes, where the combination of the c h emica l exposure and the radiological contamination wo uld lead to ser iou s injury and lon g-la sting effects or even death. Specific design features of the liquid confinement barrier, a liquid barri er to uncontrolled areas and worker radiation expos ur e from leaked so lution , include: * * * *

Co ntinuou s , impervious floor with an acid-or ca u stic-r es i sta nt surface finish Hot cell wa ll s and ceiling designed to contro l worker dose from liquids acc umul ate d in s ump s Monitors with alarms to indicate a liquid release has occurred Sealed penetrations de signe d to prevent liquid l eaks through th e barrier to uncontroll e d areas Sump solution collection vesse ls for accumulating le ake d proce ss so lution 13.2.2.8.2 IROFS RS-03, Hot Cell Secondary Confinement Boundary IROFS RS-03 functions to mitigate the impact of liquid spills and sprays from process vesse ls in the hot cells. As a system of PECs and AECs, the hot ce ll seco ndar y confinement boundary safety feature is engineered to provide backup to credible up sets in the primary confinement system u sing the following sa fety functions:

Provide negative air pre ssure in the hot cell (Zone I) relative to lower zo nes outside the hot cell using exhaust fans equipped with HEP A filters and HEGA modul es to remove the relea se of radionuclides (bo th particulate and gaseous) to outside the primar y confinement boundary to below 10 CF R 20 relea se limits durin g normal and abnormal operations.

Com pon ents credi ted include: Zone I Inlet HEPA filters to provide a n efficiency of 99 .97 percent for remo va l of radiological particulates from the air that ma y reverse flow from Zone I to Zone 11 Zone I ducting to ensure that negati ve air pre ss ure can be maintain e d by conveying exhaust air to the stack Zon e I exha u st train HEPA filter s to provide 99.97 percent removal of radiological particulate s from the air that flows to the s tack Zone I exha u st train HEGA modul es to provide 90 percent removal of iodine gas from the air that flows to the stack Zone I exha u st stack to provide dispersion of radionuclides in normal an d abno rmal relea ses at a discharge point of 22.9 m (75 ft) above the building ground level 13-51

NWMI ...... *

ffOATHWEST MEDtCAl ISOTDP(S NWMl-2013-021, Rev. 2 Chapter 13.0 -Accident Analysis Stack monitoring and interlocks to monitor discharge and signal changing on service filter trains during normal and abnormal operations As a system of PECs and AECs , the purpose of this IROFS is to mitigate high-dose radionuclide releases to maintain exposure to acceptable levels to both the worker and the public in a highly reliable and available manner. The hot cell secondary confinement boundary will perform this function using the following engineered features to ensure a high level of reliability and availability.

As a PEC , the hot cell floor, walls, ceilings , and penetrations are designed to provide an air intrusion barrier sufficient to allow the exhaust system to maintain negative air pressure under normal and credible abnormal conditions.

This barrier is not required to be air-t ight, but must be controlled to the extent that the design capacity of the exhaust fans can maintain negative pressure.

Design features associated with this function include airlocks for normal egress, cask and bagless transfer ports that can only open when the cask or container is properly sealed to the port , and appropriate l y sized venti lation ports between zones. Along with the AECs of the filtered ventilation system , this boundary will provide secondary confinement and prevent uncontrolled release of general radiological airborne gases and particulates that escape the primary confinement to reduce releases to the monitored stack to acceptable release levels during normal and abnormal operations. The Zone I exhaust system will serve the hot cell , high-integrity canister (HIC) loading area , and solid waste loading area. This exhaust system will maintain Zone I spaces at negative pressure with respect to atmosphere.

All make-up air to Zone I spaces will be cascaded from Zone II spaces. HEPA filters will be included on both the inlet and outlet ducts to Zone I. The hot cell outlet HEPA filters will minimize the spread of contamination from the hot cell into the ductwork leading to the exhaust filter train but are not credited with reducing exposure to workers and/or the public. The hot cell inlet HEPA filters will prevent contamination spread during an upset condition that results in positive pressurization of Zone I spaces with respect to Zone II spaces. The process off gas subsystem will enter the Zone I exhaust subsystem just upstream of the filter train. The exhaust train outlet HEP A filters wi II prevent contamination from entering the stack. The stack will disperse radiological gases and particulate to levels below release limits in normal operations and below intermediate consequence levels during process upsets. As an AEC, the hot cell secondary confinement system will also serve as backup to the primary offgas treatment system by providing a backup stage of carbon retention bed removal (consisting of an iodine removal) capacity before exhausting into the ventilation system described above. This system will have limited availability for iodine adsorption if the primary system fails. 13.2.2.8.3 IROFS RS-04, Hot Cell Shielding Boundary IROFS RS-04 functions to prevent worker dose rates from exceeding exposure criteria due to the presence of radioactive materials in the hot cell vessels before or after a liquid spi II accident.

As a PEC and safety feature , the hot cell shielding boundary wi ll provide an integrated system of features that protect workers from the high-dose radiation generated during the 99 Mo recovery process. The primary safety function of the hot cell shielding boundary will be to reduce the radiation dose at the worker/hot cell interface to ALARA. The shield will also protect workers and the public at the controlled area or exclusion area boundary.

13-52 NWMl-2013-021, Rev. 2 Chapter 13.0 -Accident Analysis The hot cell shielding boundary will provide shielding for workers and the public during normal operations to reduce worker exposure to an average of 0.5 mrem/hr , or less , in normally accessible workstations and occupied areas outside of the hot cell. The hot cell shielding boundar y will provide shielding for workers and the public during process upsets to reduce the worker exposure to a TEDE of 5 rem , or less, at workstations and occupied areas outside of the hot cell. As a PEC , shielding will be provided by a thick concrete, stee l-reinforced wall with steel cladding that reduces the normally expected operational exposures from within the boundary to an average of 0.5 mrem/hr, or less , outside of the boundar y. Where direct visual access is required, leaded-glass windows with appropriate thicknesses will be u se d to reduce normally expected operational exposures from within the boundar y to an average of 0.5 mrem/hr , or le ss, outside of the boundar y. Some shielding will be movable , such as around the high-dose waste cask loading area. Where penetrations are required , the engineered design provides for access-controlled, non-occupied corridors or airlocks where potential radiation streaming is safe l y mitigated by multiple la yers of shie ldin g or through a torturous path. The shielding is also designed to reduce the exposure from postulated upsets within the hot cell shielding boundary to less than a low-con se quence exposure to workers and the public of 5 rem , or less , per incident.

These incidents include spills, sprays, fires, and other releases of radionuclides contained within the boundary.

The shield may be divided into protection areas for the purposes of applying limiting conditions of operation.

Each shie lded protected area will be operable when the equipment in that area i s in the operating or sta ndb y mod es. 13.2.2.8.4 IROFS CS-07, Pencil Tank and Vesse l Spacing Control using Fixed Interaction Spacing of lndivid ual Tanks or Vessels IROFS CS-07 functions to ensure that potential interactions between full vessels and a sump filled by a liquid spill or spray have been considered to pre ve nt a nuclear criticality event. As a P EC, pencil tanks and other standalone vessels (controlled with safe geometry or vo lum e constraints) are designed and fabricated with a fixed interaction spacing for safe storage and processing of the fissile solutions. The safety function of fixed interaction s pacing of individual barrels in pencil tanks and between other single processing vessels or components is designed to minimize interaction of neutrons between vessels such that under normal and credible abnormal proce ss upsets , the syste ms will remain subcritical.

The fixed in t eraction control of tanks, vessels, or components containing fissile solutions will pre ve nt accidental nuclear criticality, a high consequence event. The fixed interaction control distance from the safe slab depth spill containment berm is s pecified where applicable.

13.2.2.8.5 IROFS CS-08, Floor and Sump Geometry Control on Slab Depth, Sump Diameter or Depth for Floor Spill Containment Berms IROFS CS-08 functions to ensure that sump designs have been considered to prevent a nuclear criticality event b y geometry if filled with liquid from a sp ill or spray release. As a PEC, the floor under designated tanks , vessels, and workstations will be constructed with a spill containment berm that maintains a geometry slab depth to be determined with final design , and one or more collection sumps with diameters or depths to be determined in final design. The safe ty function of this spill containment berm is to safely contain spi lled fissile solution from systems overhead and prevent an accidental nuclear critica lit y if one of the tanks or related piping leaks , ruptures, or overflows (if so equipped with overflows to the floor). Each spill containment berm will be sized for the largest single credible leak associated with the overhead systems. The interaction distance for the spill containment area is provided in IROFS CS-07. The sump will ha ve a monitoring system to alert the operator that the IROFS has been u se d and ma y not be available for a follow-on event. A spill containment berm will be operable if it contains reserve volume for the largest single credib l e spill. Spill containment berm sizes and location s will be determined by the final design. 13-53 NWM I ...... ' ! e * ' NORTHWtn MEDICAL ISOTOPU NWMl-2013-021, Rev. 2 Chapter 13.0 -Accident Analysis 13.2.2.8.6 IROFS CS-09, Double-Wall Piping IROFS CS-09 functions to control liquid spills or sprays in a similar manner to IROFS CS-08. As a PEC , the piping system conveying fissile solution between credited locations will be provided with a wall barrier to contain any spills that may occur from the primary confinement piping. IROFS CS-0 9 is used at locations that pass through the facility where creating a spill containment berm (IROFS CS-08) under the piping is neither practical nor desirable for personnel chemical protection purposes.

The double-wall piping arrangement is designed to gravity drain to a safe-geometry set of tanks or to a geometry containment berm. The safety function of this PEC is to safely contain spilled fissile solution from system piping and prevent an accidental nuclear criticality if the primary confinement piping leaks or ruptures. The double-wall piping arrangement will maintain the safe-geometry diameter of the solution.

The secondary safety function of the double-wall piping is to prevent personnel injury from exposure to acidic or caustic licensed material solutions that are conveyed in the piping. Defensive-in-Depth The following defense-in-depth features were identified by the liquid spill and spray accident evaluations.

Alarming radiation area monitors will provide continuous monitoring of the dose rate in occupied areas, and alarm at an appropriate setpoint above background.

Continuous air monitoring will be provided to a l ert operators of high airborne radiation levels that exceed derived air concentration (DAC) limits. HEPA filters on hot cell outlets are not credited and will reduce the impact of spills or sprays to the public. Most product solution and uranium solution processing systems will operate at or slightly below atmospheric pressure , or solutions will be pumped between tanks that are at atmospheric pressure to reduce the likelihood of system breach at high pressure. Tanks, vessels , components, and piping are designed for high reliability with materials that will minimize corrosion rates associated with the processed solutions.

13.2.2.9 Mitigated Estimates The controls selected will mitigate both the frequency and consequences of this accident.

The controls selected and described above will prevent a criticality associated with accidental spills and sprays of SNM. The selected IROFS have reduced the off-site consequences to acceptable levels (less than 500 rnrem to the public). Section 13.2.2.7.2 provides the mitigated public dose estimates. Workers will be protected by the selected secondary confinement and shielding IROFSs. Additional detailed information, including worker dose and detailed frequency estimates , will be developed for the Operating License Application. 13.2.3 Target Dissolver Offgas Accidents with Radiological Consequences The offgas accident discussed in Chapter 19.0 , Environmental Report, is a complete release of the iodine (and noble gases) from a loaded dissolver offgas iodine removal unit (IRU). This accident is the loss of efficiency of the IRU due to a process upset (e.g., flooding of the nitrogen oxide [NO x] scrubber) or equipment fai lur e (e.g., loss of the IRU heater) during the dissolution of irradiated targets. The primary components of the dissolver off gas include: NO x scrubbers (caustic and absorbers)

IR Us Pressure-relief vessel Primary adsorbers (carbon media beds for 6 days noble gas holdup) 13-54 NWM I ...... *

NOfITTfWHT MfDtCAL ISOTOP£S NWMl-2013-021 , Rev. 2 Chapter 13.0 -Accident Analysis Iodine guard beds (remove any iodine not trapped in the IRUs) Filters Vacuum receiver tanks Vacuum pumps (draw a downstream vac uum on the target dissolver offgas treatment train) Secondary adsorbers (additional carbon media beds to hold up noble gases for a n additional 60 days) The IR Us nominall y removes about 99.9 percent of the iodine in the off gas stream after the NO x scrubbers.

NWMJ expects the availability and operation ofIRUs will become part of the technical specification to meet annual release limits. The iodine released from dis so lution of the irradiated target s will have three primary pathways:

(1) a fraction of the iodine will stay in the dissolver solution (this iodine is a key dose contributor to liquid spills and sprays accidents

[see Section 13.2.2]), (2) a significant portion of the iodine gas exiting the dissolver will be captured in the caustic scrubber (and other NO x tr ea tment absorbers) and end up in the high do se liquid waste tanks, and (3) the remainder of the iodine will be captured in the IRUs. T hese IR Us wi ll remove the bulk of the radioactive iodine that passes through the dissolver scrubbers during the dissolution process. As demonstrated by the analysis discuss e d in Chapter 19.0, iodine will be the greatest contributor to the effective dose equivalent (EDE) for gaseous accident-related releases from the RPF. The primary and secondary ad sorbers will be important for delaying the release of radioactive noble gases (radioisotopes of Kr and Xe) until these isotopes have had time to decay. However , as shown in the analysis in Chapter 19.0 , the dose impact of noble gases will be orders of magnitude below that of radioiodine. Therefore , this evaluation focuses on accidents or upsets negatively impacting the IRU performance as the bounding offgas accident.

13.2.3.1 Initial Conditions The target dissolver and associated off gas treatment train are assumed to be operational and in service prior to the occurrence of any accident sequence that affects the IR Us. The JR Us are assumed to be loaded with the conservative boundin g holdup in ven tory of iodine, as determined in NWMI-2013-CALC-01 I. No credible event has been identified where the total captured inventory on the IR Us would be released. This accident evaluation is for the release of the iodine generated from a single dissolution of [Proprietary Information].

The maximum amount of iodine [Proprietary Information]

is shown in Table 13-21. The mas s balance projects about 20 percent of the io di ne will stay in the dissolver solution and Table 13-21. Maximum Bounding Inventory of Radio iodine [Proprietary Information]

Isotope 1J 2 m 1 13 3m J Total I Ci = iodine. [Proprietary Information]

[Proprietary Information]

[Proprietar y In form a ti on] [Proprietary lnformati on] [Proprietary Information]

[Proprietary Information]

[Proprietar y Information]

[Proprietary Information]

nearly 50 percent of the e lem ental iodine (h) that does volatize will be captured in the NO x scrubbers (primary the caustic scrubber) and transferred to the high dose liquid waste system. However , for this analysis , all of the iodine is assumed to evolve and remain in the off gas stream going to the IR Us. 13-55

. .-.;; .. NWMI ..... .*.******* * *

NORTHWEST MEDK:Al ISOTIN'ES NWMl-2013-021, Rev. 2 Chapter 13.0 -Accident Analysis Therefore , this evaluation focuse s on accident sequences where the inventory at risk is that generated directl y from the dissolution of [Proprietary Information].

13.2.3.2 Identification of Event Initiating Conditions There are a number of events identified in the PHA that have the potential to impact the normal efficient operation of the target dissolver off gas treatment train. The three most likel y sequences with the potential to impact efficient operation include: (I) excessive moisture carryover in the gas stream due to a process upset in the NO x units , (2) high gas flow rates due to process conditions in the dissolver (e.g., excessive sweep air) or poor NO x recovery , and (3) loss of temperature control (loss of power or failure of temperature controller) to the IRU. All three of these accidents have the potential to reduce the IRU efficiency.

13.2.3.3 Description of Accident Sequences The accident sequence s for loss ofIRU efficienc y include the following. [Proprietary Information]

is being dissolved.

A process upset occurs that reduces the IRU efficiency by an unspecified amount. The event i s identified by the operator either from a process control alarm (e.g., low heater temperature) or a radiation alarm on the gas stream or piping exiting the hot cell. Following procedure , the operator turns the steam off to the dissolver (to slow down the dissolution process).

The operator troubleshoots the upset condition and switches to the back IRU , if warranted , and/or manually open s the valve to the pressure-relief tank in the dissolver off gas system to capture the off gas stream. If the initiator for the event is loss of power or the event creates a condition where vacuum in the dissolver off gas system is lost , the pressure-relief tank valve would automatically open to capture the off gas stream. This tank has been sized to contain the complete gas volume of a dissolution cycle. 13.2.3.4 Function of Components or Barriers The IR Us will be the primary iodine capture devices; however , there will be iodine guard beds downstream of each of the primar y noble gas adsorbers.

The vent system piping will direct the dissolver off gas to the pressure-relief tank or through the guard beds and into the primary process vessel vent system. This system will also have iodine removal beds located downstream of the point where the target dissolver off gas treatment train discharges into the process vessel vent system. Thus , the system will provide a redundant iodine removal capacity that backs up the target dissolver offgas treatment train IR Us. The process vessel vent system will discharge to the Zone I exhaust header , which has a HEGA module that is a defense-in-depth component for this accident sequence.

13.2.3.5 Unmitigated Likelihood Loss of iodine removal efficiency can be initiated by operations or maintenance personnel error or equipment failures.

Failure rates for tanks , vessels , pipes , and pumps are estimated from WSRC-TR-93-262.

Table 13-2 shows qualitative guidelines for applying the likelihood categories. Operator error and equipment failure as initiating events are estimated to have an unmitigated likelihood of "not unlikely." 13-56 NWMl-2013-021, Rev. 2 Chapter 13.0 -Accident Analysis Additional detailed information describing a quantitative evaluation, including assumptions, methodology , uncertainties, and other data , will be developed for the Operating License Application. 13.2.3.6 Radiation Source Term The radioiodine inventory is given in Section 13 .2.3.1. As discussed with regard to the analysis in Chapter 19.0, the dose consequences of noble gas radioisotopes are orders of magnitude less than that of i o dine radioisotopes.

Therefore , the iodine source term is the focus of this accident sequence evaluation.

No credit is taken for any iodine removal in th e dissolver scrubbers or residual iodine remaining in the dissolver solution.

Conversely , in this accident , the previous capture iodine is not part of the source term. Therefore, the source term is 27, 100 Ci. Additional detailed information describing the validation of models , codes, assumptions , and approximations will be developed for the Operating License Application.

The source term for this accident is based on a set of initial conditions th at were designed to bound the credible offgas scenarios. These assumptions include: [Proprietar y Information]

All the iodine in the targets released into the offgas system , and no iodine or noble gases captured in the NO x scrubbers or retained in the dissolver solution Iodine removal efficiency of the dissol ver off gas IRU goes to zero Greater than expected release of material (e.g., no plating out of iodine , or subsequent iodine capture in downstream of unit operations)

The bounding iodine value includes the 1.32 safety factor used in NWMI-2013-CALC-Ol

1. The breakdown of the radionuclide inventory used in NWMI-2013-CALC-Ol l is extracted from NWMI-2013-CALC-006 using the reduced set of 123 radioisotopes.

NWMI-2014-CALC-014 identifie s th e 123 dominant radioisotopes included in the MURR material balance (NWM I-2013-CALC-006).

NWMI-2014-CALC-014 provides the basis for using the 123 radioisotopes from the total list of 6 6 0 radioisotopes potentially present in irradiated target s. The majority of omitted radioisotopes will exist in trace quantities and/or decay swift l y to stable nuclides. The reduced set of 123 radioisotopes consists of those that dominate the radioactivity and decay he at of irradiated targets. 13.2.3. 7 Evaluation of Potential Radiological Consequences R a diological consequences are bounded by those evaluated in the Section 19.4 analysis.

The unmitigated dose consequences should be about 3.4 times less than the results for the public, based on the source term ratio. Realistic radiological consequences are negligible due to the presence of defense-in-depth iodine capabilities in the dissolver offgas system and in the process vessel vent system that backs up the pe rfo rmance of the target dissolver off gas treatment train IR Us. Additional detailed information de s cribing validation of the model s, codes, assumptions , and approximations will be developed for the Operating License Application.

Assuming this accident has similar release characteristics as Section 19.4 , the radiological dose consequences can be estimated using the ratio of source terms. Tills is reasonable since a dissolution takes 1 to 2 hr. The entire in ventory wou ld also be released over a 2-hr period directly to the 22.9 m (75-ft) stack and into the environment.

RSAC 6.2 was used to model the dispersion , and the fo llo wing parameters were used for model runs: Mixing depth: 400 m (1 , 312 ft) (default)

Air densit y: 1,240 g/m 3 (1.24 o zl ft 3) (sea level) Pasquill-Gifford o (NRC Regulatory Guide 1.145) No plume rise (i.e., buoyancy or stack momentum effects) 13-57

- - .. NWMI ..... .*.* .. *.*. *

NOffTlfW'En MEDICAi. lSOTOPfS No p l ume depletion (wet or dry deposition) 2-hr release (constant release of all activity) 2-hr exposure ICRP-30 inhalation model Finite c l oud immersion model Breathing rate: 3.42E-4 m 3/sec (l .2E-2 ft 3/sec) (ICRP-30 heavy activity)

Resp i ratory fraction:

1.0 Table 13-22 shows the distance-dependent total receptor accident doses versus distance from the RPF stack for 2-hr exposure. This table was developed using the results from the Section 19.4 dose consequences and dividing by a ratio of the accident source terms. The maximum pub l ic dose is 6.65 rem at 1 , 100 m. RSAC 6.2 calculates inhalation doses using the ICRP-30 model with Federal Guidance Report No. 11 dose conversion factors (EPA 520/1-88-020 , Limiting Values of Radionuclid e Intak e and Air Con ce ntration and Dose Conv e rsion Fa c tors for Inhalation , NWM l-2013-0 2 1 , R ev. 2 Chapter 1 3.0 -Acci den t A nalysis Tab l e 13-22. Target Dissolver Offgas Accident Total Effective Dose Equivalent TEDE (rem) Distance (m) ' Total 100 2.05E-01 200 l.98E-OI 300 2.21E-01 400 6.41 E-OI 500 l.76E+OO 600 3.18 E+OO 700 4.50E+OO 800 5.47 E+OO 1 , 000 6.50E+OO 1 , 100 6.65E+OO 1,200 6.62E+OO 1 , 300 6.50 E+OO 1,400 6.29E+OO 1 , 500 6.06 E+OO 1,600 5.82E+OO 1,700 2.05 E-OI P e ak total dose is balded a nd italicized.

Submersion , and Ingestion). The committed dose T E DE = total effective dose equiv a lent. equivalent (CDE) is calculated for individual organs and tissues over a 50-y ear period after inhalation.

The CDE for each organ or tissue is multiplied by the appropriate ICRP-26 , Re c omm e ndations of th e Int e rnational Commission on Radiological Prot ec tion, weighting factor and then summed to calcu l ate the committed effective dose equivalent (CEDE). The RSAC 6.2 gamma dose from the cloud is the EDE (the person may or may not be immersed in the cloud depending on the plume position in relation to the ground surface), which is the sum of the products of the dose equivalent to the organ or tissue and the weighting factors applicable to each of the body organs or tissues that is irradiated.

The summation of the two RSAC 6.2 doses is the TEDE, which is the sum of the EDE (for external exposures) and the CEDE (for inhalation exposures). The RSAC 6.2 dose calculations and dose terminology are consistent with 10 CFR 20 terminology based on ICRP-26/30. The doses and dose commitments rem) are within intermediate consequences severity categories ( <25 rem). 13.2.3.8 Identification of Items Relied o n for Safety and Associated Functions IROFS RS-03, Hot Cell Secondary Confinement Boundary The applicable part of IROFS RS-03 that specifically mitigates target dissolver off gas treatment train IRU failures is the process vessel vent iodine removal beds. These beds are located downstream of where the target dissolver offgas treatment train discharges into the process vessel vent system; hence, the beds provide a backup to the target dissolver offgas treatment train IRUs. IROFS RS-03 is categorized as an AEC. 13-58

...... * * !' NORTHWEST MEDICAL ISOTOPES IROFS RS-09, Primary Offgas Relief System NWMl-2013-021 , Rev. 2 Chapter 13.0 -Accident Analysis As an AEC, a relief device will be provided that relieves pres s ure from the system to an on-service receiver tank maintained at vacuum, with the capacity to hold the gases generated by the dissolution of one batch of targets in the target dissolution tank. The safety function of this system is to prevent failure of the primary confinement system by capturing gaseous effluents in a vacuum receiver.

To perform this function, a relief device wi ll relieve into a vacuum receiver that is sized and maintained at a vacuum consistent with containing the capacity of one batch of targets in dissolution. Defensive-in-Depth T h e following defense-in-depth features preventing target dis so lver offgas accidents were identified by the accident evaluations. Releases at the stack will be monitored for radionuclide emissions to ensure that the overall removal efficiency of the system is reducing emissions to design levels and well below regulator limits. A spare dissolver offgas IRU will be available ifthe online IRU unit loses efficiency. The primar y carbon retention bed will include an iodine adsorption stage that reduces iodine as a normal backup to the IRU. 13.2.3.9 Mitigated Estimates The contro l s selected do not affect the frequency of this accident but mitigate the consequences.

The process vessel vent iodine removal bed and the HEGA module in the Zone I exhaust system will mitigate the dose consequences by a factor of 100. The selected IROFS have reduced the off-site consequences to acceptable levels (less than 66 mrem to the public). Additional detailed information, including worker dose estimates and detailed frequency , will be developed for the Operating License Application. 13.2.4 Leaks into Auxiliary Services or Systems with Radiological and Criticality Safety Consequences In the unmitigated scenario, liquid solution leaks into secondary containment (e.g., cooling water jackets) were identified by the PHA to represent a hazard to workers from direct radiological exposure or inhalation and an inhalation exposure hazard to the public. The PHA also identified fissile solution leak s into sec ondary containment as an event that could lead to an accidental nuclear criticality.

The accident s covered by this analysis bound the family of accidents where highly radioactive or fissile solution leaves the hot cell or other shielded areas v ia auxiliar y systems and creates a worker safety or criticality concern. 13.2.4.1 Initial Conditions Initial conditions are described as a tank or vessel (with a heating or cooling jacket) filled with process solution.

Multiple vessels are projected to be at this initial condition throughout the process. The second primar y configuration of concern is the hot cell and target fabrication condensers associated with the four concentrator or evaporator systems. The evaporator(s) initial conditions are normal operations, in which boiling solutions generate an overhead stream that need s to be condensed. The bounding source term is expected to be the dissolvers or the feed tanks in the Mo recovery and purification system. Ta b le 13-23 li sts the radionuclide liquid concentration for [Proprietary Information].

The [Proprietary Information]

stream is used to represent and bound the uranium recovery and recycle and target fabrication evaporators feed streams. 13-59 NWMl-2013-021, Rev. 2 Chapter 13.0 -Accident Analysis Table 13-23. Bounding Radionuclide Liquid Stream Concentrations (4 pages) Unit operation Target dissolution Uranium recovery and recycle [Proprietary Information]

[Proprietary Information]

Dissolver roduct Uranium se aration feed 24 1Am [Proprietary Information]

[Proprietary Information]

I36mBa [Proprietary Information]

[Proprietary Information]

137mBa [Proprietary Information]

[Proprietary Information]

1 39 Ba [Proprietary Information]

[Proprietary Information]

i4o Ba [Proprietary Information]

[Proprietary Information]

141ce [Proprietary Information]

[Proprietary Information]

14 3 Ce [Proprietary lnformation]

[Proprietary Information]

144Ce [Proprietary Information]

[Proprietary Information]

242 cm [Proprietary Information]

[Proprietary Information]

243C m [Proprietary Information]

[Proprietary Information]

2 44Cm [Proprietary Information]

[Proprietary Information]

134Cs [Proprietary Information]

[Proprietary Information]

134m Cs [Proprietary Information]

[Proprietary Information]

136Cs [Proprietary Information]

[Proprietary Information]

137 Cs [Proprietary Information]

[Proprietary Information]

1s sE u [Proprietary Information]

[Proprietary Information]

1 s6 Eu [Proprietary Information]

[Proprietary Information]

1s1Eu [Proprietary Information]

[Proprietary Information]

129 1 [Proprietary Information]

[Proprietary Information]

130J [Proprietary Information]

[Proprietary Information]

13 I J [Proprietary Information]

[Proprietary Information]

1 32 1 [Proprietary Information]

[Proprietary Information]

132m I [Propri etary Information]

[Proprietary Information]

133 1 [Proprietary Information]

[Proprietary Information]

1 33 m I [Proprietary Information]

[Proprietary Information]

1341 [Proprietary Information]

[Proprietary Information]

135 J [Propri etary Information]

[Proprietary Information]

83m Kr [Proprietary Information]

[Proprietary Information]

85 Kr [Proprietary Information]

[Proprietary Information]

85m Kr [Proprietary Information]

[Proprietary Information]

87 Kr [Proprietar y Information]

[Proprietary Information]

88Kr [Proprietary Information]

[Proprietary Information]

140La [Proprietary Information]

[Proprietary Information]

141La [Proprietary Information]

[Proprietary Information]

1 42 La [Proprietar y Information]

[Proprietary Information]

99 Mo [Proprietary Information]

[Proprietary Informat i on] 95 Nb [Proprietary Information]

[Proprietary Information]

95mNb [Proprietary Information]

[Proprietary Information]

96 Nb [Proprietar y Information]

[Proprietary Information]

13-6 0 NWMl-2013-021 , Rev. 2 Chapter 1 3.0 -Acci d ent Analysis Table 13-23. Bounding Radionuclide Liquid Stream Concentrations (4 pages) Unit operation Target dissolution Uranium recovery and recycle [Proprietar y Information]

[Proprietary Information]

Dissol ve r roduct Uranium se aration feed 97 Nb [Proprietary Information]

[Proprietary Information]

97mNb [Proprietary Information]

[Proprietar y Information]

14 1 Nd [Proprietary Information]

[Proprietary Information]

236m Np [Proprietar y Information]

[P roprietar y Information]

23 1 Np [Proprietary Information]

[Proprietary Information]

23s Np [Proprietary Information]

[Proprietar y Information]

23 9Np [Proprietary Information]

[Proprietary Information]

233 Pa [Proprietar y Information]

[Proprietary Information]

234 pa [Proprietary Information]

[Proprietary Information]

234 m Pa [Proprietar y Information]

[Proprietar y Information]

11 2 pd [Proprietary Information]

[Proprietary Information]

1 4 1 Pm [Proprietar y Information]

[Propri e tary In formation]

14 8 Pm [Proprietary Information]

[Proprietary Information]

I4 8 mpm [Proprietar y Information]

[Proprietary Information]

149Pm [Proprietary Information]

[Proprietary Information]

ISOpm [Proprietar y Information]

[Proprietary Information]

ISIPm [Proprietary Information]

[Proprietary Information]

14 2 Pr [Propriet ary Informati o n] [Proprietary Information]

14 3 Pr [Proprietary Information]

[Proprietary Information]

I44 pr [Propri etary Information]

[Proprietary Information]

144mpr [Proprietary Information]

[Proprietary Information]

I4 S pr [Proprietary In fo rmation] [Proprietary Information]

2Js pu [Proprietary Information]

[Proprietary Information]

239 Pu [Proprietar y Information]

[Proprietary Information]

240 pu [Proprietary Information]

[Prop r ietary Information]

24 1Pu [Proprietar y Informati on] [Propri etary Information]

10 3 mRh [Proprietary Information]

[Proprietary Information]

I OS Rh [Proprietar y Informati o n] [Proprietary Information]

10 6 Rh [Proprietary Information]

[Proprietary Information]

J06mRh [Proprietar y Information]

[Proprietary Information]

10 3 Ru [Proprietary Information]

[Proprietary Information]

1o s Ru [Proprietary Information]

[Proprietary Information]

106Ru [Proprietary Information]

[Prop r ietary Information]

122 sb [Proprietar y Information]

[Proprietary Information]

124 Sb [Proprietary Information]

[Proprietary Information]

125 Sb [Proprietar y Information]

[Proprietary Information]

1 26 Sb [Proprietary Information]

[Proprietary Information]

127 Sb [Proprietar y Information]

[Proprietary Information]

128 Sb [Proprietary Information]

[Proprietary Information]

13-61 N W Ml-201 3-0 21, Rev. 2 Chapter 13.0 -Acci dent An alysis Tab l e 13-23. Boundi n g Ra d ionuclide Liq u id Stream Concentrations (4 pages) Unit operation Target dissolution Uranium recovery and recycle [Proprietary Information]

Dissolver roduct 1 2 smsb [Proprietary Information]

[Proprietary Information]

129Sb [Proprietary Information]

[Proprietary Information]

1 s 1sm [Proprietary Information]

[Proprietary Information]

1s3sm [Proprietary Information]

[Proprietary Information]

1 s6 sm [Proprietary Information]

[Proprietary Information]

89Sr [Proprietary Information]

[Proprietary Information]

9 osr [Proprietary Information]

[Proprietary Information]

91sr [Proprietary Information]

[Proprietary Information]

92 Sr [Proprietary Information]

[Proprietary Information]

9 9Tc [Proprietary Information]

[Proprietary Information]

99 mTc [Proprietary Information]

[Proprietary Information]

12smTe [Proprietary Information]

[Proprietary Information]

121 Te [Proprietary Information]

[Proprietary Information]

1 2 1mTe [Proprietary Information]

[Proprietary Information]

1 29 Te [Proprietary Information]

[Proprietary Information]

129mTe [Proprietary Information]

[Proprietary Information]

1 3 1Te [Proprietary Information]

[Proprietary Information]

1 3 1mTe [Proprietary Information]

[Proprietary Information]

1 32 Te [Proprietary Information]

[Proprietary Information]

1 33 Te [Proprietary Information]

[Proprietary Information]

133m Te [Proprietary Information]

[Proprietary Information]

1 3 4Te [Proprietary Information]

[Proprietary Information]

23 1Th [Proprietary Information]

[Proprietary Information]

2 3 4Th [Proprietary Information]

[Proprietary Information]

232 u [Proprietary Information]

[Proprietary Information]

2 3 4U [Proprietary Information]

[Proprietary Information]

23s u [Proprietary Information]

[Proprietary Information]

23 6u [Prop r ietary In formation]

[Proprietary Information]

231 u [Proprietary Information]

[Proprietary Information]

mu [Proprietary Information]

[Proprietary Information]

1 J 1mxe [Proprietary Information]

[Proprietary Information]

133 Xe [Proprietary Information]

[Proprietary Information]

1 JJ mxe [Proprietary Information]

[Proprietary Information]

135 Xe [Proprietary Information]

[Proprietary Information]

1 Js mxe [Proprietary Information]

[Proprietary Information]

89my [Proprietary Information]

[Proprietary Information]

90 y [Proprietary Information]

[Proprietary Information]

90my [Proprietary Information]

[Proprietary Information]

9J y [Proprietary Information]

[Proprietary Information]

1 3-62 NWMl-2013-021 , Rev. 2 Chapter 13.0 -Accide n t Analysis Table 13-23. Bounding Radionuclide Liquid Stream Concentrations (4 pages) Unit operation Target dissolution Uranium recovery and recycle [Proprietar y Information]

[Proprietary Information]

Dissolver roduct Uranium se aration feed 9Imy [Proprietary Information]

[Proprietary Information]

92 y [Proprietary Information]

[Proprietary Information]

93y [Proprietary Information]

[Proprietary Information]

9 3 zr [Proprietary Information]

[Proprietary Information]

9szr [Proprietary Information]

[Proprietary Information]

97 Zr [Proprietary Information]

[Proprietary Information]

Totals [Proprietary Information]

[Proprietary Information]

Source: Table 2-1 ofNWM l-2013-CALC-0 11 , Sourc e T e rm C a l c ulation s, Rev. A, Northwe s t Medical Isotopes , LLC , Corvallis , Oregon , February 2015. EOI = end of irradiation.

In each case, a jacketed vessel is assumed to be filled with process so lution appropriate to the process location , with the process offgas venti lation system operating.

A level monitoring system will be available to monitor tank transfers and stagnant store vo lume s on all tanks processing LEU or fission product so lution s. The source term used in this analysis is from NWMI-2013-CALC-Ol

1. The breakdown of the radionuclide inventory used in NWMI-2013-CALC-Ol 1 is extracted from NWMl-2013-CALC-006 using the reduced set of 123 radioisotopes.

NWMI-2014-CALC-014 identifies the 1 23 dominant radioisotopes included in the MURR material balance (NWMI-2013-CALC-006).

NWMI-2014-CALC-014 provides the basis for usin g the 123 radioisotopes from the total list of 660 radioisotopes potentially present in irradiated targets. The majority of omitted radioisotopes exist in trace quantities and/or decay swiftly to stable nuclides.

The reduced set of 123 radioisotopes consists of those that dominate the radioactivity and decay heat of irradiated targets. 13.2.4.2 Identification of Event Initiating Conditions The accident initiating event is genera lly described as a process equipment failure. The PHA identified simi l ar accident sequences in four nodes associated with l eaks of enric h ed uranium so lution into heatin g and/or cooling coils surro undin g safe-geometr y tanks or vessels. The PHA identified predominately corrosive degradation of the tank or overpressure of the tank as potential causes that might damage this inter face and allow enriched uranium so lution to leak into the cooling system media or into the steam condensate for the heating system. The primary containment fails , whic h allows radioactive or fissi le so lution s to enter an aux iliar y system. Radioactive or fissile so lution l eaks across the mechanical boundary bet ween a process vesse l and associated heatin g/coo lin g jacket into the heatin g/coo lin g media. Where heating/cooling jackets or heat exchangers are used to heat or cool a fissile and/or high-do se process solution, the potential exists for the ba1Tier between the two to fail and a llow fissile and/or hi gh-dose process solutio n to enter the auxiliary system. If the auxiliary system is not designed with a safe-geometry configuration , or if this system exits the hot cell containment, confinement, or shielding boundary in an uncontrolled manner , either an accidental criticality is possible or a high-dose to workers or the public can occur. 13-63 NWMl-2013-021, Rev. 2 Chapter 13.0 -Accident Analysis Where auxiliary services enter process solution tanks , the potential exists for backflow of high-dose radiological and/or fissile process solution into the auxiliary service systems (purge air , chemical addition line , water addition line, etc.). Since these systems are not designed for process solutions, this event can lead to either accidental nuclear criticality or to high-dose radioactive exposures to workers occupying areas outside the hot cell confinement boundary.

13.2.4.3 Description of Accident Sequences The PHA made no assumption about the geometry or the extent of the heating/cooling subsystem. Consequently , an assumption is made that without additional control, a credible accidental nuclear criticality could occur , as the fissile solution enters into the heating/cooling system not designed for fissile solution , or as the solution exits the shielded area and creates a high worker dose consequence.

If the system is not a closed loop, a direct release to the atmosphere can a l so occur. Either of these potential outcomes can exceed the performance criteria of one process upset , leading to accidental nuclear criticalit y or a release that exceeds intermediate or high consequence levels for dose to workers, the public, or environment.

The accident sequence for a tank leak into the cooling water (or heating) sy s tem includes the following. The process vessel wall fails and the tank contents leak into the cooling jacket and medium , or the process medium leaks into the vessel. Tank liquid level monitoring and liquid level instrumentation are functional; however, depending on the size of the leak, the tank level instrumentation may or may not detect that a tank has leaked. The cooling water system monitor (conductivity or pH) detects a change in the cooling water , and an alarm notifies the operator.

The operator places the system in a safe configuration and troubleshoots the source of the leak. Maintenance activities to identify, repair , or replace the cause of the leak are initiated after achieving the final stable condition. Additional PHA accident sequences include the backflow (siphon) or backup of process solutions into the chemical or water addition systems. The controls for these accidents are described in Section 13.2.4.8. 13.2.4.4 Function of Components or Barriers This accident sequence requires the failure of the primary confinement in a safe-geometr y vessel or tank , the normal condition criticality safety control for the process. This same barrier will provide primary containment of the high-dose process solution to maintain the solution within the hot cell containment , confinement , and shielding boundary.

The heating and cooling systems will have secondary loops (closed loops), so a second failure is required for the fissile solution to enter into a non-geometric-safe auxiliary system or into a non-shielded auxiliary system out of the hot cells. 13.2.4.5 Unmitigated Likelihood Leaks into auxiliary services can be initiated by mechanical failure of equipment boundaries between the process so lution s and auxiliary system fluids , or backflow of high-dose radiological or fissile solution to a chemical supply system. Failure rates for tanks, vessels, pipes, and pumps are estimated from WSRC-TR-93-262.

Table 13-2 shows qualitative guidelines for applying the likelihood categories. Failures resulting in leaks or backflows as initiating events are estimated to have an unmitigated likelihood of "not unlikely." 13-64 NWM I ..*... * * ! NOfllTHWEn MEDCAl. ISOTOPfS NWMl-2013-021, Rev. 2 Chapter 13.0 -Accident Analysis Additional detailed information describing a quantitative evaluation , including assumptions , methodology, uncertainties, and other data, will be developed for the Operating License Application.

13.2.4.6 Radiation Source Term The following source term descriptions are based on information developed for the Construction Permit Application.

Additional detailed information describing source terms will be developed for the Operating License Application. Source terms associated with leaks and backflows into auxiliary system are dependent on vessel location i n the process system. The high-dose uranium solution source term bounds this analysis.

Solution leaks i nto the cooling or heating system were bounded by the irradiated target dissolver after dissolution is complete. The target dissolution process produces an aqueous solution containing uran y l nitrate , nitric acid, and fission products.

The fission product inventory is bounded by dissolution of a batch of MURR targets that is decayed [Proprietary Information], with an equivalent uranium concentration of 283 g U/L. T he primary attribute of high-dose uranium solutions used for consideration of direct exposure consequences is that equipment operation and maintenance must be conducted in a shielded hot cell environment due to the presence of fission products.

13.2.4.7 Evaluation of Potential Radiological Consequences The following evaluations are based on information developed for the Construction Permit Application.

Additional detailed information describing radiological consequences will be developed for the Operating License Application.

1 3.2.4.7.1 Direct Exposure Consequences The potential radiological exposure hazard of liquid spills discussed in Section 13.2.2 bound the consequences from radiation exposure for these accident sequences.

Even the low-dose uranium solutions , while generally contact-handled , ha v e similar e x posure consequences due to the criticalit y hazard. Auxiliary systems located within hot cells will require shielding to control worker radiation exposure independent of whether process solution is contained in the vessel or leaked into the auxiliary system. Thus , in a very short period of time , a worker can receive a significant intermediate or high consequence dose rate. B a sed on the analysis of several accidental nuclear criticalities in industry , LA-13638 identifies that a uranium solution criticality can yield between 10 1 6 to 10 1 7 fissions.

Dose rates for anyone in the target fabrication area can have high consequences.

Consequences for a shielded hot cell criticality will be developed for the Operating License Application.

13.2.4. 7.2 Confinement Release Consequences Not applicable to this accident sequence.

13.2.4.8 Identification of Items Relied on for Safety and Associated Functions Hot cell shielding is designed to provide protection from leaks into the heating and cooling closed loop auxiliary systems that result in redistribution of high-dose uranium solutions in the hot cell. From a direct exposure perspective , this type of accident does not represent a failure or adverse challenge to the hot cell shielding boundary function. 13-65

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NOllTHWUT MEDICAL tSOTOPES 13.2.4.8.1 IROFS RS-04, Hot Cell Shielding Boundary NWMl-2013-021, Rev. 2 Chapter 13.0 -Accident Analysis IROFS RS-04 functions to prevent worker dose rates from exceeding exposure criteria due to the presence ofradioactive materials in the hot cell vessels before or after a leak to the cooling and heating auxiliary syste ms. As a PEC and safety feature , the hot cell shielding boundary will provide an integrated system of features that protect workers from the high-dose radiation generated during radioisotope processing. A primary safety function of the hot cell shielding boundary will be to reduce the radiation dose at the worker/hot cell interface to ALARA. While protecting workers, the shield will also protect the public at the controlled area boundary.

The hot cell shielding boundary will provide shielding for workers and the public during normal operations to reduce worker exposure to an average of 0.5 mrem/hr , or less , in normally accessible workstations and occupied areas outside of the hot cell. 1 The hot cell shielding boundary will also provide s hielding for workers and the public during proces s upsets to reduce worker exposure to a TEDE of 5 rem , or less , at workstations and occupied areas outside of the hot cell. 2 As a PEC, shielding wi ll be provided by a thick concrete, steel-reinforced wall with steel cladding that reduces the normally expected operational exposures from within the boundar y to an average of 0.5 mrem/hr , or less , outside of the boundary.

Where direct visual access is required, leaded-glass windows with appropriate thicknesses will be used to reduce normally expected operational exposures from within the boundary to an average of 0.5 mrem/hr , or less , outside of the boundary.

Some shielding will be movable, such as around the high-dose waste cask loading area. Where penetrations are required , the engineered design provides for access-controlled, non-occupied corridors or airlocks where potential radiation streaming is safely mitigated by multiple layers of shielding or through a torturous path. The shielding is also designed to reduc e the exposure from postulated upsets within the hot cell shielding boundary to l ess than a low consequence exposure to workers and the public of 5 rem , or less, per incident.

These incidents include spills, sprays, fires, and other releases of radionuclides contained within the boundary. The shield may be divided into protection areas for the purposes of applying limiting conditions of operation.

Each shielded protected area will be operab l e when the equipment in that area is in the operating or standby mode s. 13.2.4.8.2 IROFS CS-06, Pencil Tank and Vessel Spacing Control using the Diameter of the Tanks, Vesse ls , or Piping All tanks , vessels, or piping systems invo l ved in a process upset will be controlled with a safe-geometry confinement IROFS that consists of IROFS CS-06 to provide a diameter of the vessels confinement or IROFS CS-26 to provide safe volume confinement.

13.2.4.8.3 IROFS CS-10, Closed Safe Geometry Heating or Cooling Loop with Monitoring and Alarm As a PEC , a closed-loop safe-geometry heating or coo lin g loop with monitoring for uranium proces s so lution or high-do se process solution will be pro v ided to safely contain fissile process solution that leaks across this boundary , if the primary boundary fails. The dual-purpo se safety function of this closed loop is to prevent fissile process solution from causing accidental nuclear criticality and to prevent high-do se process solution from exiting the hot cell containment, confinement , or shielded boundary (or, for systems located outside of the hot cell containment , confinement , or shielded boundary , to prevent low-dose solution from exiting the facility), causing excessive dose to workers and the public, and/or release to the environment.

1 So me operations may have high er doses durin g short per i ods of th e operat ion. The average worker expos ur e rate is d es i gned to be 0.5 mrem/hr , or le ss. Areas not norm a ll y accessible b y the worker ma y hav e higher dose rates (e.g., streaming radiation around norm a ll y inaccessible r emote manipulator p e netration s well above th e worker's he a d). 2 The shielding i s not c r e dited for mitig at ing do se rates durin g an acci d e ntal nucl ear criticality; inste ad, a ddition a l IRO FS are identified to provide double-contingency protection to prevent (reduc e the likelihood of) a n accidental nuclear criticality. 13-66 NWMl-2013-021 , Rev. 2 Chapter 13.0 -Accident Analysis The heat exchanger materials will be compatible with the harsh chemical environment of the tank or vessel process (this may vary from application to application).

Sampling of the h eating or cooling media (e.g., steam condensate conductivity, cooling water radiological activity , uranium concentration, etc.) wi ll b e conducte d to alert the operator that a brea ch has occurred a nd that additional corrective actions are required to identi fy and isolate the failed component and restore the closed loop inte grity. Dischar ged solutions from this syste m wi ll be handled as potentially fissile an d sampled according to IROFS CS-16 and CS-17 prior to discharge to a non-safe geometry.

13.2.4.8.4 IROFS CS-27, Closed Heating or Cooling Loop with Monitoring and Alarm As a PEC, on the evaporator or concentrator condensers , a close d cooling loop with monitoring for b r eakthro ugh of process so luti on will be provided to contain process solution that leaks across this b ou nd a r y, if the boundary fails. IROFS CS-27 is a pplied to those high-h eat capacity coo ling jacket s (requiring very lar ge loop heat exc hanger s) servic in g condensers where the leakage i s always from the cooling loop to the con d enser, reducing back-l eakage, and the ri sk of product so luti ons en terin g th e condenser is very low by evaporator or concentrator design. The purpose of this safety function is to monitor the condition of the condenser coo lin g jacket to ensure that in the unlik e l y event that a condense r overflow occurs , fissile and/or high-dose process so lution wi ll not flow into this non-safe geometry cooling loop and cause nuclear criticality.

The closed loop will also i solate any hi g h-d ose fissile product so lid s (from the same event) from penetrating the hot cell s hieldin g boundary, and any high-dose fission gases from penetrating the hot cell shield in g boundar y durin g normal operations.

The heat exc han ger materials will be compatible wit h the har sh che mical enviro nment of the t ank or vesse l pro cess (t hi s ma y vary from application to application).

Samp lin g of the coo lin g medi a (e.g., coo ling water radiological activity, uranium concentration, etc.) will be conducted to alert th e operator that a breach h as occurred and that additional corrective actions are required to identify and i so lat e the failed component and to restore the c lo se d loop integrity.

Closed loop pre ssure will also be monitored to id entify a leak from the closed loop to the process system. Discharged solutions from this system wi ll be handled as potentially fissile and samp l ed accord in g to IROFS CS-16 and CS-17 prior to discharge to a non-safe geo m etry. 13.2.4.8.5 IROFS CS-20, Evaporator or Concentrator Condensate Monitoring As an AEC, th e condensa t e t anks w ill use a continuously active uranium detection system to detect high carryover of uranium that shuts down the evaporator feeding the tank. The purpose of this sys tem is to (1) detect an anomaly in the evaporator or concentrator indi cating high uranium content in the condenser (due to flooding or excess i ve foaming), and (2) prevent high conce ntrat ion uranium solution from being available in the condensate tank for discharge to a non-favorable geometry system or in the condenser for leakin g to the non-safe geometry coo ling loop. The safety function of this IROFS is to prevent an accidental n uclear critica li ty. The detection system will work by continuously monitoring condensate uranium content and detecting high uranium concentration, and then shutting down the evaporator to isol ate the condensate from the condenser and condensate tank. At a limitin g se tpoint , the uranium monitor-detecting devi ce will close an isolation va lve in the inlet to the eva porator (or otherwise secure the evaporator) to stop the discharge of high-uranium content solution into the condenser and condensate collection tank. The uranium monitor is de signed to produce a va l ve-open permissive signa l that fails to an open state , closing the valve on loss of electrical power. The isolation va l ve is de s igned to fail-closed on loss of instrument air , and the so lenoid is designed to fai l-clo se d on lo ss of s ign al. The locati ons w here thi s IROFS is used will be d e t ermined during final design. 13-67

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*---------* *-13.2.4.8.6 IROFS CS-18, Backflow Prevention Device NWMl-20 13-021 , Rev. 2 Chapter 13.0 -Accident Analysis As a PEC or AEC, chemical and gas addition ports to fissile process solution systems will enter through a back:flow prevention device. This device may be an anti-siphon break, an overloop seal, or other active engineering feature that addresses the conditions of backflow and prevents fissile solution from entering non-safe geometry systems or high-dose solutions from exiting the hot cell shielding boundary in an uncontrolled manner. The safety function of this IROFS is to prevent fissile solutions and/or high-dose so luti ons from backflowing from the tank int o systems that are not designed for fissile solutions that cou ld lead to accidental nuclear criticality or to l ocations outside of the hot cell shielding boundaries that might lead to high exposures to the worker. Each hazardous location will be provided an engineered back:flow prevention device that provides high reliabi lit y and avai l ability for that location.

The backflow prevention device features for high-dose product so lution s will be l ocated inside the hot cell shielding and confinement boundaries ofIROFS RS-04 and RS-01 , respectively.

The feature i s designed such that spills from overflow are directed to a safe geometry confinement berm contro ll ed by IROFS CS-08 (described in NWMI-2015-SAFETY-004, Quantitative Risk Ana lysis of Process Upsets Associated with Passive Engineering Controls Leading to Criticality Accident Sequences , Section 3.1.6.3) or to safe-geometry tanks controlled by IROFS CS-11. 13.2.4.8.7 IROFS CS-19 , Safe Geometry Day Tanks As a PEC , safe-geometry day tanks will be provided where the first barrier cannot be a back:flow prevention device. The safety function of this PEC i s to prevent accidental nuclear criticality by providing a safe-geometry tank if a fissile solution backs-up into an auxi li ary chem i ca l addition system. IROFS CS-19 will be used where conventional backflow prevention in pre ss urized systems is not reliable.

The safe-geometry day tank will be provided for those chemical addition activities where the reagent cannot be added via an anti-siphon break since the tank or vessel is not vented and operates under some backpressure conditions.

The feature works by providing a safe-geometry vesse l that is filled with chemical reagent using the conventional backflow prevention devices , and then provide s a pump to add the reagents to the respective process system under pressure.

Safe-geometry day tanks servicing dose product solutions systems will be located in the hot cell shie ldin g or confinement boundaries of IROFS RS-04 and RS-01 , respectively.

Defensive-in-Depth The following defense-in-depth features preventing leaks into auxiliary services or systems were identified by the accident evaluations.

All tanks will be vented and unpressurized under normal use. The heating and coo lin g systems will operate at pressures that are higher than the processing systems that they heat or cool. The majority of system leakage would typically be in the direction of the heat transfer media to the processing system. All vented tanks are designed with level indicators that are available to the operator to detect the level of solution in a tank remote l y. Operating procedures will identify an operational high-level fill operating limit for each tank. As part of the l evel detector, a high-level audible alarm and light will be prov id ed to indicate a high level above this operating limit so that the operator can take action to correct conditions leading to fai lur e of the operating limit. With batch-type operation with typically low volume transfers , the sizing of the tanks wi ll include sufficient overcapacity to handle reasonable perturbations in operations caused by variations in chemical concentrations and operator errors (adding too much). 13-68

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NCMO'HWHT MEOfC.Al ISOTOPU NWMl-2013-021, Rev. 2 Chapter 13.0 -Accident Analysis Tank and vesse l walls will be made of corrosion-resistant materials and have wall thicknesses that are rated for long service with harsh acid or basic chemicals.

Purge and gas reagent addition lines (air , nitrogen , and oxygen) will be equipped with check valves to prevent flow of process solutions back into uncontrolled geometry portions (tanks , receivers , dryers , etc.) of the delivery system. 13.2.4.9 Mitigated Estimates The controls selected wil l mitigate both the frequency and consequences of this accident.

The controls se l ected and described above will prevent a criticality associated with SNM leaks into auxiliary systems. The selected IROFS have reduced the potential worker safety consequences to acceptable levels. Additiona l detailed information , including worker dose and detailed frequency estimates , will be developed for the Operating License Application. 13.2.5 Loss of Power 13.2.5.1 Initial Conditions Initial conditions of the event are described b y normal operation of all process systems and equipment.

13.2.5.2 Identification of Event Initiating Conditions Multiple initiating events were identified by the PHA that could result in the loss of normal electric power. 13.2.5.3 Description of Accident Sequences The loss of power event sequence includes the following.

1. Electrical power to the RPF is lost due to an initiating event. 2. The uninterruptible power supply automatically provides power to systems that support safety functions , protecting RPF personnel and the public. The following systems ar e supported with an uninterruptible power supply: Process and facility monitoring and control systems Facility communication and security systems Emergency I ighting Fire alarms Criticality accident alarm systems Radiation protection systems 3. Upon loss of power , the following actions occur: Inlet bubble-tight isolation dampers within the Zone I ventilation system close , and the heating , ventilation , and air conditioning (HY AC) system is automatically placed into the passive venti lation mode of operation. Process vessel vent system is automatically placed into the passive ventilation mode of operation , and all electrica l heaters cease operation as part of the passive operation mode. Pressure-relief confinement system for the target dissolver off gas system is activated on reaching the system relief setpoint , and dissolver offgas is confined in the offgas piping , vessels , and pressure-relief tank (IROFS RS-09). 13-69

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NOflTtfWUTMEDtCALISOTOPES NWMl-2013-021, Rev. 2 Chapter 13.0 -Accident Analysis Process vessel emergency purge system is activated for hydrogen concentration control in tank vapor spaces (IROFS FS-03). Uranium concentrator condensate transfer line valves are automatically configured to return condensate to the feed tank due to residual heating or cooling potential for transfer of process fluids to waste tanks (IROFS CS-l 4/CS-15). All equipment providing a motive force for process activities cease, including:

Pumps performing liquid transfer s of process solutions Pumps supporting operation of the steam and cooling utility heat transfer fluids Equipment supporting physical transfer of items (primarily cranes) 4. Operators follow alarm response procedures.

5. The facility is now in a stable condition.

13.2.5.4 Function of Components or Barriers All facility structural components of the hot cell seco ndary confinement boundary (in a passive ventilation mode) and hot cell shielding boundary (walls, floors, and ceilings) will remain intact and functional.

The engineered safety features requiring power will activate or go to their fail-safe configuration.

13.2.5.5 Unmitigated Likelihood Loss of power can be initiated by off-site events or mechanical failures of equipment.

Failures resulting in loss of power as initiating events are estimated to have an unmitigated likelihood of "not unlikely." Additional detailed information describing a quantitative evaluation, including assumptions , methodology , uncertainties , and other data , will be developed for the Operating License Application.

13.2.5.6 Radiation Source Term The loss of power evaluation is ba sed on information developed for the Construction Permit Application.

Detailed information describing radiation source terms for the loss of power event will be developed for the Operating License Application.

13.2.5.7 Evaluation of Potential Radiological Consequences The loss of power evaluation is based on information developed for the Construction Permit Application.

A detailed evaluation of potential radiological consequences, including a summary of radiological consequences from the analysis of other accidents where loss of power was an initiator , will be provided in the Operating License Application.

13.2.5.8 Identification of Items Relied on for Safety and Associated Functions No additional IROFS have been identified specific to this event other than maintain operability of the IROFS listed in Section 13.2.5.3. The loss of normal electric power will not result in unsafe conditions for either workers or the public in uncontrolled areas. Defensive-in-Depth The following defense-in-depth feature, minimizing the impact of a loss of power event , was identified by the accident evaluations.

A standby diesel generator will be available at the RPF. 13-70 13.2.6 Natural Phenomena Events NWMl-2013-021, Rev. 2 Chapter 13.0 -Accident Analysis C hapter 2.0, "Site Characteristics,'

' and Chapter 3.0 discuss the design of SSCs to withstand external events. The RPF is designed to withstand the effects of natural phenomena events. Consequences of natural phenomena accident se quenc es have been evaluated. Sections 13.2.6.l through 13.2.6.6 provide event descriptions and identify any additional controls required to protect the health and safety of workers, the public, and environment.

13.2.6.1 Tornado Impact on Facility and Structures, Systems, and Components The adverse impact of a tornado on facility operations has a number of facets that must be evaluated.

This evaluation addresses the facility design as impacted by the maximum-sized tornado with a return frequency of 10-5 /yea r (yr).

High winds can lead to significant damage to the facility structure.

Damage to the structure is a function of the strength of the tornado winds, duration , debris carried by the winds, direction of impact , and facility design. This evaluation determines the impact of tornado winds on the facility from a design basis perspective to ensure that the design prevent s impact to SS Cs in the building. The local area impact may result in loss of utilities (e.g., electrical power) and reduced access b y local emergency responders. Loss of power is evaluated (Section 13.2.5) as a potential cause for all adverse events. The individual PHA nodes evaluate the loss of site power and loss of power distribution to each subsystem. High winds may directl y impact SSCs important to safety (e.g., co mponents of the fire protection system are located in areas adjacent to the building) and reduce the reliability of those SS Cs to respond to additional events (like los s of electrical power) that can be initiated concurrently with the tornado (either as an indirect result or as an additional random failure).

This evaluation ana l yzes the impact of tornado winds on these SSCs. Tornado impact on the facility structure

-High wind pre ss ures could cause a partial or complete collapse of the facility structure , which ma y cause damage to SS Cs important to safety or impact the availability and reliability of those SSCs. A partial or complete structural collapse ma y also lead dir ect ly to a radiological or chemical release or a pot e ntial nuclear criticality , if damage caused by the collap se creates a v iolation of criticality spacing requirements.

Tornado wind-driven missiles could penetrate the fa c ilit y building envelope (walls a nd roof), impacting the availabi lit y and reliability of SSCs important to safety , or may lead directly to a radiological or chemical release. Tornado impact on SSCs important to safety located outside the main facility -High wind pres sures and tornado wind-driven missiles could damage SSCs important to safety located outside the RPF building envelope.

The damage sustained ma y impact the availability and reliability of the SSCs important to safety. Loss of site power ma y affect the ability of SSCs important to safety located within the facility building envelope to respond to additional events. A pa rtial or complete collapse of the facility structure could also lead directl y to an accident with adverse intermediate or high consequences.

The on l y IROFS loc ated outside the RPF building envelope is the exhaust stack. Buckling or toppling of the exhaust stack would affect its ability and availabi lit y to mitigate other events with intermediate consequences.

The return frequenc y of the design basis tornado is 10-5/yr , making the initiating event highly unlikel y. No additional IROFS are required.

13-71 NWMl-2013-021, Rev. 2 Chapter 13.0 -Accident Analysis 13.2.6.2 High Straight-Line Winds Impact the Facility and Structures, Systems, and Components Similar to the tornado, high straight-line winds can also damage the facility structure, which in tum can lead to damage to SSCs relied on for s afety. This evaluation demonstrates how the facilit y design addressed straight-line winds with a return interval of 100 years or more, as required by building codes. Buckling or toppling of the exhaust stack would affect the ability and availability to mitigate other events with intermediate consequences.

A partial or complete collapse of the facility structure may also lead directly to an accident with adverse intermediate or high consequences. The facility is designed as a Risk Category IV structure, a standard industrial facility with equivalent chemical hazards , in accordance with American Society of Civil Engineers (ASCE) 7 , Minimum D e sign Loads for Buildings and Other Stru c tur e s. The return frequency of the basic (design) wind speed for Risk Category IV structures is 5.88 x l0 4/yr (mean return interval , MRI= 1 , 700 yr). At this return frequency , the straight-line wind event is not likely but credible during the design life of the facility and is considered in the structural design as a severe weather event. The provisions of the ASCE 7 standard, when used with companion standards such as American Concrete Institute (ACI) 318, Building Cod e R e quirements for Structural Concr e t e, and American Institute of Steel Construction (AISC) 360 , Specification for Structural Steel Buildings , are written to achieve the target maximum annual probabilities of established in ASCE 7. The highest maximum probability of failure , which is for a failure that is not sudden and does not lead to a wide-spread progression of collapse , targeted for Risk Category IV structures is 5.0 x 10-6. Therefore, the likelihood of failure of the structure when subjected to the design basis straight-line wind in conjunction with other loads, as required by ASCE 7 , is highly unlikely.

No additional IROFS are required. 13.2.6.3 Heavy Rain Impact on Facility and Structures, Systems, and Components Localized heavy rain can overwhelm the structural integrity of the RPF roofing system. This evaluation determines the impact of probable maximum precipitation (PMP) in the form of rain on the roofing structure.

The PMP is defined as " theoretical greatest depth of precipitation for a given duration that is physically possible over a particular drainage area at a certain time of year." In other words , the PMP represents the theoretical worst-case of the most precipitation the atmosphere is capable of discharging to a particular area over a selected period of time. The PMP is based on an empirica l methodology with no defined annua l exceedance probability.

For impact on the facility, the PMP for 25.9 square kilometers (km 2) (10 square miles [mi 2]) is eva lu ated. Large amounts of rain water accumulating on the roof could lead to collapse of the roof. A partial or complete co llap se of the faci lit y roof may impact the availability or reliability of SS Cs relied on for safety located within the RPF building envelope to respond to other events of high consequence.

From the Nationa l Weather Service (NWS)/National Oceanic and Atmospheric Administration (NOAA) Hydrometeorological Report No. 51 , Probabl e Maximum Pr e cipitation Estimat e s , Unit e d States East of th e 105th Meridian, the PMP is defined as "theoretical greatest depth of precipitation for a given duration that is physically possible over a particular drainage area at a certain time of year." In other words, the PMP represents the theoretical worst-case of the most precipitation the atmosphere is capable of discharging to a particular area over a selected period of time. The PMP is based on an empirica l methodology with no defined annual exceedance probability.

Although the NWS/NOAA has historically stated that it is not possible to assign an exceedance probability to the PMP (NOAA Technical Report NWS 25, Comparison of Generali ze d Estimate s of Probabl e Maximum Pr e cipitation with Greatest Observed Rainfalls), several academic studies and papers have undertaken the exercise to determine the annual exceedance probability for PMP using modern probabilistic techniques and storm modeling and have found that the exceedance probability varies by location but is quite low (NAP , 1994). As such , the PMP event has been determined to be highly unlikely. 13-72 NWMI ...... *

NORTMWUT MEOtC.Al ISOTOl'£S No additional IROFS are required. NWMl-2013-021, Rev. 2 Chapter 13.0 -Accident Analysis The roof of the RPF is designed to prevent rainwater from accumulating on the roof. In accordance with 2012 International Building Code (IBC) and ASCE 7, the roof structure is designed to safe ly support the weight of rainwater accumulation with the primary drainage system blocked and the secondary drainage system at its design flow rate when subjected to a rainfall intensity based on the I 00-yr hourly rainfall rate. Deflections of roof members are limited to prevent rainwater ponding on the roof. The roof structure is also designed to support the extreme winter precipitation load discussed in Section 13 .2.6.6. 13.2.6.4 Flooding Impact to the Facility and Structures, Systems, and Components Regional flooding from large precipitation events raising the water levels oflocal streams and river s to above the 500-yr flood level can have an adverse impact on the structure and SS Cs within. These impacts include the structural damage from water and the damage to power supplies and instrument control systems for SSCs relied on for safety. The infiltration of flood water into the facility could cause the failure of moderation control requirements and lead to an accidental nuclear criticality.

Direct damage or impairment of SSCs could also be caused by flooding in the facility.

The site will be graded to direct the stormwater from localized downpour s with a rainfall intensity for the I 00-yr storm for a 1-hr duration around and away from the RPF. Thus , no flooding from local downpours is expected based on sta ndard industrial design. Rainwater that falls on the waste management truck ramp and accumulates in the trench drain has lo w to no consequence for radiological , chemical , and criticality hazards. Situated on a ridge , the RPF will be located above the 500-yr flood plain according to the flood insurance rate map for Boone County, Missouri , Panel 295 (FEMA, 2011). The site is above the elevation of the nearest bodies of water (two small ponds and a lake), and no dams are located upstream on the local streams. This data conservatively provides a 2 x 10*3 year return frequency flood , which can be considered an unlikely event according to performance criteria.

However , the site i s located at an elevation of 2 4 8.4 m (815 ft), and the 500-year flood plain starts at an elevation of 242 m (795 ft), or 6.1 m (20 ft) below the site. Since the site is located only 6.1 m (20 ft) below the nearest high point on a ridge (relative to the local topography), is well above the beginning of the 500-yr flood plain, and is considered a dr y site, the probable maximum flood from regional flooding is considered highl y unlikel y, without further evaluation.

3 No additional IROFS are required.

13.2.6.5 Seismic Impact to the Facility and Struct ure s, Systems, and Components Beyond the impact on the facility structure and the potential for falling facility components impacting SSCs or direct damage to SSCs causing adverse events, other activities were identified as sensitive to seismic events. Durin g the irradiated target shipping cask unloading preparations, the shield plug fasteners wi ll be removed from an upright cask before mating the cask to the cask dockin g port. During the short period between that activity and installing the cask , a seismic event could dislodge the lift/cask combination and result in dislodging the s hield plug in the presence of personnel.

This event would result in potentially lethal doses to workers in a short period of time. Se i smic ground shaking can directly damage SSCs relied on for safety or lead to damage of the facility, including partial or comp l ete collapse, which could impact SSCs relied on for safety inside and outside the RPF. Damage to the facility could also impact SSCs, causing radiological and chemical releases of int e rmediate consequence.

3 The recommend e d sta ndard for de termi nin g the probably maximum flood, ANS 2.8 , D e t er minin g Design Basis Flooding at Power R eac tor Sites, has been withdrawn. 13-73

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NORTHWtsT MEOtCAL ISOTOPfS NWMl-2013-021, Rev. 2 Chapter 13.0 -Accide nt Analys is Leaks of fissile solution, compromising the safe-geometry and safe interaction storage in solid material storage arrays and pencil tanks or vessels containing enriched uranium solutions, could lead to a criticality accident, a high consequence accident.

NWMI-2015-SAFETY-004 , Section 3.1 , identifies IROFS to prevent and mitigate this accident scenario.

Dislodging the irradiated target shipping cask during unloading preparations could expose workers to a potentially lethal radiation dose. This event is considered a high consequence accident.

The safe-shutdown earthquake, or design basis earthquake, for the RPF is specified as the risk-targeted maximum considered earthquake (MCER), as determined in accordance with ASCE 7 and Federal Emergency Management Agency (FEMA) P-753 , NE HRP R ecom m en ded Seismic Provisions for New Building s and Other Structures.

The MCfa for this site is governed by the probabilistic considered earthquake ground-shaking, which has an annual frequency of exceedance of 4 x 10-4 (2,500-yr return period). This event is considered unlikely.

Using the provisions of ASCE 7 for standard industrial facilities with equivalent chemical hazards, Risk Category IV results in a design basis earthquake equal to the safe-shutdown earthquake specified.

When designed in accordance with ASCE 7 and companion standards, the maximum probability of a complete or partial structural failure is 3 percent conditioned on the occurrence of the considered earthquake ground-shaking, or a probability of failure of l .2 x 1 o-5. Therefore , failure of the facility subject to the maximum-considered earthquake ground-shaking is considered highly unlikely.

No credit can be taken for physical features of the irradiated target cask lifting fixture for the unmitigated case; therefore, the unmitigated likelihood is equal to the annual probability of exceedance for the safe shutdown earthquake, /earthqu ake = 4 x 10-4. 13.2.6.5.1 IROFS FS-04, Irradiated Target Cask Lifting Fixture As a PEC , the irradiated target cask lifting fixture wi ll be designed to prevent the cask from tipping within the fixture and prevent the fixture itself from toppling during a seismic event. 13.2.6.6 Heavy Snow Fall or Ice Buildup on Facility and Structures, Systems, and Components This evaluation addresses snow loading on the facility structure.

The facility protects the SSCs , and an extreme snow-lo ading event may cause failure of the roof , impacting the SSCs' ability to perform associated safety functions.

NRC DC/COL ISG-07 , Int e rim Staff Guidance on Assessment of Normal and Extreme Winter Precipitation Loads on the Roofs of Seismic Category I Structures, provides guidance on the design of Category I structures for snow load that conservatively bound s the RPF. The normal snow load as defined in the NRC ISG is the 100-yr snowpack, which is equivalent to the design snow load for Risk Category IV structures determined in accordance with ASCE 7. Collapse of the roof may damage SSCs that are relied on for safety, leading to accident sequences such as accidental nuclear criticality (e.g., a pencil tank was crushed and interaction controls violated) or a radiological release (e.g., if a hot cell confinement boundary was breached and a primar y confinement boundary damaged), or may prevent an SSC from being available to perform its function. The extreme winter precipitation load, as defined in the NRC ISG , is a combination of the 100-yr snowpack and the liquid weight of the probable maximum winter precipitation.

The probable maximum winter precipitation is based on the seasonal variation of the PMP , given in NWS/NOAA Hydrometeorological Report 53, S e asonal Variation of 10-Square Mile Probable Maximum Pr ecipi tation Estimates, United State s East of th e 105 1 h Meridian, for winter months. The PMP is defined in Section 13.2.6.3. Considering the extreme winter precipitation load is a combination of the 100-yr snowpack and the theoretical worst-case precipitation event, the extreme winter precipitation load is highly unlikely. 13-74 NWM I ...... * * ! N<HllTHWE.ST MEDJCAL ISOTOP£S NWMl-2013-021, Rev. 2 Chapter 13.0 -Accident Analysis The normal snow load is the 100-yr snowpack, which is equivalent to the design snow load for a Risk Category IV structure determined in accordance with ASCE 7. The return frequency of the normal snow load is relatively high and expected to be likely to occur during the design life of the facility.

The provisions of the ASCE 7 standard, when used with companion standards such as ACI 318 and AISC 360, are written to achieve the target maximum annual probabilities of failure established in ASCE 7. The highest probability of failure, which is for a failure that is not sudden and does not lead to a wide-spread progression of collapse, targeted for Risk Category IV structures is 5.0 x 1 o-6. Therefore, the likelihood of failure of the structure when subjected to the normal design snow load in conjunction with other loads as required by ASCE 7 is highly unlikely.

No additional IROFS are required.

13.2. 7 Other Accidents Analyzed A total of 75 accident sequences identified for further evaluation by the PHA were analyzed for the Construction Permit Application.

A summary of all accidents analyzed is provided in Table 13-24. This table includes the accidents evaluated in Section 13 .2.2 to 13.2.6 for completeness.

Table 13-24 lists each accident sequence number, a descriptive title of the accident, and IROFS identified (if needed) to prevent or mitigate the consequences of the accident sequence.

The preliminary IROFS for each sequence are listed in the far right column of Table 13-24. The IROFS number and title are provided.

If the accident sequence is bounded by the accidents discussed in Section 13.2.2 to 13.2.6, a pointer to the bounding accident sequence is listed. After further analysis, if the IROFS level controls were determined to not be required either due to reduced consequences or reduced frequency, this is stated. Other accident sequences have IROFS identified, and a pointer is included to the section where the control is discussed in more detail. Accident sequence designator Table 13-24. Analyzed Accidents Sequences (9 pages) from PHA Descriptor Preliminary IROFS Identified S.R.01 High-dose solution or enriched uranium solution spill causing a radiological exposure hazard

IROFS RS-01, Hot Cell Liquid Confinement Boundary

IROFS RS-03, Hot Cell Secondary Confinement Boundary

IROFS RS-04 , Hot Cell Shielding Boundary

IROFS CS-07, Pencil Tank and Vessel Spacing Control using Fixed Interaction Spacing ofindividual Tanks or Vessels

IROFS CS-08, Floor and Sum Geometry Control on Slab Depth , Sump Diameter or Depth for Floor Spill Containment Berms

IROFS CS-09, Double-Wall Piping

See Section 13.2.2.8 S.R.02 Spray release of solutions spilled

Bounded by S.R.01 from primary offgas treatment solutions, resulting in radiological consequences S.R.03 Spray release of high-dose or

Bounded by S.R.01 enriched uranium-containing product solution, resulting in radiological consequences 13-75

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NORTHWEST MEDICAL 1$0TOP£S NWMl-2013-021, Rev. 2 Chapter 13.0 -Accident Analysis Table 13-24. Analyzed Accidents Sequences (9 pages) Accident sequence designator from PHA Descriptor Preliminary IROFS Identified S.R.04 S.R.05 S.R.06 S.R.07 S.R.08 S.R.09 S.R.10 S.R.12 S.R.13 S.R.14 S.R.16 Liquid enters process vessel ventilation system damaging IRU or retention beds, releasing retained radionuclides High-dose solution enters the UN blending and storage tank High flow through IRU causing premature release of high-dose iodine gas Loss of temperature control on the IRU leading to release of high-dose iodine Loss of vacuum pumps Loss ofIRU or carbon bed media to downstream part of the system Wrong retention media added to bed or saturated retention media Mo product cask removed from the hot cell boundary with improper shield plug installation High-dose containing solution leaks to chilled water or steam condensate system IX resin failure due to wrong reagent or high temperature Backflow of high-dose radiological and/or fissile solution into auxiliary system (purge air, chemical addition line, water addition line, etc.)

IROFS RS-09, Primary Offgas Relief System

IROFS RS-03, Hot Cell Secondary Confinement Boundary

See Section 13.2.3.8

Not credib l e or low consequence

Bounded by S.R.04

Bounded by S.R.04

Bounded by S.R.04

Bounded by S.R.04

Event unlikely with intermediate consequence

Event unlikely with intermediate consequence

IROFS RS-04, Hot Cell Shielding Boundary

IROFS CS-06, Pencil Tank and Vessel Spacing Control using the Diameter of the Tanks, Vessels, or Piping

IROFS CS-10, Closed Safe-Geometry Heating or Cooling Loop with Monitoring and Alarm

IROFS CS-27, Closed Heating or Cooling Loop with Monitoring and Alarm

IROFS CS-20, Evaporator or Concentrator Condensate Monitoring

IROFS CS-18, Backflow Prevention Device

IROFS CS-19, Safe-Geometry Day Tanks

See Section 13.2.4.8

Bounded by S.R.01

Bounded by S.R.13 13-76

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NORTtfWf.$T MEDICAL ISOTOf'ES Accident sequence designator Table 13-24. Analyzed Accidents Sequences (9 pages) from PHA Descriptor Preliminary IROFS Identified S.R.17 S.R.18 Carryover of high-do se solution into condensate (a low-dose waste stream) High-dose solution flows into the solidification media hopper

IROFS RS-08, Sample and Analysis of Low Dose Waste Tank Dose Rate Prior to Tran sfe r Outside the Hot Cell Shielded Boundary

IROFS RS-10 , Active Radiation Monitoring and Isolation of Low-Dose Waste Transfer

See Section 13.2.7.1

Low consequence event that does not challenge IROFS RS-04 S.R.19 High target basket retrieval dose

Design evolved after PHA , accident sequence eliminated S.R.20 S.R.21 S.R.22 S.R.23 S.R.24 S.R.25 S.R.26 S.R.27 S.R.28 rate Radiological spill of irradiated LEU target material in the hot cell area Damage to the hot cell wall providing shielding Decay heat buildup in unprocessed LEU target material removed from targets leads to higher-dose radionuclide off gassing

Bounded by S.R.01

Low consequence event that does not damage shielding function ofIROFS RS-04

Low consequence event Offgassing from irradiated target

IROFS RS-03 , Hot Cell Secondary Confinement Boundary dissolution tank occurs when the

See Section 13.2.2.8 upper valve is opened Bagless transport door failure HEPA filter failure Failed negative air balance from zone-to-zone or failure to exhaust a radionuclide buildup in an area Extended outage of heat leading to freezing, pipe failure, and release ofradionuclides from liquid process systems Target or waste shipping cask or container not loaded or secured according to procedure, leading to personnel exposure

IROFS RS-03, Hot Cell Secondary Confinement Boundary

IROFS RS-04, Hot Cell Shielding Boundary

See Section 13.2.2.8

IROFS RS-03 , Hot Cell Secondary Confinement Boundary

See Section 13.2.2.8

IROFS RS-03, Hot Cell Secondary Confinement Boundary

See Section 13.2.2.8

Highly unlikely eve nt for proces s solutions containing fission products Bounded by S.C.04 for target fabrication syste ms

Information will be provided in the Operating License Application 13-77

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NORTHWEST MEDtcAL ISOTOP£S NWMl-2013-021, Rev. 2 Chapter 13.0 -Accident Analysis Table 13-24. Analyzed Accidents Sequences (9 pages) Accident sequence designator from PHA Descriptor Preliminary IROFS Identified S.R.29 S.R.30 S.R.31 S.R.32 S.C.01 S.C.02 S.C.03 S.C.04 High dose to worker from release of gaseous radionuclides during cask receipt inspection and preparation for target basket removal Cask docking port failures lead to high-dose to worker due to streaming radiation and/or high airborne radioactivity Chemical burns from contaminated solutions during sample analysis Crane load drop accidents Failure of facility enrichment limit Failure of administrative control on mass (batch limit) during handling of fresh U, scrap U, LEU target material, targets, and samples Failure of interaction limit during handling of fresh U, scrap U, LEU target material, targets , container s , and sample s Spill of process solution from a tank or process vessel leading to accidental criticality

IROFS RS-12 , Cask Containment Sampling Prior to Closure Lid Removal

IROFS RS-13, Cask Local Ventilation During Closure Lid Removal and Docking Preparations

See Section 13.2.7.1

IROFS RS-04, Hot Cell Shielding Boundary

IROFS RS-15, Cask Docking Port Enabling Sensor

See Sections 13.2.2.8 and 13.2.7.l

Judged unlikely event with intermediate con s equence

IROFS FS-01, Enhanced Lift Procedure

IROFS FS-02, Overhead Cranes

See Section 13.2. 7.1

Judged highly unlikely based on supplier's checks and balances

IROFS CS-02, Mass and Batch Handling Limits for Uranium Metal, [Proprietary Information], Targets, and Laboratory Sample Outside Process Systems

IROFS CS-03, Interaction Control Spacing Provided by Administrative Control

IROFS CS-04, Interaction Control Spacing Provided by Passively Designed Fixtures and Workstation Placement

See Section 13.2. 7.2

IROFS CS-02 , Mass and Batch Handling Limits for Uranium Metal , [Proprietary Information], Targets , and Laboratory Sample Outside Process Systems

IROFS CS-03 , Interaction Control Spacing Provided by Administrative Control

IROFS CS-04, Interaction Control Spacing Provided by Passively Designed Fixtures and Workstation Placement

See Section 13.2.7.2

IROFS CS-06, Pencil Tank, Vessel, or Piping Safe Geometry Confinement using the Diameter of Tanks, Vessels, or Piping

IROFS CS-07, Pencil Tank and Vessel Spacing Control using Fixed Interaction Spacing oflndividual Tanks or Vessels

IROFS CS-08, Floor and Sump Geometry Control of Slab Depth, Sump Diameter or Depth for Floor Spill Containment Berms

IROFS CS-09, Double-Wall Piping

IROFS CS-26, Processing Component Safe Volume Confinement

See Section 13.2.7.2 13-78

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MEDtcAL ISOTOPES NWMl-2013-021 , Rev. 2 Chapter 13.0 -Accident Analysis Table 13-24. Analyzed Accidents Sequences (9 pages) Accident sequence designator from PHA Descriptor Preliminary IROFS Identified S.C.05 Leak of fi s sile s oluti o n into the

Bound ed b y S.R 1 3 heatin g or coolin g jacket on th e tank o r vesse l S.C.06 System overflow to process ventilation involving fissile material S.C.07 Fi ss ile s olution l e ak s ac ros s m e chanical boundar y b e twe e n proc ess vesse l s and h ea tin g/coolin g jack ets into heatin g/c oo lin g medi a S.C.08 S.C.09 S.C.10 Backflow of high-dose radiological and/or fis s ile solution into auxiliary system (purge air , chemical addition line , water addition line , etc.) Hi g h co n ce ntrati o n s of uranium e nter th e co n c entrat o r o r eva p ora tor co nden sa t es High concentrations of uranium enter the low-dose or high-dose waste collection tanks

IROFS CS-11, Simple Overflow to Normally E mpty Safe Geometry Tank with Level Alarm

IROFS CS-12 , Condensin g Pot or Seal Pot in Ventilation Vent Line

IROFS CS-13, Simple Overflow to Normally E mpty Safe Geometry Floor with Level Alarm in the Hot Cell Containment Boundary

See Section 13.2.7.2

Bo und ed b y S.R.1 3

Bounded by S.R.13

IROFS CS-06 , Pencil Tan k, Vesse l , or Pipin g Safe Ge om etry Co nfin e m e nt u s in g th e Di ame t e r of T ank s , Vesse l s, o r Pipin g

IROFS CS-07 , P e ncil Tank and V esse l S p ac in g Co ntr o l Us in g Fixe d Int erac ti o n S p ac in g ofl ndi v idu al T an ks or V esse l s

IR O FS CS-26 , Pr ocess in g Co mp o n e nt Safe Vo lum e Co nfin e m e nt

See S e c t ion 1 3.2.7.2

IROFS CS-14, Active Discharge Monitorin g and Isolation

IROFS CS-15 , Independent Active Dischar g e Monitoring and Isolation

IROFS CS-16, Sampling and Analysis of Uranium Mass or Concentration Prior to Discharge or Disposal

IROFS CS-17, Independent Sampling and Analysis of Uranium Concentration Prior to Discharge or Disposal

See Section 13.2. 7 .2 13-79 I

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NOITKWHT MlDtCAl ISOTOi-fS NWMl-2013-021 , Rev. 2 Chapter 13.0 -Accident Analysis Table 13-24. Analyzed Accidents Sequences (9 pages) Accident sequence designator from PHA Descriptor Preliminary IROFS Identified S.C.11 S.C.12 S.C.13 S.C.14 S.C.15 S.C.17 High concentrations of uranium in contactor so lv ent regeneration aqueous waste High concentrations of uranium in the LEU target material wash solution High concentrations of uranium in the nitrous oxide scrubber High concentrations of uranium in the IX waste collection tanks effluent High concentrat i ons of uranium in the IX resin waste High concentrations of uranium in the solid waste encapsulation process

Bounded by S.C.04 and S.C. l 0

IROFS CS-04, Interaction Control Spacing Provided by Passively Designed Fixtures and Workstation Placement

IROFS CS-06, Pencil Tank, Vessel, or Piping Safe Geometry Confinement using the Diameter of Tanks, Vessels, or Piping

IROFS CS-07, Pencil Tank and Vessel Spacing Control Using Fixed Interaction Spacing oflndividual Tanks or Vessels

See Section 13.2.7.2

IROFS CS-06, Pencil Tank, Vessel, or Piping Safe Geometry Confine ment using the Diameter of Tanks, Vessels, or Piping

IROFS CS-16 , Samp lin g and Analysis of Uranium Mass or Concentration Prior to Discharge or Disposal

IROFS CS-17 , Independent Samp lin g and Analysis of Uranium Conce ntrati on Prior to Discharge or Disposal

See Section 1 3.2.7.2

IROFS CS-16, Sampling and Analysis of Uranium Mass or Concentration Prior to Discharge or Disposal

IROFS CS-17, Independent Sampling and Analysis of Uranium Concentration Prior to Discharge or Disposal

See Section 13.2.7.2

IROFS CS-06 , Pencil Tank, Ve ss el, or Piping Safe Geometry Confinement using the Diameter of Tanks , Vessels, or Piping

IROFS CS-07 , Pencil Tank and Vessel Spacing Contro l Using Fixed Interaction Spacing oflndividual Tanks or Vessels

IROFS CS-16 , Samp lin g and Ana l ys i s of Uranium Mass or Concentration Prior to Discharge or Disposal

IROFS CS-17 , Independent Samp ling and Analysis of Uranium Concentration Prior to Dischar ge or Disposal

See Section 13.2.7.2

IROFS CS-16, Sampling and Analysis of Uranium Mass or Concentration Prior to Discharge or Disposal

IROFS CS-17, Independent Sampling and Analysis of Uranium Concentration Prior to Discharge or Disposal

IROFS CS-21, Visual Inspection of Accessible Surfaces for Foreign Debris

IROFS CS-22, Gram Estimator Survey of Accessible Surfaces for Gamma Activity

IROFS CS-23, Nondestructive Assay ofltems with Inaccessible Surfaces

IROFS CS-24, Independent Nondestructive Assay ofltems with Inaccessible Surfaces

IROFS CS-25, Target Housing Weighing Prior to Disposal

See Section 13.2. 7.2 13-80 Accident sequence designator NWMl-2013-021 , Rev. 2 Chapter 13.0 -Accident Analysis Table 13-24. Analyzed Accidents Sequences (9 pages) from PHA Descriptor Preliminary IROFS Identified S.C.19 S.C.20 S.C.21 S.C.22 S.C.23 S.C.24 Failure of PEC -Component safe geometry dimension or safe volume Failure of concentration limits Target basket passive design control failure on fixed interaction spacing High concentration of uranium in the TCE evaporator residue High concentration in the spe nt si li cone oi l waste High uranium content on HEPA filters and subsequent failure

IROFS CS-06, Pencil Tank, Vesse l , or Piping Safe Geometry Confinement u s ing the Diameter of Tanks, Vesse l s, or Piping

IROFS CS-07, Pencil Tank and Vesse l Spacing Contro l Using Fixed Interaction Spacing oflndividual Tanks or Vessel s

IROFS CS-26, Processing Component Safe Vol um e Confinement

See Section 13.2.7.2

No credible path leading to criticality identified or not credible by design

IROFS CS-02, Mass and Batch Handling Limits for Uranium Meta l , [Proprietary Information], Targets , and Laboratory Sample Outside Process Systems

IROFS CS-03, Interaction Contro l Spacing Provided by Administrative Control

See Section 13.2. 7.2

IROFS CS-04, Interaction Control Spacing Provided by Passively Designed Fixtures and Workstation Placement

IROFS CS-06, Pencil Tank, Vessel, or Piping Safe Geometry Confinement Using the Diameter of Tanks, Vessels, or Pipin g

IROFS CS-07, Pencil Tank and Vessel Spacing Control Using Fixed Interaction Spacing oflndividual Tanks or Vessels

IROFS CS-16, Sampling and Analysis of Uranium Mass or Concentration Prior to Discharge or Di s posal

IROFS CS-17, Independent Sampling and Analysis of Uranium Concentration Prior to Di sc harge or Di s po sal

See Section 13.2. 7 .2

IROFS CS-04, Interaction Contro l Spac in g Provided b y Passively Designed Fixtures and Workstation Placement

IROFS CS-05 , Container Batch Volume Limit

IROFS CS-06 , Penci l Tank, Vessel, or Piping Safe Geo m etry Confi n ement Using th e Diameter of Tanks , Vesse ls , or Piping

IROFS CS-07 , Pencil Tank and Vessel Spacing Control Using Fixe d Interaction S pacin g oflndivid u a l Tank s or Vessels

IROFS CS-16, Sampling and Analysis of Uranium Mas s or Conce ntrati on Prior to Discharge or Disposal

IROFS CS-17, Independent Samp lin g and Ana l ys i s of Uranium Conce ntr ation Prior to Discharge or Disposal

See Sec ti on 13.2. 7 .2

Bounded by S.C.17 13-81 NWMl-2013-021, Rev. 2 Chapter 13.0 -Accident Analysis Table 13-24. Analyzed Accidents Sequences (9 pages) Accident sequence designator from PHA Descriptor Preliminary IROFS Identified S.C.27 Failure of administratively controlled container volume limits S.C.28 S.F.01 S.F.02 S.F.03 S.F.04 S.F.05 S.F.06 Crane load drop accidents Pyrophoric fire in uranium metal Accumulation and ignition of flammable gas in tanks or systems Hydrogen detonation in reduction furnace Fire in reduction furnace Fire in a carbon retention bed Accumulation of flammable gas in ventilation system components

IROFS CS-03, Interaction Control Spacing Provided by Administrative Control

IROFS CS-04 , Interaction Control Spacing Provided by Passively Designed Fixtures and Workstation Placement

IROFS CS-05, Container Batch Volume Limit

See Section 13.2.7.2

IROFS FS-01, Enhanced Lift Procedure

IROFS FS-02, Overhead Cranes

See Section 13.2.7.2

Event highly unlikely based on credible physical conditions

IROFS FS-03, Process Vessel Emergency Purge System

See Section 13.2.7.3

Judged highly unlikely based on credible physical conditions

Judged unlikely based on event frequency

IROFS FS-05, Exhaust Stack Height

See Section 13.2.7.3

Bounded by S.F.02 S.F.07 Fire in nitrate extraction system -* Event unlikely with intermediate or low consequences combustible solvent with S.F.08 S.F.09 S.F.10 S.F.11 S.N.01 S.N.02 S.N.03 uranium General facility fire Hydrogen exp l osion in the facility due to a leak from the hydrogen storage or distribution system Combustible fire occurs in hot cell area Detonation or deflagration of natural gas leak in steam generator room Tornado impact on facility and SSCs important to safety High straight-line winds impact the facility and SSCs important to safety Heavy rain impact on facility and SSCs important to safety

Information will be provided in the Operating License Application

Information will be provided in the Operating License Application

Information will be provided in the Operating License Application

Information will be provided in the Operating License Application

Judged highly unlikely event based on return frequency

Judged highly unlikely to result in structure failure

Bounded by S.N.06 13-82 NWMl-2013-021, Rev. 2 Chapter 13.0 -Accident Analysis Table 13-24. Analyzed Accidents Sequences (9 pages) Accident sequence designator from PHA Descriptor Preliminary IROFS Identified S. .04 Flooding impact to the facility and SSCs important to safety S.N.05 Seismic impact to the facility and SSCs important to safety

Judged highly unlikely event based on facility location above the 500-year flood plain

Judged highly unlikely to result in structure failure

IROFS FS-04 , Irradiated Target Cask Lifting Fixture

See Section 13.2.6.5 S.N.06 Heavy s nowfall or ice buildup on

Judged highly unlikely to re su lt in structure failure facility and SSCs important to safety S.M.01 Vehicle strikes SSC important to

Judged likely event with low consequence safety and causes damage or S.M.02 S.M.03 S.CS.01 H E PA IROFS IRU IX LEU Mo leads to an accident sequence of intermediate or high consequence Facility evacuation impacts on operations Localized flooding due to internal system leaka ge or fire suppression sprinkler activation Nitric acid fume release high-efficienc y particulate air. items relied on for safety. iodine removal unit. ion exchange.

low-enri c hed uranium. mol ybdenum.

Judged likely event with low consequence

IROFS CS-08, Floor and Sump Geometry Control of Slab Depth, Sump Diameter or Depth for Floor Spill Containment Berms

See Section 13.2. 7.2

No IROFS currently identified P EC pa ssive engineered co ntrol. PHA preliminary hazard s analysis. SSC st ru ct ures , systems , and co mpon e nt s. TCE trichloroethylene U uranium. UN ur anyl nitrat e. T a ble 13-25 provides a summary of all IROFS identified by the accident analyses performed for the Construction Permit Application.

Table 13-25 also identifies whether the IROFS were considered engineered safety features or administrative controls.

Engineered safety features are described in Chapter 6.0, and the administrative controls are discussed in Chapter 14.0, "Technical Specifications." Additional IROFS are anticipated to be identified (or the current IROFS modified) by additional design de ta il developed for the Operating License Application. 13-83

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NORTHWEST MfDICAL lSOTOP£S NWMl-2013-021, Rev. 2 Chapter 13.0 -Accident Analysis Table 13-25. Summary of Items Relied on for Safety Identified by Accident Analyses (2 pages) IROFS Engineered Administrative designator Descriptor safety feature control RS-01 Hot cell liquid confinement boundary ./ RS-02 Reserved RS-03 Hot ce ll secondary confinement boundary ./ RS-04 Hot cell shielding boundary ./ RS-05 Reserved RS-06 Reserved RS-07 Reserved RS-08 Samp l e and analysis oflow-dose waste tank dose rate prior to transfer outside the hot cell shielded boundary RS-09 Primary offgas relief system ./ RS-10 Active radiation monitoring and isolation oflow-dose waste transfer ./ RS-11 Reserved RS-12 Cask containment sampling prior to closure lid removal RS-13 Cask local venti lati on during closure lid remova l and docking pr eparations RS-14 Reserved RS-15 Cask docking port enabling sensor CS-01 Reserved CS-02 Mass and batch handling limits for uranium metal, [Proprietary

./ Information], targets, and laboratory sample outside process systems CS-03 Interaction control spacing provided by administrative control ./ CS-04 Interaction control spacing provided by passively designed fixtures ./ and workstation placement CS-05 Container batch volume limit CS-06 Pencil tank, vesse l , or piping safe geometry confinement using the ./ diameter of tanks, vessels, or piping CS-07 Pencil tank and vessel spacing control using fixed interaction

./ spacing of individual tanks or vessels CS-08 Floor and sump geometry control of slab depth, s ump diameter or ./ depth for floor sp ill containment berms CS-09 Double-wall piping ./ CS-10 C los e d safe geometry heating or cooling loop with monitoring and ./ alarm CS-11 Simple overflow to normally empty safe geometry tank with level ./ alarm CS-1 2 Co nd ensing pot or seal pot in ve ntilati on vent line ./ CS-13 Simple overflow to normally empty safe geometry floor with level ./ alarm in the hot cell containment boundary 13-84 NWM I ...... *.*

NORTHWtST MEDICAL ISOTI>f'ES NWMl-2013-021, Rev. 2 Chapter 13.0 -Accident Analysis Table 13-25. Summary of Items Relied on for Safety Identified by Accident Analyses (2 pages) IROFS Engineered Administrative designator Descriptor safety feature control CS-14 CS-15 CS-16 CS-17 CS-18 CS-19 CS-20 CS-21 CS-22 CS-23 CS-24 CS-25 CS-26 CS-27 FS-01 FS-02 FS-03 FS-04 FS-05 IROFS Active discharge monitoring and isolation Independent active discharge monitoring and isolation Sampling and analysis of uranium mass or concentration prior to discharge or disposal Independent sampling and analysis of uranium concentration prior to discharge or disposal Backflow prevention device Safe-geometry day tanks Evaporator or concentrator condensate monitoring Visual inspection of accessible surfaces for foreign debris Gram estimator survey of accessible surfaces for gamma activity Nondestructive assay of items with inaccessible surfaces Independent nondestructive assay of items with inaccessible surfaces Target housing weighing prior to disposal Processing component safe volume confinement Closed heating or cooling loop with monitoring and alarm Enhanced lift procedure Overhead cranes Process vessel emergency purge system Irradiated target cask lifting fixture Exhaust stack height items relied on for safet y. ./ ./ ./ ./ ./ ./ ./ ./ ./ ./ ./ ./ ./ ./ ./ ./ ./ ./ ./ The following subsections describe the IROFS that are not previously discussed elsewhere in this chapter. The IROFS are grouped according to their respective accident sequence categories , as shown in Table 13-26. Table 13-26. Accident Sequence Category Definitions 13.2.7.1 Items Relied on for Safety for Radiological Accident Seq u ences (S.R.) The following IROFS fall under the radiological accident sequence category and are not discussed elsewhere in this chapter. * .

S.R. S.C. S.F. S.N. S.M. s.cs. IROFS I Definition Radiological Criticality Fire or exp l osion Natural phenomena Man-made Chemical safety Section containing related IROFS description 13.2. 7.1 13.2.7.2 13.2.7.3 13.2.7.4 13.2.7.5 13.2.7.6 items relied on for safet y. 13.2.7.1.1 IROFS RS-08, Sample an d Analysis of Low Dose Waste Tank Dose Rate Prior to Transfer O u tside the Hot Ce ll Shie ld ed Boundary As an augmented administrative contro l (AAC), prior to transferring the solution from the low-dose waste tank to the low-dose waste encapsulation system outside of the hot cell shielded boundary, the low-dose waste tank will be administratively locked out, samp l ed, and the sample analyzed for high radiation.

Batches that satisfy the sample criteria can be transferred to the low-dose waste encapsulation system. 1 3-8 5 NWM I ...... *. * ! . NOffTHWl:ST MEOtCAl ISOTOPfS NWMl-2013-021 , Rev. 2 Chapter 13.0 -Accident Analysis The safety function o f this AAC is to prevent transfer of low-dose solution to outside the shielded boundary at radiation dose rates that would lead to intermediate-or high-dose consequences to workers. 13.2.7.1.2 IROFS RS-10, Active Radiation Monitoring and Isolation of Low Dose Waste Transfer As an AEC , the recirculating stream and discharge stream of the low-dose waste tank will be simultaneously monitored in a background shielded trunk out s ide of the hot cell shielded cavity. The continuous gamma-ray instrument monitoring the recirculation line and the transfer line will provide an open permissive signal to a dedicated isolation valve in the transfer line. The safety function of the system is to prevent transfer of low-dose waste solutions with exposure rates in excess of approved limits (safety limits and limiting safety system settings to be determined later) to outside the shielded boundary at radiation dose rates that would lead to intermediate-or high-dose consequences to workers or the public. The system functions by monitoring both the recirculation line for the low-dose waste collection tank and the transfer line to the low-dose waste encapsulation system outside of the hot cell shielded boundary. Monitoring will be performed in a shielded trunk , which reduces the background from the normally shielded hot cell areas to acceptable levels for monitoring.

In this closed-loop system, the gamma monitor will provide an open permissive signal to a fail-closed isolation valve in the transfer line , allowing the isolation valve to open. If the radiation levels exceed a safety limit setpoint during recirculation for sampling or during transfers, the isolation valve will be closed. The isolation valve will also fail closed on loss of power and loss of instrument air. 13.2.7.1.3 IROFS RS-12, Cask Containment Sampling Prior to Closure Lid Removal As an AEC , a sampling system will be connected to the cask vent to sample the atmosphere within the cask prior to closure lid removal. The system will sample the contents of the cask and have the ability to remediate the atmosphere using a vacuum system if dose rates are too high (safety limit s to be determined). The safety function ofIROFS RS-12 is to prevent personnel exposure to high-dose gaseous radionuclides.

The system wi ll identif y a hazardous concentration of high-dose gases in the cask, and if a high dose is identified , will remediate the situation through evacuation to a safe processing system. The system works by evacuating a sample of the gas and analyzing the sample as it passes by a detector.

If high activity is detected , the system will a l ann. The operator will use the system to evacuate and backfill the cask with fresh air (from a protected pressurized source such as a compressed bottle) until the atmospheres are within approved safety limits. 13.2.7.1.4 IROFS RS-13, Cask Local Ventilation During Closure Lid Removal and Docking Preparations As an AEC , a local capture ventilation system will be used over the closure lid to remove any escaped gases from the breathing zone of the worker during removal of the closure lid , removal of the shielding block bolts, and installation of the lifting lu gs. The safety function of IROFS RS-13 is to prevent exposure to the worker by evacuating any high-dose gaseous radionuclides from the worker's breathing zone and preventing immersion of the worker in a high-dose environment.

The system will use a dedicated evacuation hood over the top of the cask during containment closure lid removal. The gases will be removed to the Zone 1 secondary containment system for processing.

13-86 r ----NWMI ...... *

NOflTHWtST MEDtcAL ISOTDitf.S 13.2.7.1.5 IROFS RS-15, Cask Docking Port Enabling Sensor NWMl-2013-021, Rev. 2 Chapter 13.0 -Accident Analysis As an AEC, the cask docking port will be equipped with sensors that detect when a cask is mated with the cask docking port door. The sensors feed an enabling circuit that will prevent the door from being opened when no cask is present. The safety function of IROFS RS-15 is to prevent the cask docking port door from being opened , allowing a streaming radiation path to an accessible area and to prevent Zone II to Zone I air pressure imbalances that would allow air to migrate into the Zone II airlock. The system will also prevent a high streaming dose to workers from targets inside the hot cell , if the cask lift fails following mating. The system is designed to provide an enabling contact signal and positive closure signal when the sensor does not sense a cask mated to the door, causing the door to clo s e. 13.2.7.1.6 IROFS FS-01, Enhanced Lift Procedure As an Administrative Control (AC), lifts of high-dose rate containers or casks or of heavy objects (weight limit to be determined in final design) that move over hot cells in the standby or operating modes will use an enhanced lift procedure to reduce the likelihood of an upset. Enhancements will use the guidelines in DOE-STD-I 090-2011, Hoisting and Rigging , for critical lifts (for nonroutine cover block lifts) and engineered production lifts (for routine container and cask lifts using pre-engineered fixtures).

The safety function of IROFS FS-01 is to prevent (by reducing the likelihood) a dropped load or striking an SSC with a heavy load , causing damage that leads to an intermediate or high consequence event. The IROFS will be administered through the use of operating and maintenance procedures.

13.2. 7.2 Items Relied on for Safety for Criticality Accident Sequences (S.C.) The following IROFS fall under the criticality accident sequence category and are not discussed el s ewhere in this chapter. 13.2.7.2.1 IROFS CS-02, Mass and Batch Handling Limits for Uranium Metal, [Proprietary Information], Targets, and Laboratory Samples Outside Process Systems As a simple AC , mass and batch limits will be applied to handling , processing, and storage activitie s where uranium metal , [Proprietary Information] (LEU target material), targets , and/or samples are used. The mass or batch limits will be set such that the handled quantity can sustain double-batching or one in t eraction control failure with another approved quantity of fissile material , approved volume of fissile material , or an approved configuration for a tank, vessel , or IX column. Where safe batches are allowed , fixtures will be used to ensure that the safe batch is not exceeded (e.g., where [Proprietary Information]

are allowed as a safe batch , the operator will be provided with a carrying fixture that allows only [Proprietary Information]). For targets , the housing is credited for maintaining the contents dry. Final limits for each activity will be set in final design. 13.2. 7.2.2 IROFS CS-03, Interaction Control Spacing Provided by Administrative Control As a simple AC, while handling approved quantities of uranium metal, approved quantities of [Proprietary Information] (LEU target material), batches of targets, or batches of samples, an interaction control will be maintained between quantities being handled; fissile solution tanks , vessels , or IX columns; and safe-geometry ventilation housings.

Interaction control spacing will be set in final design when all process upsets are evaluated.

13-87 NWM I ...*.. * * ! NORTHWEST MEDICAL ISOTOPES NWMl-2013-021, Rev. 2 Chapter 13.0 -Accident Analysis 13.2. 7.2.3 IROFS CS-04, Interaction Control Spacing Provided by Passively Designed Fixtures and Workstation Placement As a PEC , fixed interaction control fixtures or workstations will be provided for holding or processing approved containers with designated quantities of uranium metal, quantities of [Proprietary Information] (LEU target material), batches of targets , and batches of samples. The fixtures are designed to hold only the approved container or batch and are fixed with 61 centimeter (cm) (2-ft) edge-to-edge spacing from all other fissile material containers , workstations , or fissile solution tanks, vessels, or IX columns. Where LEU target material is handled in open containers , the design should prevent spills from readily spreading to an adjacent workstation or storage location.

Final workstation and fixture spacing will be determined in final design when all process upsets are evaluated.

Workstations with interaction controls will include the following (not an all-inclusive listing):

LEU target material trichloroethylene (TCE) wash column workstation containing a geometry funnel LEU target material ammonium hydroxide rinse column workstations containing safe-geometry funnels Target basket fixture that provides safe spacing of a batch of targets from another batch in the target receipt cell 13.2.7.2.4 IROFS CS-05, Container Batch Volume Limit As a simple AC to address the activity of sampling and small quantity storage , a volumetric batch limit will be applied such that the total number of small sample or storage containers is controlled to a safe total volume. Many activities at the RPF will involve very high-dose solutions; only small quantities of a sample may be removed from the shielded area for analysis due to radiological reason s. As a result , sample bottles will be relatively small. The uranium content in these containers will often be unknown. To provide safe storage and handling in the laboratory environment, a safe volumetric batch limit on these small containers will be applied. Some potentially contaminated uranium waste streams will also be generated at the RPF that require quantification of the uranium content prior to disposal.

These waste streams will need a safe volume container for interim storage while the uranium content is being identified.

The final set of approved containers and volumes will be provided during final design when a ll process upsets are evaluated.

13.2.7.2.5 IROFS CS-11, Simple Overflow to Normally Empty Safe Geometry Tank with Level Alarm As a PEC, for each vented tank containing fissile or potentially fissile process solution for which IROFS CS-11 is assigned , a simple overflow line will be insta ll ed below the level of the process vessel ventilation port and any chemical addition ports (where an anti-siphon safety feature will be installed). The overflow drain will prevent the process solution from entering the respective non-geometrically favorable portions of the process ventilation system and any chemical addition ports (where the solutions will enter through anti-siphon devices).

The safety function of this feature is to prevent accidental nuclear criticality in geometrically favorable portions of the process ventilation system. The overflow will be directed to a safe-geometry storage tank, which will normally be empty. The overflow storage tank will be equipped with a level a larm to inform the operator when use of the IROFS has been initiated so that actions may be taken to restore operability of the safety feature by emptying the tank. The locations where this IROFS is used will be determined during final design. 13-88 NWM I ...... *

NORTHWEST IWNCAL lSOTOfl'ES NWMl-2013-021, Rev. 2 Chapter 13.0 -Accident Analysis 13.2.7.2.6 IROFS CS-12, Condensing Pot or Seal Pot in Ventilation Vent Line As a PEC, downstream of each tank for which IROFS CS-12 is assigned, a safe geometry condensing pot or seal pot will be installed to capture and redirect liquids to a safe-geometry tank or flooring area with safe-geometry sumps. One such condensing or seal pot may service several related tanks within the geometry boundary of the ventilation system. The condensing or seal pot will prevent fissile solution from flowing into the respective non-geometrically favorable process ventilation system by directing the solution to a safe-geometry tank or flooring area with safe-geometry sumps. The safety function ofIROFS CS-12 is to prevent accidental nuclear criticality in non-geometrically favorable portions of the process ventilation s y stem. The safe-geometry tank or sumps will be equipped with a level alarm to inform the operator when use of the IROFS has been initiated. Each individual tank or vessel operation must be evaluated for required capacity for overflow to ensure that a suitable overflow volume is available. A monitoring and alarm circuit will be provided so that common overflow tanks or safe slab flooring or sumps may be used for multiple tanks or vessels , and limiting conditions of operation will be defined to ensure that the IROFS is made available in a timely manner or operations are suspended following an overflow event of a single tank. Where independent seal or condensing pots are credited, the drains of the seal or condensing pots must be directed to independent locations to prevent a common clog or overcapacity condition from defeating both. 13.2.7.2.7 IROFS CS-13, Simple Overflow to Normally Empty Safe Geometry Floor with Level Alarm in the Hot Cell Containment Boundary As a PEC for each vented tank containing fissile or potentiall y fissile process solution for which IROFS CS-13 is assigned , a simp l e overflow line will be installed above the high alarm setpoint.

The overflow w i ll be directed to one or more s afe-geometry flooring configurations with safe-geometry sumps. IROFS CS-13 will prevent accidental criticality by ensuring that overflowing fissile solutions are captured in a safe-geometry slab configuration with safe-geometry sumps. These flooring areas (separated as needed to support operations in different hot cell areas) will normally be empty. The flooring areas will be equipped with a sump level alarm to inform the operator when use of the IROFS has been initiated.

13.2.7.2.8 IROFS CS-14, Active Discharge Monitoring and Isolation As an AEC for discharges from safe-geometry systems to non-favorable geometry systems, an active ur a nium detection system will be used to close an isolation valve in the discharge line at a uranium concentration limit and/or cumulative mass limit (the limit[s] to be set sufficiently low to preclude on process upsets and sufficientl y high to maintain an operating limit setpoint below the safety setpoint). This system will prevent a high-concentration uranium solution from being discharged to a non-favorable geometry system. The safety function of IROFS CS-14 is to prevent an accidental nuclear criticality. The closed-lo op system is designed to iso lat e the discharge points listed below by actively monitoring the solution stream for uranium concentration using a suitable uranium monitor. At a limitin g setpoint , the uranium monitor will close an isolation valve in the discharge line to stop the discharge. The uranium monitor is designed to pro duce a va l ve-open permissive signa l that fails to an open state, closing the valve on loss of electrical power. The isolation valve is designed to fail-closed on loss of instrument air , and the solenoid is designed to fail-closed on loss of signal. The locations where this IROFS is used will be determined during final design. 13-89 NWMI ...... *

NOflTtfWEST MEDICAL ISOTOPE S NWMl-2013-021, Rev. 2 Chapter 13.0 -Accident Analysis 13.2.7.2.9 IROFS CS-15, Independent Active Discharge Monitoring and Isolation As an AEC for discharges from safe-geometry systems to non-favorable geometry systems, an independent active uranium detection system will be used to close an independent isolation valve in the discharge line at a uranium concentration limit and/or cumulative mass limit (the limit[s] to be set sufficiently low to preclude follow-on process upsets and sufficiently high to maintain an operating limit setpoint below the safety setpoint).

This system will prevent a high concentration uranium solution from being discharged to a non-favorable geometry system. The safety function ofIROFS CS-15 is to prevent an accidental nuclear criticality.

The closed-l oop system is designed to isolate the discharge points listed below by actively monitoring the solution stream for uranium concentration using a s uitable monitor to detect uranium. At a limiting setpoint, the monitor will close an isolation valve in the discharge line to stop the discharge. The monitor is designed using a different monitoring method and i so lation valve than used in IROFS CS-14 to produce a valve-open permissive signal that fails to an open state, closing the valve on loss of electrica l power. The isolation valve is designed to fail-c los ed on loss of instrument air , and the solenoid is designed to fail-closed on loss of signal. The locations where this IROFS is used will be determined during final design. 13.2.7.2.10 IROFS CS-16, Sampling and Analysis ofU Mass/Concentration Prior to Discharge/Disposal As an AAC , prior to initiating di sc harge from the safe-geometr y container, tanks, or vessels ass i gned IROFS CS-16 to non-favorable geometry systems, the container, tank , or vessel will be isolated and placed under admjnistrative control , recircu l ated or otherwise uniformly mi xe d , sampled, and the sample analyzed for uranium content. The discharge or disposal will only be approved following independent review of the sample results to confirm that the uranium content is below a concentration or a mass limit (to be determined for each individual application based on expected vo lum es and follow-on processing needs) and under the independent oversight of a supervisor (who administratively controls the locks on the discharge system). Uraruum mass in the disposal container or vessel will be tracked to ensure that the mass or concentration limit for the container is not exceeded. The safety function of IROFS CS-16 is to prevent accidental nuclear criticality caused by discharging or disposing of high-concentration uranium to an uncontrolled system. The IROFS functions as described by ensuring, through physical sampling and analysis, that the uranium content of an isolated container , tank, or vessel (both inlets and outlets isolated , as applicable) is below a safe, single parameter limit on so luti on concentration or under a safe mass for the disposal container.

Systems, tanks , or vesse l s for which IROFS CS-16 applies, include: TCE recycle tanks Spent silicone oil Condensate tanks (either as normal or backup controls) 13.2. 7.2.11 IROFS CS-17, Independent Sampling and Analysis of U Concentration Prior to Discharge/Disposal As an AAC, prior to initiating discharge from the safe-geometry tanks or vessels assigned IROFS CS-17 to non-favorable geometry systems, the tank or vessel will be isolated and placed under administrative control , recirculated , samp l ed, and the sample analyzed for uranium content. The recirculation or uniformly mixing , sampling, and analysis activities will be independent (performed at a different time , using different operators or laboratory technicians , and different ana l ysis equipment, checked with independent standards) of that performed in IROFS CS-16. 13-90 NWMl-2013-021 , Rev. 2 Chapter 13.0 -Accident Analysis The di sc harge or disposal will only be approved following independent re view of the sample result s to confirm the uranium content is below the limiting setpoint for uranium concentration or batch mass for the contents and under the indep e ndent oversi g ht of a supervisor (who administratively controls the locks on the discharge system). Uranium mass in the disposal container or vessel wi ll be tracked and independently verified to ensure that the mas s or concentration limit for the container i s not exceeded.

The safety function of IROFS CS-17 is to pre vent accidental nuclear criticality caused b y discharging concentration uranium to an uncontrolled system. The IROFS functions as described b y ensuring, throu g h ph ysica l sampling and analysis, that the uranium content of an isolated tank or vessel is below a safe, single parameter limit on solution concentration or mass for a disposal container.

Systems, tanks , or vessels for which IROFS CS-1 7 applies includ e: TCE rec ycle tanks Spent silicone oi l Condensate tanks (either as normal or backup controls) 13.2.7.2.12 IROFS CS-21, Visual Inspection of Accessible Surfaces for Foreign Debris As a simple AC, a vis ual inspection will be performed to identif y foreign matter on accessible surfaces of equipment and waste materials approved for this method prior to disposal.

All visible foreign material i s assumed to be uranium. All surfaces must be non-porous.

Materials in volve d must be so lids (no solutions or liquids present).

All s urfaces must be visually accessible either directly or through approved in s pection device s. The inspection criterion is for no foreign material of discernible thickne ss to be visible (transparent films allowed).

The safety function of thi s AC is to ensure that no sign ificant uranium deposits exist on the item being disposed , to prevent an accumulation of a minimum subcritical mass of uranium in the disposal container.

The control will be exercised at designated waste consolidation stations, holding specifically approved waste containers, and on the item s approved b y the Criticality Safety Manager. The waste will not be conso lida ted until independent measurements conducted according to IROFS CS-22 or IROFS CS-24 have been completed.

The item will be controlled during the waste measurement analysis period. Items initiall y approved include disa sse mbled irradiated or scrap target hou sing parts or pieces. 13.2.7.2.13 IROFS CS-22, Gram Estimator Su rve y of Accessible Surfaces for Gamma Activity As an AAC, a gram estimator survey will be performed on all accessible s urfaces of equipment and waste m a terials approved for this method prior to disposal.

The survey will be p erfor med on low-risk waste streams that have surfaces that are I 00 percent accessib l e with the measurement instrument.

The measurement setpoint is designed to detect activity from 15 g of 2 35 U uniformly spread over 30 kilograms (kg) of 4-mil (thousandth of an inch) thick pol yet hylene sheeting (both s ides) as a bounding waste form for disposal at the U.S. Department of Transportation (DOT) fissile-excepted limit of 0.5 g 235 U/L kg fissile material.

The purpose of this IROFS is to provide a backup instrument AAC to visual inspection (IROFS CS-21) for bulking and disposal of low-risk waste to pr eve nt accidental nuclear criticality.

All surfaces will need to be accessible to the instrument used. The waste stream mu s t not be contaminated with significant fission product radionuclides since all activity is attributed to uranium. This survey will be performed as backup to the visual inspection described in IROFS CS-21. An independent person from the one performing the visual inspection ofIROFS CS-21 will perform the survey. The control will be exercised at designated waste consolidation stations, holding specifically approved waste containers, on the waste items using survey instrument(s) and setpoint(s) approved by the Criticality Safety Manager. Waste consolidation will be conducted after independent verification of the two methods of quantifying uranium mass has been performed.

IROFS CS-22 is applicab l e to radiological waste ge nerated outside the hot cell boundary that has had a low risk for direct contact with uranium-bearing material s. 13-91 NWM I ...... ' *

NOfllffWtrT MEDICAi. ISOTIM'ES NWMl-2013-021 , Rev. 2 Chapter 13.0 -Accident Analysis 13.2.7.2.14 IROFS CS-23, Non-D estructive Assay of Items with Inaccessible Surfaces As an AAC , a nond estruct ive assay (NDA) method will be used on approved waste streams to quantify the uranium mass prior to disposal.

An approved waste container with an approved uranium mass limit will receive the waste. A running inventory of items and uranium mass will be maintained with the waste disposal container.

The purpose of this IROFS is to prevent accidental nuclear criticality by controlling the mass of enriched uranium that is disposed in a non-safe geometry waste container.

At designated waste consolidation stations holding specifically approved waste containers, the control will be exercised on the waste items using NDA techniques and mass or concentration limits approved by the Crit icality Safety Manager. The waste will not be consolidated until independent measurements conducted according to IROFS CS-24 are completed.

The item will be controlled during the waste measurement analysis period. 13.2.7.2.15 IROFS CS-24, Independent NDA of Items with Inaccessible Surfaces As an AAC, an independent NDA method will be u se d on approved waste streams to quantify the uranium mass prior to disposal.

An approved waste container with an approved uranium mass limit will receive the waste. A running inventory of item s and uranium mass will be maintained with the waste disposal container.

The purpose of this IROFS is to prevent accidental nuclear criticality by controlling the mass of enriched uranium that is disposed in a non-safe geometry waste container.

The control will be u sed as a backup to IROFS CS-16, IROFS CS-21 or IROFS CS-23, as approved b y the Criticality Safety Manager for each waste stream. At designated waste consolidation stations holding specifically approved waste containers, the control will be exerc ised on the waste items using NDA techniques and mass or concentration limit s approved by the Cr i ticality Safety Manager. Waste consolidation will be conducted after independent verification of the two methods of quantifying uranium mas s has been performed. 13.2.7.2.16 IROFS CS-25, Target Housing Weighing Prior to Disposal As an AAC, on disposal of empty target housing s, target hou s ing piece s will be weighed and the weight compared to the original housing tare weight. The removed LEU target material will be weighed, and the weight compared to the original loading of LEU target material prior to disposal.

The weights will agree within tolerances approved by the Criticality Safety Manager. Any differences will be attributed as [Proprietary Information]

mass remaining in the wastes. An approved waste container with an approved uranium mass lim it will receive the waste. A running inventor y of items and uranium ma ss will be maintained with the waste dispo sa l container.

The purpose of this IROFS is to prevent accidental nuclear criticality by controlling the mass of enriched uranium that is disposed in a non-safe geometry waste container.

The control will be used as a backup to IROFS CS-16 for the disposal of target housings. At designated waste consolidation stations holding specifically approved waste containers, the control wi ll be exercised on the waste items weighed on approved sca l es and at mass or concentration setpoint(s) approved by the Criticality Safety Manager. Waste consolidation will be conducted after independent verification of th e two method s of quantifying uranium mass (the go/no-go method ofIROFS CS-16 , and the quantitative method of IROFS CS-25) have been performed.

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NOflTHWUT M£OtCAl. ISOTOPES NWMl-2013-021 , Rev. 2 Chapter 13.0 -Accident Analysis 13.2. 7.2.17 IROFS CS-26, Processing Component Safe Volume Confinement A s a PEC, some processing components (e.g., pumps , filter housings , and IX columns) will be controlled to a safe volume for safe storage and processing of fissile solutions.

The safety function of the safe volume component is also one of confinement of the contained solution.

The safe volume confinement of fissile solutions will prevent accidental nuclear criticality, a high consequence event. The safe volume confinement conservatively includes the outside diameter of any heating or cooling jackets (or any other void spaces that may inadvertently capture fissile solution) on the component.

Where insulation is used on the outside wall of the component , the insulation will be closed foam or encapsulated type (so as not to soak up solution during a l eak) and wi ll be compatible with the chemical nature of the contained solution.

13.2.7.3 Items Relied on for Safety for Fire or Explosion Accident Sequences (S.F.) The following IROFS fall under the fire or explosion accident sequence category and are not discussed el s ewhere in this chapter. 13.2.7.3.1 IROFS FS-05, Exhaust Stack Height As a PEC, the exhaust stack is designed and fabricated with a fixed height for safe release of the gaseous effluents.

13.2. 7.3.2 IROFS FS-02, Overhead Cranes Overhead cranes will be designed , operated , and tested according to ASME B30.2 , Ov e rhead and Gantry Cranes (Top Running Bridge, Single or Multipl e Girder , Top Running Trolley Hoist). Lifting devices for shipping containers will be designed , operated , and tested according to ANSI N14.6 , Standard for Sp ec ial Lifting D e vices for Shipping Containers Weighin g 10 , 000 Pounds (4 , 500 k g) or Mor e for N uclear Materials.

The safety function ofIROFS FS-02 is to prevent (by reducing the likelihood) mechanical fai lur e of cranes during heavy lift activities. This IROFS will be implemented through the facilities configuration management and management measures programs.

13.2.7.3.3 IROFS FS-03, Process Vessel Emergency Purge System A s an AEC, an emergency backup set of bottled nitrogen gas will be provided for tanks that have the potential to reach the hydrogen lower flammability limit either through the radiolytic decomposition of w a ter or through reaction with the nitric acid (or other reagents added during processing

). The system will monitor the pressure or flow going to the header and open an isolation valve on low pressure or flow (setpoint to be determined) to restore the sweep gas flow to the system using nitrogen. The system will be configured to provide more than 24 hr of sweep gas for the required tanks. The safety function ofIROFS FS-03 is to prevent a hydrogen-air mixture in the tanks from reaching lower flammability limit conditions to prevent the deflagration or detonation hazard. The purge gases will be exhausted through the dissolver offgas or the process vessel ventilation system. The system is designed to sense low pressure or flow on the normal sweep system and introduce a continuous purge of nitrogen from a reliable emergency backup station of bottled nitrogen into each affected vessel. 13.2.7.4 Items Relied on for Safety for Natural Phenomena Accident Sequences (S.N.) The IROFS under the natural phenomena accident sequence category are discussed in Section 13 .2.6. 13-93 NWM I ...... * * ! HOmfWEST MEOfCAl lSOTOPH NWMl-2013-021, Rev. 2 Chapter 13.0 -Accident Analysis 13.2.7.5 Items Relied on for Safety for Man-Made Accident Sequences (S.M.) There are no IROFS specifically identified for the man-made accident sequence category. 13.2.7.6 Items Relied on for Safety for Chemical Accident Sequences (S.CS.) There are no IROFS specifically identified for the chemical accident sequence category. 13-94 NWMl-2013-021, Rev. 2 Chapter 13.0 -Accident Analysis 13.3 ANALYSIS OF ACCIDENTS WITH HAZARDOUS CHEMICALS T h is section analyzes the hazardous chemical-based accident sequences identified in the PHA. 13.3.1 Chemical Burns from Contaminated Solutions During Sample Analysis 13.3.1.1 Chemical Accident Description This accident sequence occurs during sampling and analysis activities performed outside the hot cell confinement and shielding boundary where facility personnel (operators and/or technicians) may handle radioactively contaminated acidic or caustic solutions.

There are two possible modes of occurrence for this accident.

A sample container is dropped during handling activities outside a laboratory hood , resulting in a spill/splash event. A spill occurs during sample handling or ana l ysis where the container is required to be opened. 13.3.1.2 Chemical Accident Consequences Either of the modes described above can result in damage to skin and/or eye tissue on exposure to the acidic or caustic sample solution.

This accident se quence may result in long-term or irreversible tissue damage, particularly to the eyes. 13.3.1.3 Chemical Process Controls Facility personnel will be required to follow strict protocols for sampling and analysis activities at the RPF. Sampling locations , techniques , containers to be used , routes to take through the RPF when transporting a sample , ana l ysis procedures , reagents , analytical equipment requirement s, and sample material disposal protocols will all be specified per procedure s and/or work plans prepared and discu sse d pr i or to sampling or analytical activities.

Operators and technicians will be required to wear personal protective equipment, specifically for eye and skin protection.

R a diologically contaminated acidic and caustic so lution samples will be handled in approved containers.

Containers will be properly sealed when removed from sample locations and vent hoods during transport and/ or storage. Sample containers will also be opened only when securely located in an approved laboratory hood , with the hood lowered for spray protection.

This process wi ll pro vi de an additional la ye r of protection for eyes and skin (e.g., protective eyewear/face shield, laboratory coat or apron, anti-contamination chemical re sis tant gloves, etc.). 13.3.1.4 Chemical Process Surveillance Requirements Specific surveillance requirements will be identified in the Operating Permit Application.

For this accident sequence, surveillance may consist of management auditing or oversight of sampling and analysis activities to ensure adherence to the specified protocol of procedures, personal protective equipment usage, approved container usage , and laboratory hood etiquette.

13-95

... NWM I ...... . ' *.* ! NOKTifWHT MEDfCAl ISOTOPES 13.3.2 Nitric Acid Fume Release 13.3.2.1 Chemical Accident Description NWMl-2013-021, Rev. 2 Chapter 13.0 -Accident Analysis This accident consists of a release of nitric acid fumes inside or outside of the RPF originating from one of the nitric acid storage tanks in the chemical storage and preparation room. 13.3.2.2 Chemical Accident Consequences Chapter 19 .0 identifie s hazardous chemical relea se scenarios for the facility using several of the stored chemicals.

A 1-hr release of the bounding RPF inventory of 5,000 L of nitric acid was shown to cause a concentration of 1 ,200 parts per million (ppm) at the controlled area fence line and 19.1 ppm at 434 m (1,425 ft) (nearest resident location) under disper s ion conditions of moderate wind. Unmitigated exposure to a nearb y worker would be much higher. The AEGL-2, 60-minute (min) exposure limit for nitric acid is 24 ppm , which is high consequence to the public. AEGL-3, the 10-min exposure limit , is 1 70 ppm for a high consequence exposure to the worker. These determinations were made using the ALOHA (Areal Locations of Ha zar dous Atmospheres) computer code for estimating the consequences of chemical releases. The use of ALOHA is recognized by the NRC in NUREG/CR-6410. The impact and consequences of a chemical releas e on RPF operations would r e quire personnel to either evacuate the facility or, under some circumstances, shelter in place depending on the location of the event. 13.3.2.3 Chemical Process Controls The RPF will follow U.S. E nvironmental Protection Agency and Occupational Safety and Health Administration regulations for design, construction, and operation of chemical preparation and storage areas. Chemical handling procedures will be provided to operators to ensure safe handling of chemicals according to applicable regulatory requirements and consistent with the applicable material safety data sheets. IROFS to prevent or mitigate events that could impact the chemical storage tanks in the RPF chemical storage and preparation room are addressed in Section 13.2.5. 13.3.2.4 Chemical Process Surveillance Requirements Specific surveillance requirements for chemical use and storage at the RPF will be identified in the Operating Permit Application.

13-96 NWM I ...... * * ! NOmfMST MEDICAL lSOTDPU

13.4 REFERENCES

NWMl-2013-021 , Rev. 2 Chapter 13.0 -Accident Analysis 10 CFR 20, "Standards for Protection Against Radiation ," Code a/Federal Regulations, Office of the Federal Register, as amended. 10 CFR 30, " Rul es of General App licability to Domestic Licens ing of Byproduct Material," Code of Federal Regulations , Office of the Federal Register, as ame nd ed. 10 CFR 50, "Domestic Licensing of Production an d Utilization Facilities," Code of Federal Regulations, Office of the Federal Register , as amended. 10 CFR 70, "Domestic Licensing of Special Nuclear Material," Code of Federal R egulations, Office of the Federal R egister, as amended. 10 CFR 70.61 , "Perfo rmance Requirements," Code a/Federal Regulations, Office of the Federal Register, as amen d ed. 10 CFR 71, "Packaging and Transportation of Radioactive Material,

Code of Federal Regulations, Office of the Federal Register , as amended. A C I 318, Building Code Requirements for Stru ct ural Concrete, American Concrete Institute, Farmington Hills , Michigan, 2014. AISC 360, Specification for Structural Steel Buildings, American In stitute of Steel Construct ion , C hicago , Illinois, 2010. ANS 2.8, Determining Design Basis Flooding at Pow er R eac tor Sites, American Nuclear Society, La Grange Park, Illinoi s, 1992 , W2002. ANSI N 14.6, Standard for Special Lifting Devic es for Shipping Containers Weighing 10 , 000 Pounds (4,500 kg) or More for Nuclear Materials , American Nuclear Society, La Grange Park, Illinoi s, 1993. ANSI/ ANS-8.1, Nuclea r Criticality Safety in Op e rations with Fissionable Material Outside Reactors, American Nuc l ear Society, La Grange Park, lllinoi s, 1998 (Reaffirmed 2007). ASCE 7, Minimum D esig n Load s for Building s and Oth er Structures, American Society of Civil Engineers, Reston, Virginia, 2010. ASME B30.2, Overhead and Gantry Cranes (Top Runnin g Bridge , Single or Mu ltipl e Girder , Top Running Trolley Hoist), American Society of Mechanical Engineers, New York, New York, 2005. CDC, 2010, NIOSH Pocket Guide to Chemical Ha z ards , 2010-168c, Centers for Disease Co ntrol and Prevention, http://www.cdc.gov/nios h/npg/, do wn lo aded February 27, 2015. DC/COL ISG-07 , Interim Staff Guidance on Assessment of Normal and Extreme Winter Precipitation Loads on the Roofs of S eis mic Category I Structures, U.S. Nuclear Regulatory Commission, Washington , D.C., 2008. DOE-HDBK-30 I 0, DOE Handbook-Airborne Release Fractions/Rat es and R esp irabl e Fractions for No nr eactor Nuclear Facilities, Change Not ice No. 1 , U.S. Department of Energy, Washington , D.C., December 1994 (R2013). DOE-STD-1090-2011, Hoisting and R igging, U.S. D epartment of Energy , Washington , D.C., September 30, 20 11. 13-97

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MEDICAL ISOTOP£S NWMl-2013-021, Rev. 2 Chapter 13.0 -Accident Analysis EPA 520 11-88-020 , Federal Guid a nce Report No. 11, Limiting Valu e s of Radionuclide Intak e and A ir Concentration and Dos e Conversion Fa c tors for Inhalation , Subm e rsion , and In ges tion, U.S. Environmental Protection Agency, Washington , D.C., September 1988. FEMA , 2011, "Flood Insurance Rate Map," Panel 295 of 470, Boone County , Missouri and Incorporated Areas , Map# 29019C0295D , Federal Emergency Management Agency , Washington , D.C., March 17 , 2011. FEMA P-753 , N EHRP R ec omm e nd ed S e ismic Provisions for Ne w Building s and Other Structur e s , Federal Emergency Management Agency , Washington , D.C., 2009. Hydrometeorological Report No. 51 , Probabl e Ma x imum Pr e cipitation Estimat e s , Unit e d Stat e s East of th e 105th M e ridian , U.S. Department of Commerce , National Oceanic and Atmo s pheric Administration , Washington , D.C., 1978. Hydrometeorological Report No. 53 (NUREG/CR-1486), S e a s onal Variation of JO-Squar e Mil e Probabl e Maximum Pr e cipitation E s timat e s , U nit e d States East of the 105 1 h M e ridian, U.S. Department of Commerce , National Oceanic and Atmospheric Administration , U.S. Nuclear Regulatory Commission, Office of H y drology National Weather Service, Washington , D.C., April 1980. IBC , 2012 , International Building C od e, as amended , International Code Council , Inc., Washington , D.C., February 2012. ICRP-26 , R e comm e ndations of th e International C ommission on Radiolo g i c al Protection , International Commission on Radiological Protection , Ottawa , Canada , 1977. ICRP-30 , Limits for Intak es of Radionuclides by Workers , International Commission on Radiological Protection , Ottawa , Canada , 1979. ICRP-72, Age-D e p e ndent Doses to the M e mber s of the Publi c from Intak e of Radionuclid e s -Part 5 Compilation of Ing e stion and Inhalation Co e ffici e nts , International Commission on Radiological Protection , Ottawa , Canada , 1995. LA-13638 , A Review of Criti c ali ty A c cid e nts , Los Alamos National Laboratory , Los Alamos , New Mexico, 2000. NAP 1994 , Estimating Bounds on Extrem e Pr ec ipitation Ev e nts , National Academy Press , National Research Council, Washington, D.C., I 994. NOAA Technical Report NWS 25 , Comparison of Generaliz e d Estimat e s of Probabl e Maximum Pr ec ipitation w ith Great es t Obs e rv e d Rainfalls , National Oceanic and Atmospheric Administration , Washington , D.C., 1980. NUREG-153 7, Guid e lin e s for Pr e paring and R ev i ew ing A ppli c ations for th e Li ce nsing of N on-Pow e r R e a c tors -Format and Cont e nt , Part 1 , U.S. Nuclear Regulatory Commission , Office of Nuclear Reactor Regulation, Washington, D.C., February I 996. NUREG-1940, RAS C AL 4: De sc ription of Mod e ls and M e thods, U.S. Nuclear Regulatory Commission , Office of Nuclear Material Safety and Safeguards, Washington, D.C., December 2012. NUREG/CR-6410 , N ucl e ar Fu el C y cle Facili ty Ac cident Anal ys i s Handbook , U.S. Nuclear Regulator y Commission, Office of Nuclear Material Safety and Safeguards , Washington , D.C., March 1998. NWMI-2013-CALC-006 , Overall Summary Mat e rial Balan c e -MURR Targ e t Batch , Rev. D , Northwest Medical Isotopes, LLC , Corvallis , Oregon , 2015. NWMI-2013-CALC-011 , Sour ce T e rm Cal c ulation s, Rev. A , Northwest Medical Isotopes , LLC , Corvallis, Oregon, 2015. 13-98 NWMl-2013-021 , Rev. 2 Chapter 13.0 -Accident Analysis NWMI-2014-051 , integrated Safety Analysis Plan for the Radioisotope Production Fa c ility, Re v. A, Northwest Medical Isotopes, Corvallis, Oregon, 2014. NWMI-2014-CALC-014, Selection of Dominant Target i sotopes for NWMJ Material Balances, Rev. A, Northwest Medical Isotopes, LLC , Corvallis, Ore gon, 2014. NWM I-2015-RPT-009 , Fission Product R elease Eva luati on, Rev. B , Northwest Medical I s otopes , LLC, Corvallis, Oregon, 2015. WMI-2015-SAFETY-OO 1 , NWMJ Radioi sotope Produ c tion Facility Preliminary Hazards Ana l ysis, Rev. A, Northwest Medical Isotopes , Co r va lli s, Oregon , 2015. NWMI-2015-SAFETY-004, Quantitative Ri sk A nal ysis of Process Upsets Associated wit h Passive Engineering Controls Leading to Criti c ality Accident Sequences , R ev. A, Northwest Medical Isotopes , Corvallis, Oregon , 2015. Regulatory Guide 1.145, Atmospheric Disp ersio n Mod e l s for Potential Accide nt Consequence Assessments at Nuclear Power Plant s, Rev. 1 , U.S. Nuclear Regulator y Commission, Washington , D.C., February 1983. WSRC-TR-93-262 , Savannah River Site Gen eric Data B ase Development, R ev. I , Westinghouse Savannah Ri ve r Company , Savannah Ri ver Site , Aiken , South Carolina, May 1988. 13-99

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Chapter 14.0 -Technical Specifications Construction Permit Application for Radioisotope Production Facility Prepared by: NWMl-2013-021 , Rev. 1 August 2017 Northwest Medica l Isotopes , LLC 815 NW g t h Ave , Sui t e 256 Corv a llis , OR 97330 This page intentionally left blank.

NWMl-2013-021 , Rev. 1 Chapter 14.0 -Technical Specifications Chapter 14.0 -Technical Specifications Construction Permit Application for Radioisotope Production Facility NWMl-2013-021, Rev. 1 Date Published:

August 5, 2017 Document Number. NWMl-2013-021 I Revision Number. 1 T i tle: Chapter 14.0 -Technical Specifications Construction Permit Application for Radioisotope Production Facility Approved by: Carolyn Haass Sianature:

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I Rev Date 0 6/29/2015 1 8/5/2017 REVISION HI ST ORY Reason for Revision Initial Application NWMl-2013-021 , Rev. 1 Chapter 14.0 -Technical Specifications Revised By Not required Incorporate changes based on responses to C. Haass NRC Requests for Add iti onal Information "NWMI ...... ..* ... ........ *.* ' * * ! . NOITMWEST MlDfCAl ISOTOPES NWMl-2013-021 , Rev. 1 Chapter 14.0 -Technical Specifications Thi s pa ge int e ntionall y l eft blank.

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NOflTHWlST MEDK:Al NWMl-2013-021 , Rev. 1 Chapter 14.0 -Technical Specifications CONTENTS 14.0 TECHNICAL SPECIFICATIONS

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14-1 14.1 Outline ..........................................................................................

.................................... 14-2 14.1. l Introduction

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14-3 14.1.3 Limi ting C ondition of Operation

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................... 14-4 14.1.6 Administrat i ve Contro ls .......................................................................

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...... 14-5 TABLES Table 14-1. Potential Techni ca l Specifications

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14-1 14-i

..... NWMI *::**:*:* ...... . * * . NORTHWUT MEOfCAL ISOTOPlS TERMS Acronyms and Abbreviations AC administrative control ANS American Nuclear Society NWMl-2013-021, Rev. 1 Chapter 14.0 -Technical Specifications ANSI American National Standards Institute CFR Code of Federal Regulations IROFS items relied on for safety ISA integrated safety analysis LCO limiting condition of operation LSSS limiting safety system setting NWMI Northwest Medical Isotopes, LLC RAM radioactive material RPF Radioi so tope Production Facility SL safety limit SNM special nuclear material SSC systems, structures , and components 14-ii NWMl-2013-021, Rev. 1 Chapter 14.0 -Technical Specifications 14.0 TECHNICAL SPECIFICATIONS T h is chapter describes the process by which the Northwest Medical Isotopes, LLC (NWMI) Radioisotope Production Facility (RPF) technical specifications will be developed and written. For the Construction Permit Application, NWMI has prepared the strategy and content of what will be required for technical specifications during RPF operations.

No technical specifications were developed for the Construction Permit Application. The technical specifications will be included in the submission of the Operating License Application.

The variables or conditions listed in Table 14-1 are probable subjects of technical specifications based on their involvement with preventing release of radioactive materials routinely or in the event of an accident.

Table 14-1. Potential Technical Specifications Item or variable Reason Uranium mass limits on batches , samp l es , and approved containers

Spacing requirements on targets and containers with SNM" Floor and sump designs* Hot cell liquid confinement*

Process tank size and spacing* Evaporator condensate monitor Criticality monitoring system In-line uranium content monitoring Air pressure differential between zones* Ventilation system filtration*

Process offgas subsystem P r imary offgas relief system Hot cell shield thickness and integrity" Hot eel 1 secondary confinement boundary" Double-wall piping Process closed heating and cooling loops S y stem backflow prevention devices Stack height" Area radiation monitoring system Criticality control Criticality control Criticality control Criticality control Criticality control Criticality control Criticality control Criticality control Control of airborne RAM Control of airborne RAM Control of airborne RAM Control of airborne RAM Occupation and general public dose reduction Control of airborne RAM Control of liquid RAM/criticality control Control of both airborne and liquid RAM Control of liquid RAM/criticality control Control of airborne RAM Occupation and general public dose reduction a Items that will significantly influence the final design. RAM = radioactive material.

SNM special nuclear material.

14-1 NWM I ...... ' * *

NORTHWEST Mf.DtCAl ISOTOPES NWMl-2013-021, Rev. 1 Chapter 14.0 -Technical Specifications The format and content of the technical specifications for the RPF will be based on the guidance provided in American National Standards Institute/ American Nuclear Society (ANSI/ ANS) 15.1 , The Developm e nt of Technical Specifications for Re s earch Reactors; NUREG-153 7, Guidelines for Preparing and R e viewing Applications for the Lic e nsing of Non-Power Reactors:

Format and Content; and the final interim staff guidance augmentingNUREG-1537 (NRC , 2012). The technical specifications will be consistent with Title 10, Code of F e deral Regulations, Part 50.34, "Contents of Applications; Technical Information," and will address the applicable paragraphs of 10 CFR 50.36 , "Technical Specifications." However , the technical specifications will be written to address the differences between the RPF and either power or research reactors.

The proposed technical specifications will form a comprehensive set of parameters to ensure that normal RPF operations will not result in off-site radiation exposures in excess of the guidelines in 10 CFR 20 , "Standards for Protection Against Radiation ," and also reasonably ensure that the RPF will function as analyzed in the Operating License Application. Adherence to the technical specifications will limit the likelihood of malfunctions and mitigate the consequences to the public of off-normal or accident events. The RPF integrated safety analysis (ISA) process identified systems , structures , or components (SSC) that are defined as items relied on for safety (IROFS). The importance of these SSCs will also need to be reflected in the technical specifications.

Each IROFS will need to be examined and likely translated into a limiting condition of operation (LCO). This translation will involve identifying the most appropriate specification to ensure operability and a corresponding surveillance periodicity for the IROFS. An IROFS could potentially be translated into a design function but this seem s less likely than translating it into a LCO. The outline for the technical specifications that will be prepared during de v elopment of the Operating License Application is provided below. 14.1 OUTLINE 14.1.1 Introduction The introductory section will identify the scope , purpose , and format of the technical specifications.

A list of definitions will be identified to provide consistent language throughout the document.

Term Actions Administrative control (AC) ' Definition Actions are that part of a limiting condition for operation that prescribes Required Actions to be taken under designated conditions within specified completion times . ... (described in Section 14.1.6) Design features ... (described in Section 14.1.5) Limiting condition

... (described in Section 14.1.3) for operation (LCO) Limiting safety system setting (LSSS) ... (described in Section 14.1.2) 14-2 Term Modes Operable/ operability Safety limit (SL) Shall Surveillance requirements Verify/verification NWMl-2013-021 , Rev. 1 Chapter 14.0 -Technical Specifications Definition Modes are used to (1) determine safety limits, limiting control settings, limiting conditions for operation, and administrative controls program applicability, (2) distinguish facility operational conditions , (3) determine minimum staffing requirements, and (4) provide an instant facility status report. A system, subsystem, component, or device shall be operable or have operability when it is capable of performing its specified safety function(s), and (1) setpoints are within limits , (2) operating parameters necessary for operability are within limits, and (3) when all necessary attendant instrumentation, controls, electrical power, cooling or seal water, lubrication , or other auxiliary equipment that are required for the system, subsystem, component, or device to perform its safety function(s) are also capable of performing their related safety support function(s) . ... (described in Section 14.1.2) Denotes a mandatory requirement that must be complied with to maintain the requirements , assumptions , or conditions of the facility safety basi s . ... (described in Section 14.1.4) A qualitative assessment to confirm or substantiate that specific plant conditions exist. This assessment may include collecting sample data or quantitative data; taking instrument readings; recording data and information on logs , datasheets , or electronic media; and evaluating data and information according to procedures. 14.1.2 Safety Limit and Limiting Safety System Setting Safety limits (SL) will be established from basic physical conditions, as determined by appropriate proces s variables, to ensure that the integrity of the principal physical barrier is maintained ifthe SLs are not exceeded.

Limiting safety system settings (LSSS) will be established for the operation of the RPF to defend the SL. The LSSS will be limiting values for setting instrumentation by which point protective action will be initiated. SLs for radiochemical and chemical processing will be developed to maintain operations within limits pursuant to 10 CFR 50.36 to protect workers and the public. As an example , the amount of radioactive material will be limited so as not to exceed the shielding and confinement capabilities of the systems and components in which the materials are processed or stored. Each SL and LSSS will have an identified applicability, objective, specification, and basis. Currently, neither the SL no r LSSS have been specifically identified but may be part of the Operating License Application.

14.1.3 Limiting Condition of Operation Administratively established constraints on equipment and operational characteristics will be identified and described.

These limits will be the lowest functional capability or performance level required for safe operation of the facility.

Each LCO will have an identified applicability , objective, specification, and basis. The basis of each LCO will be provided and consistent with analysis provided in the Operating License Application.

Anticipated systems covered in this section include containment, ventilation, effluent monitoring , and criticality monitoring. Windows, or short time periods , of approved inoperability will be established to create operational flexibility.

The basis of these windows will be analyzed in the Operating License Application.

14-3 "NWMI ...... ** ** .*.******* ! * * . NORTHWEST MEDICAL ISOTOPES NWMl-2013-021, Rev. 1 Chapter 14.0 -Technical Specifications 14.1.4 Surveillance Requirements A set ofrequirements will provide maximum intervals for checks, tests, and calibrations for each system or component identified in Section 14.1 .3 to verify a minimum performance or operability level. The basis for each will be identified and will be derived from either an analysis presented in the Operating License Application or experience , engineering judgment, or manufacturer recommendations. 14.1.5 Design Features This section will establish the minimum design functions of safety-related SSCs , particularly construction or geometric arran g ements. These design functions , if altered or modified, are implied to significantl y affect safety and will not be identified in other sections.

Anticipated areas covered in thi s section include the site and facility description, and fissionable material storage. Design features that will be provided in the technical specification s are the features of the RPF (e.g., materials of construction and geometric arrangements) that would have a significant effect on safety if those features were altered or modified.

The requirements of 10 CFR 50.36(c)(4) are specified here as they pertain to the above referenced processes.

14.1.6 Administrative Controls This section will establish the administrative structure and controls for the RPF and will identify the roles, responsibility , and reporting line s for NWMI management (e.g., Levels 1 through 4). Other requirements include: * * *

* *

Identifying minimum staffing and supervisory functions Preparing and maintaining call lists Selecting and training per s onnel Developing a process for creating and modifying procedures Identifying actions to be taken in case of an SL violation (if applicable), exceeding an LCO , or release ofradioactivity in excess of regulatory limits Developing reporting requirements for annual operating condition s or events Specifying records retention This section will also identify the creation of a Review and Audit Committee and will address the establishment of a charter , review and audit functions , quorum requirements, membership expertise , and meeting frequency for the committee.

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NOITHWEST MEDtCAL ISOTDPH

14.2 REFERENCES

NWMl-2013-021 , Rev. 1 Chapter 14.0 -Technical Specifications 1 0 CFR 20, "Standards for Protection Against Radiation ," Code of Federal Regulations , Office of the Federal Register , as amended. 1 0 CFR 50, "Domestic Licensing of Production and Utilization Faci liti es," Code of Federal Regulation s, Office of the Fe deral R egister, as amended. ANS I/AN S 15.1 , Th e D eve lopm e nt ofTechnical Sp ec ifications for R esearch R e a c tor s, American National Standards Institute/American Nuc l ear Societ y, LaGrange Park Illin ois, 2013. NRC , 20 I 2, Final Int er im Staff Guidance Augmenting NUREG-153 7, " Guidelines for Preparing and Reviewing A pplications for the Licensin g of No n-Po we r R e a c tor s," Parts 1 and 2 , for Lic e nsing Radioisotop e Production Facilities and Aqueous Homogen e ous R e a c tors , Docket ID: NRC-2011-0135 , U.S. Nuclear Regulatory Commission, Washington , D.C., October 30, 2012. NUREG-153 7 (Part 1 ), Guidelin es for Pr e parin g and R ev i ewi ng A ppli ca tions for th e Licensing of Power R eactors: Format a nd Content, U.S. Nuclear Regulatory Commiss ion , Washington , D.C., February 1996. 14-5 NWMl-2013-021 , Rev. 1 Chapter 14.0 -Technical Specifications This page intentionally left blank. 14-6

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 | issue date = 08/05/2017
 | title = Attachment 3 to NWMI-2013-021, Rev. 2, Chapter 13.0 - Accident Analysis Construction Permit Application for Radioisotope Production Facility and Chapter 14.0 - Technical Specifications
-| author name = Haass C C
+| author name = Haass C
 | author affiliation = Northwest Medical Isotopes, LLC
 | addressee name =

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